Bulletin of the American Physical Society
60th Annual Meeting of the APS Division of Plasma Physics
Volume 63, Number 11
Monday–Friday, November 5–9, 2018; Portland, Oregon
Session YO5: Spherical Tokamaks, Scenarios, Alternate Configurations |
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Chair: Theodore Golfinopoulous, Massachusetts Institute of Technology Room: OCC B113-114 |
Friday, November 9, 2018 9:30AM - 9:42AM |
YO5.00001: MAST Upgrade Status Update Nizar Ben Ayed 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, pellet injector, a cryoplant to serve the cryopumps and 2 additional neutral beams to increase the heating power from 5 to 10MW. Plans for the completion of MAST Upgrade and associated infrastructure will be presented and preparations for achieving the first plasma. |
Friday, November 9, 2018 9:42AM - 9:54AM |
YO5.00002: Status and Plans for the NSTX-U Recovery Project Jonathan Edward Menard, Stefan P Gerhardt, Russ Feder The NSTX Upgrade device began operation in 2016 and performed 10 weeks of commissioning activities and initial scientific research. However, a number of technical issues, including the failure of a divertor magnetic field coil, resulted in the suspension of operations. In response, a facility-wide “Extent of Condition” review was initiated at the request of the Department of Energy. This review generated a comprehensive corrective action plan and organization of a dedicated “Recovery Project” to enable NSTX-U to be the most capable Spherical Tokamak in the world program while also improving facility reliability. There are eight major scope items in the NSTX-U Recovery Project including: (1) six redesigned inner PF coils, (2) redesigned upper and lower polar region structures, (3) redesigned select plasma facing components, (4) improved bake-out, (5) additional component stress/strain trending instrumentation, (6) enhanced test cell shielding, (7) implementation of the accelerator safety order, and (8) reassembly of NSTX-U components with improved alignment. Progress, status, and plans for the NSTX-U Recovery Project will be described. |
Friday, November 9, 2018 9:54AM - 10:06AM |
YO5.00003: Research Directions on the Pegasus Toroidal Experiment J. A. Reusch, G. M. Bodner, M. W. Bongard, R. J. Fonck, C. Pierren, A. T. Rhodes, N. J. Richner, C. Rodriguez Sanchez, C. E. Schaefer, J. D. Weberski In recent years, the Pegasus research program has focused on developing the physics basis and predictive models for non-solenoidal tokamak startup using Local Helicity Injection (LHI). This has resulted in demonstrating solenoid-free ST plasma startup to ~ 0.2 MA. An expansion of the scope of this activity to enable a comprehensive examination of non-solenoidal startup is planned. This will include the deployment and direct comparison of leading startup techniques in a single experiment. Proposed new capabilities include: increasing BT 4× to 0.6 T to support critical scaling tests to near-NSTX-U field levels; deploying impurity and internal plasma kinetics diagnostics; advanced LHI injectors with shaped electrodes and active control of the helicity injection rate; sustained and transient coaxial helicity injection; tailored poloidal field induction; and a modest (200–400 kW) EBW RF heating and current drive capability. These efforts will address scaling of LHI to higher Ip and BT, comparative studies of helicity injection techniques, and the use of EBW to improve target plasmas for subsequent non-inductive sustainment. The ultimate goal is to validate techniques for producing a ~1 MA plasma in NSTX-U and beyond. |
Friday, November 9, 2018 10:06AM - 10:18AM |
YO5.00004: Non-Solenoidal Startup Using High-Field-Side Local Helicity Injection on the Pegasus ST G. M. Bodner, M. W. Bongard, R. J. Fonck, C. Pierren, J. A. Reusch, A. T. Rhodes, N. J. Richner, C. Rodriguez Sanchez, C. E. Schaefer, J. D. Weberski Local Helicity Injection (LHI) is a non-solenoidal startup technique that utilizes electron current injectors at the plasma edge to initiate a tokamak-like discharge. Injection on the high-field-side (HFS) provides significantly more helicity input than the low-field-side (LFS), but geometry constraints and increased PMI narrow the operating space as BT increases. LFS-to-HFS handoff enables full-field operation up to Ip ~ 0.2 MA. Thomson measurements show a flat Te profile (Te,0 ~ 50 eV) during initial LFS startup that transitions to peaked Te and ne profiles (Te,0 ~ 125 eV, ne,0 ~ 1×1019 m-3) during the HFS drive. During HFS injection, high-amplitude n = 1 magnetic oscillations, attributed to large-scale instability of the injected current streams, can abruptly disappear while broadband magnetic fluctuations in the plasma edge region shift to higher frequency and presumably shorter wavelengths. This transition is coincident with a 10–20% increase in the net current drive efficiency. Initial visible bremsstrahlung measurements indicate average Zeff < 2.5 at the end of the LHI injection phase. New experiments with the HFS injectors moved to smaller Rinj will test scalability to higher Ip via increased helicity injection rate. |
Friday, November 9, 2018 10:18AM - 10:30AM |
YO5.00005: Enhanced Pedestal H-mode Regime on NSTX Devon Battaglia, Ronald E Bell, Stefan P Gerhardt, Walter Guttenfelder, Stan Kaye, Rajesh Maingi The Enhanced Pedestal (EP) H-mode on NSTX features the development of a wider pedestal with a significant increase in the ion temperature and carbon rotation gradients in an ELM- and MHD-free period following an ELM. These discharges achieve improved energy and momentum confinement with a beneficial decrease in the impurity accumulation relative to a standard ELM-free H-mode regime. The bifurcation to improved confinement is triggered by a transient period of lower collisionality during the ELM recovery where the ∇Ti increases consistent with neoclassical ion thermal transport. The larger ∇Ti reinforces the lower edge collisionality via a reduction of the neoclassical impurity pinch and moving the region of large Er shearing rate and the density pedestal toward the core. The larger ∇Ti also acts to stabilize electron-scale turbulence, improving the electron thermal confinement. The favorable positive feedback initiated by transiently accessing lower ion collisionality leading to a significant improvement in the thermal confinement and reduced impurity accumulation motivates the development of actuators for controlling the edge density that are compatible with large core density and heat flux mitigation. |
Friday, November 9, 2018 10:30AM - 10:42AM |
YO5.00006: Validation of novel hybrid scale ETG simulations in NSTX via comparisons of simulated turbulence with a new high-k scattering synthetic diagnostic. Juan Ruiz Ruiz, Walter Guttenfelder, Anne Elisabeth White, Yang Ren, Jeff Candy, Nathan T Howard A rigorous validation study of ETG turbulence in an NSTX H-mode using nonlinear gyrokinetic simulation and a new high-k scattering synthetic diagnostic is presented, suggesting that electron scale turbulence (e- scale) can account for experimental heat fluxes (Qe) in highly unstable ETG regimes, but cross-scale interactions might be necessary to match Qe in marginally stable turbulence regimes. At high ETG drive, ion scale modes are stabilized by ExB shear and e- scale simulations (kθρs = [1.5, >40]) can marginally match Qe. Hybrid scale simulations capturing both ion and e- scale modes (kθρs = [0.3, >40]) underpredict Qe. At low ETG drive, hybrid scale simulations can match the experimental Qe within uncertainty (dominant transport is from e- scales with non-zero transport from ion scales), but e- scale simulations underpredict Qe. Quantitative comparisons of the experimental high-k fluctuation spectra with a new high-k scattering synthetic diagnostic reveal that the measured frequency spectrum is not a critical constraint on simulation. The measured wavenumber spectral shape can be matched at low ETG drive (consistent with the agreement in Qe), but not at high ETG drive. |
Friday, November 9, 2018 10:42AM - 10:54AM |
YO5.00007: Assessment of equilibrium field coil misalignments on the divertor footprints in NSTX-U Stefano Munaretto, Todd E Evans, Nathaniel M Ferraro, Dmitri M Orlov, Gregorio L Trevisan, Wen Wu A study of the heat load distribution at the divertor plates of NSTX-U due to misalignment of the equilibrium magnetic field coils indicates that a rigid shift in the installation of the toroidal field coils will produce the largest impact. Non-axisymmetric magnetic field perturbations due to error fields produce complex 3D edge magnetic topologies that alter the properties of the heat and particle flux distributions on the target plates. The perturbed equilibria due to shift and tilt of several coils are simulated using the MHD code M3D-C1. The produced footprints are estimated using the field line tracing code TRIP3D through the OMFIT framework. The effect the misalignment of several poloidal field coils separately has on the footprint is negligible for shifts up to 5mm and tilts up to 5mrad with the maximum current flowing in the coils.A rigid 5mm shift of the toroidal field coils relative to the centerline of NSTX-U will lead to a 10cm spread footprint on each of the outer divertor plates. Combination of shift and tilt of the toroidal field coils and estimation of the heat flux are also presented. These results will be used to determine the precision needed to install the equilibrium field coils in NSTX-U. |
Friday, November 9, 2018 10:54AM - 11:06AM |
YO5.00008: New aspects of error fields from high field side in tokamaks Jong-Kyu Park, Nathaniel M. Ferraro, Clayton E. Myers, Tomas Markovic, Matej Peterka, Stefan P. Gerhardt, Jonathan E. Menard, Radomir Panek, Alberto Loarte, Yuri Gribov Recent error field (EF) studies in NSTX-U and COMPASS are revealing new aspects of magnetic perturbations from the high field side (HFS) in tokamaks, challenging conventional EF correction strategies. First, attempts to correct EFs produced by a misaligned TF coil in NSTX-U were not fully successful, often driving a disruptive m/n=1/1 mode and almost always generating 2/1 locking in the edge. The optimal correction often varied dynamically depending on scenarios despite the statically generated nature of EFs. The controllable HFS EFs in COMPASS are correctable against 2/1 locking but not fully against other degrading effects such as a disruption during L-H transition, even with top and bottom (T/B) coils in addition to low field side (LFS) coils. Response simulations using IPEC, MARS, and M3D-C1 all indicate the complex m/n mode coupling and scenario dependences of HFS EF, as opposed to the single-mode characteristics of LFS EFC. NSTX-U recovery is accordingly preparing various EFC strategies with multi-level physics requirements including locking, NTV, and heat flux mitigation. COMPASS is testing various EFC scenarios for ITER using its unique coil sets, as will be discussed in detail. |
Friday, November 9, 2018 11:06AM - 11:18AM |
YO5.00009: Initial Operation of the ST40 Spherical Tokamak Paul Richard Thomas Advances in High Temperature Superconductor magnet technology allow a significant increase in the toroidal field (TF), which has been found to improve confinement in STs. The combination of the high b that has been demonstrated in STs and the high TF that can be produced by HTS TF magnets opens a path to lower-volume fusion reactors, since fusion power is proportional to b2 Bt4 V (V is the plasma volume). Tokamak Energy is aiming to exploit this concept as a route to fusion energy production. A high field ST, ST40 (R=0.4-0.6m, R/a=1.6-1.8, Ip=2MA, Bt=3T, κ=2.5, tpulse~1-10sec, 2MW NBI), was partially assembled and an experimental campaign conducted between January and June 2018. The main aim was to test Merging/Compression (MC) start-up. Plasma currents up to 350kA at 0.8T were obtained. Following MC, plasma was sustained for up to 15ms, even without a solenoid. Experiments on MAST and TS3 show efficient ion heating due to reconnection and this has been confirmed on ST40, with Ti > 1keV measured using Doppler broadening spectrometry. Plasmas with H-mode signatures were observed. ST40 is being moved to a larger facility, able to accommodate NBI and neutron shielding. The experimental results, an outline of the parallel HTS R&D and future plans will be presented. |
Friday, November 9, 2018 11:18AM - 11:30AM |
YO5.00010: Effects of Ohmic heating on the time scale of current quench during a disruption on EAST Debabrata Banerjee, Ping Zhu, Yanmin Duan, Songtao Mao The physical mechanism governing the process of thermal and current quench (TQ & CQ) during disruption has not been well understood. Performing resistive MHD simulation using NIMROD code, after choosing initial profiles from a recent EAST disruption discharge induced by massive helium gas injection, we compare experimental and simulation results, and find good agreement in the time history of total radiation power and MHD activity preceding thermal quench (TQ) and current quench (CQ) phases. It is demonstrated that the post-TQ re-increase of plasma temperature is due to the Ohmic heating (OH) power, which despite its imbalance with the radiation power loss, continuously feeds a lower temperature plasma during CQ phase. This produces a CQ time up to 8 times longer than the case without OH effect at all. Further simulations indicate that the effect of OH on CQ time scale is prominent in both natural and MGI induced disruption. |
Friday, November 9, 2018 11:30AM - 11:42AM |
YO5.00011: The hybrid scenario for high power, zero torque reactor operation Francesca Turco, Craig Petty, Thomas Petrie, Thomas Osborne, Tim Luce, Brian Grierson Operation at high density with off-axis EC power injection has been achieved in the high betaN hybrid scenario in DIII-D, expanding the parameter space to regimes compatible with a radiating divertor, for steady-state reactor relevant regimes. A variation on the high-betaN hybrid scenario has been found, that remains stable to the deleterious 2/1 mode without central CD deposition. Off axis EC heating leads to core impurity accumulation and lower confinement, but this effect is offset by the density increase, which leads to an enhanced pedestal and lower fast-ion losses. A second ingredient of a reactor-like scenario has been explored separately, obtaining passively stable hybrid plasmas at zero torque with and without ECCD. The lower torque discharges exhibit lower confinement. However, with reduced EC power (0.5 MW), tauE is found to increase (although H98y2 does not change significantly), such that the same betaN~2.3 can be reached with 2.3 MW less power. |
Friday, November 9, 2018 11:42AM - 11:54AM |
YO5.00012: Burn Control Mechanisms for Power Excursion Mitigation in ITER Maxwell D Hill, Weston M Stacey The strong increase of the D-T fusion cross-section with increasing plasma temperature in the ITER operating range has long been noted1-3 as a potential triggering mechanism for positive thermal power excursions, and this possibility remains an issue for ITER4-7. The efficacy of several inherent and active burn control mechanisms for limiting positive power excursions are explored. Mechanisms being evaluated include line and recombination radiation of seeded impurity ions5; MARFEs8; direct loss mechanisms such as ion orbit loss (IOL), ELMs, and impurity radiation enhancement by massive gas or pellet injection. We have made calculations to estimate the probable efficacy of such mechanisms in ITER. We find, e.g., that impurity seeding combined with increased massive D gas injection or pellets in the divertor or edge plasma may provide a plausible mechanism to radiate power to terminate a thermal power excursion. Refs: 1) Mills, LA4250, 1970; 2) Mori, Control Nuc Fus, 1105, 1976; 3) Stacey, NucFus 13,843,1973; 4) Borrass, Nuc Fus 31,1035,1991; 5) Mandrekas&Stacey, NucFus35,843,1995; 6) Humphreys, PhysPlas 22,021806,2015; 7) Boyer&Schuster, PPCF 56, 104004(2014); 8) Stacey, FusTech 52, 29 (2007).
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Friday, November 9, 2018 11:54AM - 12:06PM |
YO5.00013: Parameter dependences of runaway electron dynamics on ASDEX Upgrade and TCV Gergely Papp, Gabriella Pautasso, Joan Decker, Daniele Carnevale, Stefano Coda, Basil Duval, Ralph Dux, Ola Embreus, Boglarka Erdos, Ondrej Ficker, Rainer Fischer, Christoph Fuchs, Marco Gobbin, Linnea Hesslow, Mathias Hoppe, Andrej Lier, Patrick J McCarthy, Jan Mlynar, Alexander Mlynek, Gergo I Pokol, Umar Sheikh, George J Wilkie, the ASDEX Upgrade Team, the TCV Team, the EUROfusion MST1 Team The institutes collaborating within the EUROfusion consortium are executing a coordinated research program to better understand disruption-generated runaway electron (RE) generation, control and mitigation. The generation and subsequent suppression of REs on both ASDEX Upgrade (AUG) and TCV is achieved using massive gas injection (MGI) of neon or argon [PAPP IAEA 2016, Pautasso PPCF 2017]. |
Friday, November 9, 2018 12:06PM - 12:18PM |
YO5.00014: 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. © 2018 Lockheed Martin Corporation. All Rights Reserved. |
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