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 ZP06: Poster Session: Magnetic Confinement: Low-Aspect Ratio Tokamaks (9:30am - 12:30pm)On Demand
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ZP06.00001: Integrated Studies of Solenoid-Free Tokamak Startup with Pegasus-III M.D. Nornberg, M.W. Bongard, S.J. Diem, J.A. Goetz, R.J. Fonck, B.A. Kujak-Ford, B.T. Lewicki, A.C. Palmer, J.A. Reusch, A.C. Sontag, G.R. Winz Minimizing or completely eliminating the need for induction from a central solenoid during startup, ramp-up and sustainment of a tokamak plasma is a critical challenge in magnetic fusion. To address that challenge, the Pegasus-III facility is under construction to provide a dedicated US platform for non-solenoidal tokamak startup and sustainment studies. Its mission is to perform comparative studies of leading techniques for solenoid-free startup and provide guidance for 1 MA startup in NSTX-U and beyond. These include: local helicity injection (LHI); coaxial helicity injection (CHI); RF-aided and RF-only startup; and PF induction. Topics of study include current drive efficiency and confinement scalings at increasing $B_{T}=0.6$ T and $I_{p}=0.3$ MA, testing synergistic effects by combining techniques, and supporting technology development. The new facility features: a solenoid-free centerstack; an innovative 24-turn TF coil system with dynamic stress compensation; new divertor coils that also support CHI; new digital control systems for all magnet and HI power systems; and an initially-modest electron Bernstein wave (EBW) heating system. The science program will first establish high-$I_{p}$ LHI scenarios, followed by deployment and test of transient CHI, and eventually a modest sustained CHI system. Low-power EBW studies will be developed in the second and third year of operation. Expansion of the RF systems to provide RF-only initiation and sustainment is under evaluation. [Preview Abstract] |
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ZP06.00002: CHI Research on Pegasus-III R. Raman, J.A. Reusch, J.D. Weberski, M.W. Bongard, F. Ebrahimi, R.J. Fonck, A.C. Palmer, T.R. Jarboe, B.A. Nelson, M. Ono, G.R. Winz The spherical tokamak (ST) may require and the advanced tokamak would considerably benefit from the elimination of the central solenoid. \textsc{Pegasus}-III is a ST non-solenoidal startup development station under design and fabrication dedicated to solving the startup problem. On \textsc{Pegasus}-III, Transient and Sustained coaxial helicity injection (T- and S-CHI) will be explored, as well as possible synergies of CHI with local helicity injection and EBW heating and current drive. T-CHI has shown promising capability on the HIT-II and NSTX STs. However, in both these machines the vacuum vessel was electrically cut. For reactor applications a simpler biased electrode configuration is required in which the insulator is not part of the external vacuum vessel. To develop this capability \textsc{Pegasus}-III will use a double biased electrode configuration, which would a first of its kind for the reactor-relevant development of the CHI concept. The system is capable of generating plasma start-up currents at the levels that can be supported by the external poloidal field coils, which is estimated to be $\sim 300$ kA. The CHI design for \textsc{Pegasus}-III will be described. [Preview Abstract] |
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ZP06.00003: Magnetic Activity During LHI Startup and Sustainment N.J. Richner, G.M. Bodner, M.W. Bongard, S.J. Diem, R.J. Fonck, M.D. Nornberg, C.E. Schaefer, J.D. Weberski Local helicity injection (LHI) is a non-solenoidal tokamak startup technique using biased plasma sources for DC helicity injection. This process relies upon magnetic reconnection and relaxation mechanism(s) that convert the helicity from injected current streams into bulk plasma current through helicity-conserving instabilities. To inform this process, high-bandwidth local magnetic measurements have been obtained in a broad survey of LHI operational ($I_{inj}$, $V_{inj}$, $B_{T}$, injector geometry) and physics regimes ($e.g.$ stream-only, actively driven, decaying, etc.). Significant broadband high-frequency activity is present in LHI discharges compared to Ohmic plasmas. $\tilde{B}\thinspace $features power-law behavior with spectral indices of $\sim 5/3$ for $f$\textless $f_{ci}\thinspace $and $\sim 8/3$ for $f$\textgreater $f_{ci}$. Similar signatures are attributed to MHD and KAW/whistler wave turbulence, respectively, in astrophysical contexts, and is predicted to have an inverse cascade of magnetic helicity. Such turbulence has also been observed in reconnection systems. High frequency activity $f$\textgreater $f_{ci}$ is correlated with LHI drive voltage $V_{inj}$ and/or injected beam velocity $v_{b}\propto V_{inj}^{1/2}$, further suggesting a kinetic role. Activity at $f\sim 2\thinspace $MHz (2--4 $f_{ci})$ is found to scale linearly with applied LHI drive. Its potential role in the current drive process is under investigation. [Preview Abstract] |
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ZP06.00004: Ohmic Sustainment of Local Helicity Injection Initiated Plasmas on the Pegasus ST C. Pierren, G.M. Bodner, M.W. Bongard, S.J. Diem, R.J. Fonck, M.D. Nornberg, N.J. Richner, C. Rodriguez Sanchez, C.E. Schaefer Local helicity injection (LHI) is being developed on the \textsc{Pegasus} ST for non-solenoidal tokamak startup. The startup viability of LHI-initiated plasmas will depend upon how readily they efficiently handoff to subsequent heating and current drive (CD) sustainment phases. Final experiments on the \textsc{Pegasus} ST tested handoff to sustainment using the Ohmic solenoid (OH). LHI-initiated plasmas were found to robustly couple to OH CD, even at high $I_{p}$ ramp rates \textgreater 100 MA/s. The relative impurity content during LHI appeared higher than comparable OH-only discharges; however, these impurities appeared to quickly decay to OH-only like levels during the OH phase of handoff scenarios. Magnetic energy was readily conserved across the LHI-OH handoff as evidenced by minimal drops in $I_{p}$. The LHI target's $j\left( R \right)$ profile appears to be MHD favorable as the onset of internal low $m$, $n=1$ tearing modes typical of \textsc{Pegasus} OH discharges was mitigated with LHI startup. Unique high $\beta_{T}\sim 1$ plasmas at extremely low $B_{T}$ could not be Ohmically sustained because they suffered from an edge kink instability that was presumably stabilized by the LHI edge current streams. Reconstructions and stability analyses are underway to further explore the MHD stability properties of LHI plasmas. [Preview Abstract] |
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ZP06.00005: The NSTX-U Research Program Stanley Kaye, Devon Battaglia, Walter Guttenfelder, Rajesh Maingi NSTX-U is a high-powered Spherical Tokamak (ST) whose mission is to establish the physics basis for next-step ST facilities, broaden the scientific understanding of plasma confinement for ITER, and maintain U.S. world leadership in ST research capabilities. In particular, the research in NSTX-U will be critical for informing the design of a Compact Fusion Pilot Plant. NSTX-U will have a mission-oriented research program that addresses key gaps over the next five years. 1) Extend confinement and stability studies to low collisionality and high-$\beta $. NSTX-U will operate at collisionalities 5x lower than NSTX, allowing validation of the strong improvement of core and pedestal confinement with decreasing collisionality. 2) Develop operation at large bootstrap fraction and advance the physics basis required for non-inductive, high-performance and low-disruptivity operation. NSTX-U will develop operational scenarios with f$_{\mathrm{BS}}=$60--90{\%} and $\beta_{\mathrm{N}}=$4-6 for multiple current redistribution times, and it will operate fully non-inductively for plasma currents up to 1 MA. 3) Develop and evaluate conventional and innovative power and particle handling techniques to optimize plasma exhaust in high performance scenarios. Research in this area will evolve towards implementation of flowing liquid lithium components. [Preview Abstract] |
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ZP06.00006: Upgrades to the Materials Analysis and Particle Probe for the next NSTX-U campaign Hanna Schamis, Camilo Jaramillo, Jean Paul Allain Boron and lithium conditioning of plasma-facing components have both improved plasma performance in previous NSTX campaigns. Controlled laboratory experiments have shown that oxygen plays a role in hydrogen retention in B and Li. This may contribute to low recycling regimes that improve plasma performance. In the next NSTX-U campaign, boronization will be the first conditioning technique used, followed by lithium conditioning. The interactions of these mixed layers with H and O is not well understood and needs to be further explored. In addition, more research needs to be done on the link between plasma performance and wall conditioning. In order to contribute to these pending surface science questions, the Materials Analysis and Particle Probe (MAPP) will be reinstalled with upgraded capabilities. These include a new hemispherical analyzer, which will improve the energy resolution of the acquired data. This will allow for better operation of diagnostics, as well as more detailed stoichiometric data of the B or Li layers. Additionally, a new custom-designed in situ quartz microbalance will allow for real-time measurements of material deposition and erosion at the MAPP location. These surface measurements will benefit modeling and understanding of wall conditioning. [Preview Abstract] |
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ZP06.00007: Diagnostic Ungrades for NSTX-U Calvin Domier, Jon Dannenberg, Yilun Zhu, Yang Ren, Brent Stratton, Neville C. Luhmann, Jr. UC Davis is currently funded to develop three plasma diagnostics for NSTX-U. An 8-channel 693 GHz poloidal high-k scattering system to study high-k electron density fluctuations on NSTX-U. The probe beam is launched from Bay G towards Bay L where optics collect radiation at 8 simultaneous scattering angles ranging from 2 to 15°. This yields measurement of poloidal wavenumbers from 7 cm-1 to >40 cm-1, while translatable optics allow placement of the scattering volume from r/a = 0.1 out to the pedestal region. The second is the Far-Infrared Tangential Interferometer/ Polarimeter (FIReTIP) system, which will provide real-time density feedback control for the Real-Time Control and Protection (RTCP) system. The third is a 5×4 channel (5 poloidal, 4 radial) microwave imaging reflectometer (MIR) system employing system-on-chip (SoC) transmitter and receivers and sharing the same Bay L port as high-k scattering. Details of the diagnostics, including installation and commissioning schedules, will be presented. [Preview Abstract] |
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ZP06.00008: Evaluating NSTX-U Operational Space Relative to PFC Engineering Limits Tom Looby, Matthew Reinke, Andreas Wingen, David Donovan, Mike Messineo, Jonathan Klabacha Sustaining high beam power on NSTX-U, $ \text{P}_{NBI} \sim 10 $ MW for $ \Delta \text{t} \sim 5 $s , may be limited by overheating of the divertor plasma facing components (PFCs). New castellated and fish-scaled graphite PFCs are inherently 3-dimensional, creating non-axisymmetric features that vary with plasma shape. A new Heat flux Engineering Analysis Toolkit (HEAT) has been developed to simulate 3D plasma effects coupled to 3D CAD geometry. Features of HEAT are described and predictions compared with 2D models originally used to scope design requirements. Results show how HEAT can be used to map an operational space for NSTX-U to reduce the risk of PFC failure. A side by side comparison of the 2D design analysis with 3D HEAT results is provided for a variety of discharge configurations and features that are missed with a 2D toroidally symmetric model are highlighted. [Preview Abstract] |
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ZP06.00009: Current-Drive via Plasmoid-Mediated Reconnection in Spherical Tokamaks Fatima Ebrahimi, Roger Raman Fully solenoid-free current start-up is one of the primary objectives of the ST program.[1] Due to its simplicity and favorable scaling, the transient helicity injection technique via plasmoid reconnection [2] has shown to be a promising startup method for advanced ST scenarios. We investigate stability and physics of plasmoid mediated reconnection during startup helicity injection at high current in spherical tokamaks. Our recent 3-D NIMROD simulations [3] in NSTX/NSTX-U have shown stable current-carrying axisymmetric plasmoid formation. Here, we first examine the accessibility to the regimes of maximum current (MA level) of start-up plasma formation with increased toroidal and injector fluxes. The goal is to achieve ohmically self-heated plasma formation via generation of large axisymmetic plasma current in the absence of large 3-D fluctuations. Preliminary results of 3-D simulations of plasmoid reconnection during helicity injection in PEGASUS-like configuration will also be presented. [1] M. W. Bongard et al. APS-DPP-CPP initiative whitepaper (2019).[2] F. Ebrahimi, R. Raman, PRL 114, 205003 (2015); [3] F. Ebrahimi PoP, 26, 092502 (2019). Work supported by DOE grants DE-AC02-09CHI1466, and DE-SC0010565. [Preview Abstract] |
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ZP06.00010: Fast Ion Transport by Coupled Tearing and Kink Modes in NSTX Jeong-hun Yang, Mario Podesta, Eric Fredrickson Fast ion transport by coupled tearing and kink modes is observed in NSTX. The relative phase of tearing and kink modes affects the fast ion pressure up to 10{\%} and NB-driven current density by up to 20{\%} in the low aspect ratio plasma. The fast ion interactions with either tearing or kink modes have been reported, but the interaction with coupled tearing and kink modes is reported for the first time. The analysis uses the ``Kick'' model in TRANSP, where fast ion dynamics in perturbed magnetic field are computed using the particle-following code ORBIT, with input provided by analytic eigenfunctions of tearing and kink modes scaled based on SXR data. With experimental inputs only, the modeled neutron rate agrees with the measured neutron rate, whereas with classical runs the neutron rate is overestimated by 20{\%}. It is found that the relative phase of tearing and kink modes affects the fast ion transport and the mode combination affects the energy transfer between modes and fast ions. The synergetic effects of the tearing and kink modes in fast ion transport suggest that fast ion distributions may contribute to the mode coupling as well. [Preview Abstract] |
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ZP06.00011: Relevant time-scales for the assessment of Alfven Eigenmode stability and saturation Mario Podesta Energetic particles from Neutral Beam (NB) injection, fusion reactions or RF acceleration can destabilize Alfvenic instabilities (AEs) that, in turn, redistribute energetic particles thus potentially leading to degraded plasma performance. This work investigates the characteristic time scales that need to be included in the assessment of AE stability and saturation. NB-heated NSTX-U scenarios featuring simultaneous low-frequency kink modes and Toroidal AEs are used as reference. Analysis is performed through the NUBEAM fast ion module of TRANSP, enhanced by the reduced-physics kick model for fast ion transport by instabilities [M. Podestà et al., Plasma Phys. Control. Fusion 59 (2017) 095008]. The results indicate that time scales from hundreds of microseconds or shorter up to several NB slowing down times need to be included for quantitative estimates of AE saturation. This suggests that time dependent simulations are generally required, instead of a simpler time-slice analysis. Remaining issues for a reduced fast ion transport and AE stability model for integrated simulations will be discussed. [Preview Abstract] |
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ZP06.00012: Core-Edge Coupled Modeling for Fast Ion Transport Xin Zhang, Francesca Poli Fast particles in tokamak experiments have large drift orbits that frequently extend into the Scrape-Off-Layer (SOL). The confinement of these particles are therefore influenced by the SOL, which subsequently impacts the overall heating and confinement of the core plasma. In order to study the effect of SOL on fast particle confinement, we have developed a full orbit fast ion tracer code that samples the entire tokamak plasma. The code uses a reduced SOL model and fast grid generator to provide a time-dependent plasma background for the SOL, whereas the core plasma profiles are provided by TRANSP. The particle motions are integrated with the Boris algorithm, with collisions against the thermal plasma modeled by the Langevin equation for Coulomb collisions and solved with a newly developed energy conserving algorithm [Zhang, Fu, and Qin, arXiv:2006.10877]. The coupled evolution of the plasma can therefore be studied self-consistently, which could provide valuable insights into existing experiments and aim future experimental design. The simulations will be performed using NSTX/NSTX-U plasma profiles. [Preview Abstract] |
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ZP06.00013: Active Radiative Liquid Lithium Divertor for Handling Transient High Heat Flux Events Masayuki Ono, Roger Raman The extreme heat flux anticipated in fusion reactor divertor plasma facing components (PFCs) is perhaps the most challenging technology issue for fusion energy development. Most divertor PFCs are designed based on the maximum steady-state operational limits. However, in addition to the high steady-state heat flux, the fusion reactor divertor PFCs could also experience significant transient heat flux such as ELMs and/or other magnetic reconnection events which can deposit large transient heat flux onto the divertor PFCs. If unprotected, it could damage the divertor PFC surfaces which could lead to a highly undesirable unplanned shutdown for PFC repair and/or replacement. In this presentation, we explore feasibility of active radiative liquid lithium divertor concepts for protecting the divertor PFCs from the extreme transient heat flux while maintaining the normal plasma operations. We also suggest a possible implementation technique using inductive pellet injector for the reactor PFC protection from transient heat flux which can be tested on NSTX-U. [Preview Abstract] |
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ZP06.00014: Feasibility study for a high-k temperature fluctuation diagnostic based on soft X-ray imaging Xiang Chen, Juan Ruiz Ruiz, Nathan Howard, Walter Guttenfelder, Jeff Candy, Jerry Hughes, Robert Granetz, Anne White Turbulence transport can significantly limit fusion gain. A deep understanding of turbulent transport requires sufficient experimental data of turbulence quantities, such as the fluctuations of electron density and temperature. In this work, we explored the feasibility of using soft X-ray imaging to measure electron temperature fluctuations in NSTX-U, a spherical tokamak. We devised a pseudolocal tomography algorithm to reconstruct local electron temperature fluctuations from the measurements of line-integrated soft X-ray emissivity and a model to test this algorithm. The reconstructed wavenumber spectrum of the electron temperature fluctuations is optimized by varying the number of viewing chords and viewing angles of X-ray detectors to best match the synthetic wavenumber spectrum. The dependence of the reconstructed results on the relationship between the emissivity and the electron temperature is studied. A sensitivity analysis has also been done to demonstrate the requirements that the X-ray detector needs to fulfill if we want to build such a realistic diagnostic based on this idea. The requirements include the aperture, the time resolution, the electronic noise level and other factors. [Preview Abstract] |
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ZP06.00015: Linear analysis of geodesic curvature driven instability in divertor legs in spherical tokamaks D. A. Baver, J. R. Myra, F. Militello, D. Moulton Turbulence in the divertor region impacts the heat flux width, which is important for evaluating advanced divertor configurations such as the super-X. In this work, the ArbiTER$^{\mathrm{1}}$ code is used to analyze the underlying linear instabilities that drive divertor leg turbulence. Simulations of the divertor leg of MAST-U have revealed a type of instability that is driven predominantly by geodesic curvature. This drive mechanism allows a ballooning-type mode to exist in regions that would normally be categorized as having good curvature. This type of instability is particularly prominent in the MAST-U super-X divertor, but also exists in the standard divertor configuration. We will compare the location, mode structure, and growth rates of these modes in different divertor configurations and examine their toroidal mode number spectrum. 1. D. A. Baver, J. R. Myra and M. V. Umansky, Comm. Comp. Phys. 20, 136 (2016). [Preview Abstract] |
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ZP06.00016: Toroidicity-induced Alfv\'{e}n eigenmode structure measured with multiple diagnostics in MAST H.H. Wong, C A. Michael, N.A. Crocker, T. Carter, A.R. Field, N. Fil, M. Fitzgerald, A. Jacobsen, K.G. McClements, P. Liu, H. Wang, G.J. Choi, Z. Lin, M. Cecconello, D. Dunai Measurements of plasma fluctuations associated with toroidicity-induced Alfv\'{e}n eigenmodes (TAE) are obtained using multiple diagnostics in the MegaAmp Spherical Tokamak (MAST). The redistribution and loss of fast-ions caused by the excitation of TAEs would potentially reduce the effectiveness of beam heating and pose a threat to the device. In this study the contribution of TAEs to measured plasma fluctuations is isolated using cross-correlation analysis of measurements from Mirnov coils, beam emission spectroscopy (BES), motional Stark effect (MSE) and tangential soft x-ray (SXR) diagnostics on MAST. The SXR and MSE data allow measurement of the radial structure of TAE across a large spatial range. In contrast, BES offers a 2D view in the poloidal plane that is comparatively narrow in the radial direction, with much finer spatial sampling that can be adjusted from pulse to pulse to view different portions of the radius. The measurements from BES, SXR and MSE are to be compared with simulations carried out with the NOVA and MISHKA ideal MHD linear eigenmode codes, as well as linear simulations carried out with the Gyrokinetic Toroidal Code (GTC)[1]. In future work, details of time-dependent, shorter scale radial and poloidal structures obtained from BES will be compared with nonlinear simulations from GTC, thus providing a basis for validating nonlinear physics models for TAE, e.g. those based on the concept of hole-clump instabilities. [1] Z. Lin el al., Science 281, 1835(1998) [Preview Abstract] |
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ZP06.00017: Development of weight functions for 3 MeV proton diagnostics in FIDASIM Alvin Garcia, William Heidbrink, Werner Boeglin, Alexander Netepenko, Mirko Salewski, Luke Stagner The weight function of a diagnostic is important in Energetic Particle research. It describes the phase space sensitivity of a diagnostic that is used in forward modelling and tomographic reconstructions of the fast-ion distribution function. The challenges of the weight function calculation for a 3 MeV proton diagnostic are attributed to the curved ``sightline'' trajectories, which are dependent upon the proton velocity. For example, curves must be considered in the calculation of the detector solid angle accepted by the collimating structure. At MAST, the charged fusion product diagnostic (CFPD) measures 3 MeV protons from beam-target fusion reactions between fast ions and thermal deuterium in the plasma core. In an effort to simulate energy-resolved proton signals and use CFPD data for tomographic inversions, the weight function of a 3 MeV proton diagnostic is developed. This poster discusses the implementation of the algorithm in the FIDASIM framework and its benchmark against existing independent calculations at MAST. [Preview Abstract] |
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ZP06.00018: Overview of Results and Plans at the Lithium Tokamak Experiment -- $\beta $ D.P. Boyle, R.E. Bell, P.E. Hughes, R. Kaita, B.P. LeBlanc, A. LeViness, A. Maan, R. Majeski, E. Merino, X. Zhang, J.K. Anderson, W. Cappechi, P. Beiresdorfer, F. Scotti, V.A. Soukhanovskii, T.M. Biewer, D.B. Elliott, D. Donovan, C. Hansen, B. Koel, E. Ostrowski, S. Kubota, T. Rhodes, N. Yoneda, L.E. Zakharov Following extensive investigation of plasmas almost fully surrounded by solid or liquid lithium walls coatings, the Lithium Tokamak Experiment was upgraded to LTX-$\beta $ with the addition of a neutral beam, higher magnetic fields, and improved diagnostics. The main goal of LTX-$\beta $ is to extend the low-recycling regime first observed in LTX to higher performance, steadier discharges with neutral beam heating and fueling. Initial measurements of confinement and plasma-material interactions have been made with new and enhanced diagnostics, and additional diagnostic upgrades underway will allow better kinetic profiles and the determination of wall recycling. Machine upgrades also in progress will further increase plasma current, enhance neutral beam coupling, and improve lithium wall conditioning. [Preview Abstract] |
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ZP06.00019: Plasma performance dependence on lithium surface conditions for the Lithium Tokamak eXperiment-$\beta$ A. Maan, D.P. Boyle, R. Kaita, E.T. Ostrowski, D.C. Donovan, R. Majeski, B.E. Koel, T.M. Biewer, P.E. Hughes, C. Hansen, V. Soukhanovskii The characteristics of lithium-coated plasma-facing components (PFCs) have been correlated with plasma performance on the Lithium Tokamak eXperiment-$\beta$ (LTX-$\beta$). Previous experiments on LTX showed that the application of lithium to PFCs was needed to achieve higher performance discharges with flat electron temperature profiles and high edge temperatures. Samples that match the LTX-$\beta$ PFCs were exposed to plasmas before and after PFCs were coated with lithium and transferred under vacuum to a surface analysis station. Measurements using X-ray photoelectron spectroscopy (XPS) revealed that the primary surface constituent was lithium oxide. Earlier XPS analysis of lithium-coated PFCs on LTX was only able to show the presence of surface oxygen. The new XPS data from LTX-$\beta$ have sufficient resolution to clearly identify lithium compounds for the first time, and enable them to be correlated with how lithium-coated PFCs can reduce impurities and retain hydrogen to reduce recycling. Measurements from the Lyman-$\alpha$ detector array, newly installed to view the high field side limiting surface, are presented to illustrate progress made towards recycling measurements for LTX-$\beta$. [Preview Abstract] |
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ZP06.00020: MHD Dynamics in Beam-Injected LTX-$\beta$ Plasmas P.E. Hughes, W. Capecchi, D.B. Elliott, L.E. Zakharov, R.E. Bell, C. Hansen, D.P. Boyle, R. Majeski, R. Kaita The implementation of a toroidal Mirnov array in the Lithium Tokamak eXperiment--Beta (LTX-$\beta$) has provided the means to study MHD stability and mode dynamics in neutral beam injected LTX-$\beta$ plasmas [D.B. Elliott et al. IEEE TPS April 2020]. PSI-Tri reconstructions enhanced by the addition of Thomson scattering profiles inform stability analysis, as well as providing key transport and profile parameters to model beam-plasma coupling. An array of particle tracking codes is employed to model fast ion confinement for comparison against experimental measurements, including pressure and temperature evolution, fueling as characterized by electron density, and the observed acceleration of MHD mode rotation in the counter-beam, counter-IP direction due to fast ion losses. [Preview Abstract] |
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ZP06.00021: Optimized beam fueling in LTX-$\beta $ William Capecchi, Jay Anderson, Ron Bell, Dennis Boyle, Paul Hughes, Robert Kaita, Richard Majeski, Drew Elliott, Chris Hansen, Leonid Zakharov The LTX-$\beta $ upgrade completed last year included installation of a new 20kV neutral beam for heating and fueling, but initial operation of the beam showed high first orbit losses. In the next phase of operation, core fueling though neutral beam injection (NBI) will be essential for studying the low recycling regime where cold edge fueling is undesirable. Doppler spectroscopy is used along with beam dump/scraper RTDs to analyze beam geometry and maximize throughput into the torus for various beam operational modes. Full ion orbit codes are employed to model beam coupling and deposition and to investigate drift velocities in various combinations of toroidal field and plasma current directions to optimize first orbit confinement. Here we report results of the beam performance optimization and map out a path to maximize neutral beam fueling of LTX-$\beta $ plasmas. [Preview Abstract] |
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ZP06.00022: Measurement of Hydrogen Molecular Rotational Temperatures in LTX-$\beta$ for the Thermometry of Plasma-Facing Lithium Surfaces Nao Yoneda, Filippo Scotti, Ronald Bell, Taiichi Shikama, Paul Hughes, Anurag Maan, Vlad Soukhanovskii, Dennis Boyle, Richard Majeski, Kazuaki Hanada, Masayuki Ono, Masahiro Hasuo LTX-$\beta$ is a spherical tokamak device operated with plasma-facing surfaces coated with solid and liquid lithium. We measured the variation of the H$_{2}$ $d$-state rotational temperature ($T_{\rm rot}$) in LTX-$\beta$ to deduce the surface temperature because phase change or passivation of lithium may reduce the accuracy of conventional surface thermometry using an infrared camera. H$_{2}$ emission line spectra (Q1-Q5(0-0) of Fulcher-$\alpha$ band) were observed under three surface conditions: fresh solid lithium, passivated solid lithium, and liquid lithium. We used two radial viewing chords directed to the inboard limiter and a spectrometer (F/1.8, and 70 pm wavelength resolution) for the measurement. $T_{\rm rot}$ was estimated from the relative intensities of the emission lines assuming a Boltzmann distribution. The estimated $T_{\rm rot}$ was higher for the liquid surface than for the fresh solid surface. No obvious dependence of $T_{\rm rot}$ on the electron density near the limiter was observed for densities in the range of 0.2--1.2$\times 10^{18}$ ${\rm m}^{-3}$. [Preview Abstract] |
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ZP06.00023: Negative Triangularity Operation of LTX Dick Majeski, W. Capecchi, C. Hansen LTX-$\beta$, the upgrade to the Lithium Tokamak Experiment, is a high-field side limited, low aspect ratio tokamak (with R/a nominally 1.6). We have recently been exploring the use of a pair of poloidal field coils, which are internal to the vacuum vessel, but external to the lithium-coated liner system, to generate tokamak discharges with negative triangularity. Negative triangularity equilibria have several features which are of interest with low recycling walls, in LTX. Reduced drive for trapped electron modes, in combination with flat temperature profiles, may further reduce transport in a tokamak. While the collisionless scrape-off layer (SOL) in conventional, high field side limited discharges (with lithium walls) is expected to have a large, mirror trapped population, the SOL in negative triangularity discharges should have no trapped population. Diverted negative triangularity discharges may also be possible, while positive triangularity diverted discharges are not feasible in LTX, with the present coilset. Here we discuss possible negative triangularity equilibria which can be achieved with the LTX coilset. We will also briefly discuss some (speculative) implications for reactors. [Preview Abstract] |
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ZP06.00024: Numerical simulations for evaluation of EBW Heating development in LTX-$\beta$ Bhavya Kenia The electrostatic Electron Bernstein wave (EBW) can propagate at frequencies near the electron cyclotron frequency throughout the over dense plasma of a Spherical Tokamak but not in vacuum or low-density extreme edges. A scheme to couple to the wave exploits physics that allows X-mode or O-mode wave to mode convert to an EBW at the plasma edge. The mode conversion efficiency is expected to sensitively depend on the electron density scale length (L$_{n}$) at the Upper Hybrid Resonance (UHR) layer with a theoretical maximum of 100\%. Full wave modelling of the O-X coupling in LTX-$\beta$ shows that at a moderate edge density, an O mode launched with finite k$_{||}$ gives optimized coupling efficiency greater than 65\%. At very steep edge density profiles, a normal X mode launch gives highest coupling. With a recently upgraded toroidal field capability to B$_{0}$ $\leq$ 3.4 kG in LTX-$\beta$, a 9.3 GHz launch frequency allows a range of narrow heat deposition across the entire minor radius. Genray ray tracing of EBW propagation launched at the UHR layer just inboard of the LCFS yields a span of the radial positions at which localised deposition occurs – core deposition at the fundamental cyclotron resonance for B = 3.0 kG and an edge deposition at radius r/a $>$ 0.7 for B = 2.05 kG. [Preview Abstract] |
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ZP06.00025: Modeling of losses and confinement of NBI ions in LTX-beta tokamak Leonid Zakharov, Wiliam Capecchi The LTX-$\beta$ device in PPPL ($R=0.36,\ a\simeq0.24$ m, $b/a\simeq 1.6$) with litium coated wall surface recently upgraded its capacities by installation of 20/30 keV/A NBI. Three years ago LTX demonstrated zero recycling in the transient phase of a decaying plasma. With NBI installed the low recycling regimes are envisioned for the flat top phase. Because of presently low toroidal field 0.3 T and plasma current $I_{pl}\simeq75$ kA, the Larmor radius is as large as 7 cm for 20 keV H-ions. This raises concerns with their first orbit losses. Both NUBEAM of TRANSP and a special particle orbit code {\tt 3Dorb} with tracing of full and guiding center orbits were used for understanding the problem. The range of $I_{pl}=75\ --\ 150$ kA was considered in order to make specific guidance on necessary $I_{pl}$ enhancement. The effect of the plasma charge due to ion losses was shown as suppressing the losses. While first orbit losses occur within 5-10 mksec, the confined fraction of hot ions has the confinement time of about 10 msec based on {\tt 3Dorb} simulations of hot ion collisions with the plasma. This promises a high performance regime, given that the first orbit loss problem solved. The results of comprehensive analysis of hot hon losses and confinement are presented. [Preview Abstract] |
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(631) 591-4000
Office of Public Affairs
529 14th St NW, Suite 1050, Washington, D.C. 20045-2001
(202) 662-8700