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 JO08: Magnetic Confinement: High Field TokamaksLive
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Chair: Brian Grierson, PPPL |
Tuesday, November 10, 2020 2:00PM - 2:12PM Live |
JO08.00001: Recent progress on the high-field path to commercial fusion energy Robert Mumgaard CFS and the MIT are developing the high-field path to commercial fusion energy, characterized by a focus on leveraging innovation, compact size, simplicity, and near-term deployment of fusion energy systems based on the high-field tokamak enabled by recent innovations in REBCO magnets. Key developments over the last year include full-scale REBCO TF magnet demonstration, sub-scale R{\&}D on REBCO CS and PF magnets, scale-up of the REBCO supply chain, and applications of the technology to non-tokamak fields. Extensive analysis has shown a robust physics and engineering basis for SPARC, a high field, net-energy tokamak, with access to reactor-relevant burning plasma physics.. The addition of an advanced divertor configuration enables SPARC to explore heat exhaust solutions at reactor-identical parameters. The final site selection for SPARC is anticipated in the fall and construction is expected to start in 2021. In parallel, work continues on ARC, the compact pilot plant proposed by MIT and CFS. Concept refinement has been steered by market insights from end-users and led to several different classes of plant that share the same physics and engineering base. In support of ARC, R{\&}D has started on liquid immersion blankets, molten salt, tritium extraction, and supporting subsystems with funding by several agencies including ARPA-E and non-US entities. CFS continues to encourage a fusion dialogue to non-government entities, partnerships, and fusion technology programs that prioritize climate-relevant commercial deployment. [Preview Abstract] |
Tuesday, November 10, 2020 2:12PM - 2:24PM Live |
JO08.00002: Overview of SPARC on the High-Field Path to Fusion Energy Dan Brunner The SPARC mission is to create and confine a plasma that produces net fusion energy for the first time. High-temperature, high-field superconductors are the fundamental technology that enables SPARC to be built at a relatively small scale compared to other proposed net-energy tokamaks; the smaller scale enables it to be completed on a faster timeline. The two major milestones of the 3-year Phase 1 of the project to be completed in June 2021 are (1) design, construction, and operation of a SPARC-relevant toroidal field model coil (TFMC) and (2) a ready-to-construct engineering design of the SPARC tokamak and facility. The second year of R$\&$D was very successful, the TFMC design is complete and the team has begun manufacturing of the coil and procurement of other components. In parallel, the physics and engineering design of the SPARC tokamak and facility has been baselined to a ``V2'', with no expected major changes through the remainder of the project. The physics performance projections remain robust to the Q$>$2 mission and the team has moved on to more detailed design of all of the subsystems. [Preview Abstract] |
Tuesday, November 10, 2020 2:24PM - 2:36PM Live |
JO08.00003: SPARC Integrated Design and Operational Scenarios Alexander Creely, Dan Brunner, Chris Chrobak, Darren Garnier, Robert Granetz, Martin Greenwald, James Irby, Brian LaBombard, Adam Kuang, Yijun Lin, Robert Mumgaard, Matthew Reinke, Ryan Sweeney, Steven Wukitch The SPARC tokamak is a critical next step toward commercial fusion energy. Having progressed through several design iterations, SPARC has now baselined major parameters at Version 2 (V2) and is proceeding with detailed engineering design. With $B_0 = 12.2$ T, $R_0 = 1.85$ m, $a = 0.57$ m, and a goal of $Q>2$, SPARC is a compact, high-field, D-T tokamak that will directly access conditions expected in a fusion power plant [A. J. Creely et al., \emph{J. Plasma Phys.} Submitted]. Empirical projections of plasma performance indicate that there is considerable margin over the $Q>2$ goal, predicting up to $Q \approx 11$ with $H_{98,y2}=1$. In addition to this primary reference discharge, several other plasma scenarios have been scoped out for SPARC, including L-mode operation, lower field and current scenarios, and an X-point target advanced divertor scenario. The engineering design has now incorporated detailed analysis of many key systems, including the toroidal and poloidal field coils, the central solenoid, the vacuum vessel, neutron shielding, and others, leading to the current machine design. [Preview Abstract] |
Tuesday, November 10, 2020 2:36PM - 2:48PM Live |
JO08.00004: Empirical and physics-based predictions of core plasma performance for the SPARC tokamak P. Rodriguez-Fernandez, N.T. Howard, M.J. Greenwald, J.W. Hughes, A.J. Creely, C. Holland, J.C. Wright, Y. Lin, F. Sciortino SPARC is being designed to be a medium-size tokamak (R$=$1.85m) that will leverage recent advancements in high-temperature superconductor technology to operate with a toroidal magnetic field of B$=$12.2T on axis. ICRF-heated deuterium-tritium H-mode plasmas in SPARC are expected to reach Q\textgreater 2 (core mission) and both empirical and physics-based models predict ample margin with respect to this mission. The empirical scaling of energy confinement (H98y2$=$1.0) and conservative assumptions estimate 140MW of fusion power and a gain of Q$=$11. Independent high-fidelity integrated modeling simulations with physics-based models for transport and heating are in remarkable agreement with empirical predictions and yield Q$=$9.0. This work presents predictions of several scenarios expected for the SPARC research program, and comprehensive scans of the parameter space with varying physics assumptions, demonstrating that the SPARC design is robust to uncertainties in plasma physics. The work that is being done in designing SPARC and its future operation will lay the basis for the high-field path and will serve as testbed for predictive burning-plasma physics models to be used in the design of high-field compact fusion power plants. [Preview Abstract] |
Tuesday, November 10, 2020 2:48PM - 3:00PM Live |
JO08.00005: Gyrokinetic simulations of turbulence and transport in the SPARC tokamak N.T. Howard, P. Rodriguez-Fernandez, C. Holland, M. Greenwald, J.W. Hughes Gyrokinetic simulations have been used to understand the turbulence and transport in the core of the SPARC tokamak. Building off of ongoing predictive TRANSP (1.5D) modeling that utilizes physics-based models (TGLF) to predict SPARC performance, we have performed linear and nonlinear gyrokinetic simulation across the SPARC profile using the CGYRO code. Simulations were performed with high physics fidelity: 6 gyrokinetic species, E{\&}M turbulence, Sugma collisions, realistic geometry, etc. Single-scale simulations were used to probe the turbulence at both ion (ky*rho\textunderscore s \textless 1.0) and electron-scales (ky*rho\textunderscore s \textgreater 1.0) to understand the origin of heat and bulk particle transport and to estimate the steady state impurity profiles expected in the SPARC baseline condition. The results of this analysis indicate that near marginal ITG turbulence and extremely stiff transport is expected in the core of SPARC. Ion-scale turbulence will likely play a dominant role with only small contributions from intermediate and high-k turbulence over most of the plasma profile, opening the possibility for tractable gyrokinetic profile prediction. High fidelity gyrokinetic simulation results will be discussed in detail and compared directly with reduced models to quantify differences in profile predictions and thus performance. Research supported by Commonwealth Fusion Systems. [Preview Abstract] |
Tuesday, November 10, 2020 3:00PM - 3:12PM Live |
JO08.00006: Dimensionless Parameter Scaling of Intrinsic Torque in C-Mod Enhanced Confinement Plasmas John Rice, Norman Cao, Tuomas Tala, Colin Chrystal, Martin Greenwald, Jerry Hughes, Jim Irby, Yijun Lin, Earl Marmar, Matt Reinke, Pablo Rodriguez-Fernandez A dimensionless parameter dependence study of intrinsic torque has been performed on a database of H- and I-mode plasmas from the Alcator C-Mod tokamak. The torque was determined by comparing intrinsic angular momentum density profiles just before and just after L-H and L-I transitions. The intrinsic torque has been found to scale as $\beta _{\mathrm{N}}^{\mathrm{1.5}}\rho_{\mathrm{\ast }}^{\mathrm{-1.0}}\nu_{\mathrm{\ast }}^{\mathrm{0.1}}$ , with the parameter ranges 0.3 \textless $\beta_{\mathrm{N}}$ \textless 1.5, 0.004 \textless $\rho_{\mathrm{\ast \thinspace }}$\textless 0.011 and 0.04 \textless $\nu_{\mathrm{\ast \thinspace }}$\textless 0.9, and with the intrinsic torque varying between 0.04 and 0.6 Nm. Comparison with results from other tokamaks suggests that the intrinsic torque should be normalized by some measure of the device size. Normalizing to the major radius yields a scaling proportional to 1/$\rho_{\mathrm{\ast }}$ at fixed $\beta _{\mathrm{N}}$ (around 1.3), and an intrinsic torque value of 4 Nm for SPARC with R $=$ 1.85 m and $\rho_{\mathrm{\ast }} \quad =$ 0.0027. [Preview Abstract] |
Tuesday, November 10, 2020 3:12PM - 3:24PM Live |
JO08.00007: Toward a tearing resilient SPARC equilibrium R. Sweeney, R.J. La Haye, C. Paz-Soldan, N.C. Logan, A.J. Creely, D.T. Garnier, R.S. Granetz, M. Greenwald, J. Irby, K. Montes, C. Rea, P. Rodriguez-Fernandez, R.A. Tinguely, J. Zhu Careful minimization of error fields (EF) and tailoring of the current profile to improve neoclassical tearing mode (NTM) stability are planned to avoid tearing mode islands. EF thresholds for penetration and locking scale unfavorably with the toroidal field and favorably with the electron density such that the predicted critical overlap field in SPARC is comparable to that in ITER. The dominant external EF exhibits the common low field side response, but the second singular mode is substantial and is sensitive to high field side errors, which is uncommon and attributed to the low beta. All equilibrium field coils, joints, and leads are assessed for as-designed errors. In addition, a workflow assesses as-built errors and provides engineering tolerances on equilibrium field coils, and an EF correction coil set is designed. NTM stability benefits from the high field, such that the marginal poloidal beta, below which tearing stability is expected, can be achieved with a reasonable value of the classical stability index $\Delta'$. A sensitivity study of $\Delta'$ varying about the design point is presented. [Preview Abstract] |
Tuesday, November 10, 2020 3:24PM - 3:36PM Live |
JO08.00008: Vertical Stability and Control on SPARC D.T. Garnier, Granetz R.S., J. Irby, C. Rea, R Sweeney, A.J. Creely For its highest performance discharges, SPARC is being designed with a strongly shaped equilibrium with high elongation and triangularity ($\kappa_{sep} = 1.97$ and $\delta_{sep} = 0.54$). While these parameters have been achieved with similar aspect ratio ($\epsilon = 0.31$) in the KSTAR and EAST tokamaks, the tendency for reduced vertical stability with increasing elongation is well known, and the importance of triangularity has more recently been explored[1]. As a high field, high current tokamak, SPARC will have a thick-walled, close-fit vacuum vessel and sections of toroidally continuous plasma first wall components which will provide passive stabilization with a time constant of $\sim30$ msec. In-vessel copper coils will be utilized for active vertical control. Scenario modeling has provided initial estimates of required passive elements and vertical stability coil specifications. Analysis of the passive stability parameters and expected maximum actively controllable z-distrurbance will be presented. [Preview Abstract] |
Tuesday, November 10, 2020 3:36PM - 3:48PM Live |
JO08.00009: Runaway electron mitigation in SPARC with a passively driven 3D coil Valerie Izzo, Darren Garnier, Robert Granetz, Ryan Sweeney To prevent growth of a large runaway electron (RE) population during the current quench (CQ) of a SPARC disruption, suppression of the RE population is required. Deconfinement due to field stochasticity could be achieved with large amplitude 3D magnetic perturbations produced by a 3D coil passively driven by the plasma current decay [1]. This scenario is modeled in NIMROD using prescribed 3D magnetic fields calculated for a proposed coil design, specified at the boundary of the simulation domain. The time dependence of the 3D field amplitude is evaluated as the simulation progresses, as a function of the total plasma current during the current decay, reaching the maximum coil current amplitude as the plasma current goes to zero. The simulations are initiated from a SPARC equilibrium and a thermal quench is induced by radiation from massive Ne deposition, after which the CQ ensues. During the simulation, drift-orbits for RE test particles are calculated to determine RE losses along stochastic fields. Initial calculations with 3D field perturbations exceeding $\delta $B/B$=$10$^{\mathrm{-2}}$ show complete loss of the RE test population within 2 ms of the start of the simulation. [1] A. Boozer, Plasma Phys. Control. Fusion 53 (2011) 084002 [Preview Abstract] |
Tuesday, November 10, 2020 3:48PM - 4:00PM Live |
JO08.00010: Ripple-induced fast-ion loss in SPARC due to misaligned TF Coils S. Scott, G. Kramer, E. Tolman, J. Wright, P. Rodriguez, A. Snicker, J. Varje, K. Sarkimaki The expected loss of fusion alpha power due to ripple-induced transport is computed for the SPARC tokamak design by the ASCOT and SPIRAL orbit-simulation codes to assess the expected surface heating of plasma-facing components. We find good agreement between the ASCOT and SPIRAL simulation results in integrated quantities such as fraction alpha power loss and also in the spatial, temporal, and pitch-angle dependence of the losses. The SPARC edge ripple is small (0.15 - 0.30{\%}) when the toroidal field (TF) coils are perfectly aligned, the associated computed ripple-induced alpha power loss is small (\textasciitilde 0.25{\%}), and the corresponding peak surface power density is acceptable (244 kW/m$^{\mathrm{2}})$. However, the ripple and ripple-induced losses increase strongly if the toroidal field coils are assumed to suffer increasing magnitudes of misalignment. Surface heat loads may become problematic if the TF coil misalignment approaches the centimeter level, which exceeds expected assembly tolerances. Recessing the plasma-facing surface of the RF antennas by one centimeter behind adjacent protective limiters is found to be sufficient to reduce the lost-alpha power load onto the antennas to safe levels. Ripple-induced losses of the energetic ion tail driven by ICRF heating are not expected to generate significant wall or limiter heating in the nominal SPARC plasma scenario. [Preview Abstract] |
Tuesday, November 10, 2020 4:00PM - 4:12PM Live |
JO08.00011: ICRF Heating for SPARC J.C. Wright, Y. Lin, S.J. Wukitch, P. Rodriguez-Fernandez, A.H. Seltzman SPARC is designed to have a 25 MW coupled ICRF system as its sole proposed auxiliary heating method. SPARC RF scenarios are based on the successes of Alcator C-Mod as well as the TFTR and JET programs during their D-T operation. Among heating methods, ICRF is the only proven method that can effectively heat high density and high field plasmas in SPARC for both the pre-D-T and D-T operations. The single-pass-absorption for D-T-($^3$He) burning plasma combining 2$^\textrm{nd}$ harmonic T heating and minority $^3$He heating will be greater than 90% during flat top operations. In addition to operation at 120 MHz and 12 T, we will discuss the possibility of 3-species scenario operation at a lower 8 T operation. Optimum fusion yield occurs for beam on target temperatures of 110 keV. This suggests tuning of ICRF scenarios to create tail energies for fuel ions in this range can be used to maximize the D-T burn rate. AORSA/CQL3D simulations will be discussed showing the generation of T and $^3$He tails and characterizing the general increase of the fusion rate in an ICRF heated plasma over the thermal D-T rate. The impact on Q as determined by TRANSP will be shown using effective Maxwellian for the T tail. [Preview Abstract] |
Tuesday, November 10, 2020 4:12PM - 4:24PM Live |
JO08.00012: Physics and engineering drivers for the SPARC divertor shape A.Q. Kuang, S. Ballinger, D. Garnier, M. Greenwald, J. Irby, B. LaBombard, J. Terry, J. Canik, T. Gray, J.D. Lore, M. Reinke, D. Brunner, A.J. Creely, B. Lipschultz, M. Umansky The SPARC divertor shape is being designed primarily for the baseline scenario of a strike point sweep and an advanced divertor mission. Based on empirical scalings, the peak unmitigated divertor parallel heat flux in SPARC is projected to be greater than 10 GW/m$^{\mathrm{2}}$ [1]. The current baseline scenario for operations employs a \textasciitilde 1 Hz strike point sweep to spread the heat flux and assumes 50{\%} of P$_{\mathrm{SOL}}$ is loss to radiation before reaching the target; the sweep frequency is limited by the central solenoid and poloidal field coil set, and the extent by the divertor surface. SPARC employs long (poloidally) tightly baffled divertors to maximize the swept area and reduces the incident total magnetic field line angle onto the target to \textasciitilde 1\textdegree . Furthermore, SPARC is being designed to enable the assessment of an X-point target outer divertor through optimization of coils and providing sufficient divertor volume to encompass the secondary X-point. While the X-point target divertor geometry is only achievable at reduced plasma current (5.7 MA), with an estimated PSOL$=$22 MW, SPARC still provides a reactor level testbed for the evaluation of this advanced divertor geometry. The design of the divertor is ongoing to ensure a consistent assembly and maintenance scheme is included. [1] Kuang, A.Q., \textit{J. Plasma Phys.} Submitted. [Preview Abstract] |
Tuesday, November 10, 2020 4:24PM - 4:36PM Live |
JO08.00013: Developing Divertor and Edge Modelling Studies for Advanced Configurations in the Divertor Tokamak Test Facility Michael Wigram, B. LaBombard, M. Greenwald, C. Meineri, P. Innocente, M. Moscheni, F. Subba The divertor heat flux problem is an important unresolved dilemma facing future reactor-level fusion devices. A variety of divertor configurations need to be examined to assess their suitability to meet this challenge. The Divertor Tokamak Test (DTT) facility, whose construction is starting, is an important step towards answering this question. Studies are underway to develop the DTT divertor and edge modelling, to compare predictions for various configurations. Initially a code benchmarking study is performed between three edge-plasma codes: SOLPS-ITER, UEDGE and SOLEDGE2D, to assess the differences in physics/predictions between the codes, including a model validation study for the three codes for high current, high-field, and narrow SOL width plasma shots using Alcator C-Mod data. Modelling studies will then be performed for single-null configurations of the DTT divertor, and compared with long-legged double-null configurations such as the Super-X and X-point Target divertors to compare their relative performance and to explore the potential performance benefits that these configurations may offer. This contribution will present the current progress and state of this research. [Preview Abstract] |
Tuesday, November 10, 2020 4:36PM - 4:48PM Live |
JO08.00014: Recent developments in the design of ARC Brandon Sorbom, Alex Creely, Shiaoching Tse, Dan Korsun The most recent developments in the design space for an ARC-class fusion power plant and plans for future technology development are presented. The original conceptual design of the ARC fusion power plant was presented in two papers published in 2015 and 2018. Although these papers presented one particular instantiation of ARC, this was not intended to be a ``final'' design but rather be a point within the spectrum of potential compact, high-field tokamak based power plants. Since the publication of the original papers, there has been development into both the physics and technology underlying compact, high-field tokamaks, as evidenced in numerous recent SPARC publications and hardware development being carried out by CFS and MIT. The same modeling and analysis toolset that has been used to inform the plasma physics basis for SPARC has been applied to the ARC design concept in order to more fully understand and define the parameter space in which ARC devices could exist. This includes first principles modeling and zero dimensional scaling. At the same time, technology development, specifically in HTS magnets over the past few years has progressed. HTS cable architectures have been developed for both AC and DC coil operation and a prototypical TF coil is currently being built and will be tested soon. This work in both physics and technology has provided a clearer picture of the feasibility of different development pathways to an ARC-like power plant. [Preview Abstract] |
Tuesday, November 10, 2020 4:48PM - 5:00PM Live |
JO08.00015: Maximizing fusion power in an ARC-class tokamak with a heat exhaust solution D.G. Whyte, A. Creely, P. Rodriguez-Fernandez, M. Greenwald Maximizing power is a key goal for fusion as an economic energy source. Fusion power fundamentally has two limits: core plasma pressure must reside within tokamak operation limits and plasma heat exhaust must be within technology limits. A compact high-field ARC-sized tokamak operating in L-mode with a radiative mantle offers a self-consistent solution to achieving over 2 GW of fusion power within heat exhaust limits with an ignited plasma. This performance trajectory requires low but acceptable safety factor, inductive current drive, high Greenwald fraction and seeded high-Z impurities. A 1-D power balance model shows pathways to achieving robust ignition and fusion power control with minimal external heating under 10 MW through the Cordey pass. The divertor is highly dissipative and used for particle/ash removal, but is not the principal heat exhaust location. The physics and technology implications of this approach are discussed. [Preview Abstract] |
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