Bulletin of the American Physical Society
63rd Annual Meeting of the APS Division of Plasma Physics
Volume 66, Number 13
Monday–Friday, November 8–12, 2021; Pittsburgh, PA
Session JO07: MFE: High-Field TokamaksOn Demand
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Chair: Saskia Mordijck, College of William and Mary Room: Rooms 315-316 |
Tuesday, November 9, 2021 2:00PM - 2:12PM |
JO07.00001: 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 scheduled to be completed in the summer of 2021 were (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 third year of R&D was very successful with the TFMC, at the time of abstract submission, being assembled and the test facility commissioned. In parallel, the physics and engineering design of the SPARC tokamak and facility has matured to the 30% complete "ready-to-construct" design, culminating in a week-long project review. No showstoppers were found and the go-ahead was given to start construction on the site in Devens, Massachusetts. The project remains on schedule for first plasma in 2025. |
Tuesday, November 9, 2021 2:12PM - 2:24PM |
JO07.00002: SPARC as a Path to ARC and Commercial Fusion Power Alexander J Creely, Dan Brunner, Martin J Greenwald, Adam Q Kuang, Robert T Mumgaard, Matthew L Reinke, Pablo Rodriguez-Fernandez, Brandon N Sorbom, Dennis G Whyte This talk describes the current concept for the ARC pilot fusion power plant, compares parameters in various SPARC plasmas to anticipated ARC plasma parameters, and lays out the timeline in which SPARC plans to achieve milestones as handoffs to ARC design. SPARC is being designed to achieve Q > 2 in order to demonstrate net energy production in a fusion device, and given nominal physics assumptions will have the capability of exploring regimes with Q ≈ 11 [A. J. Creely et al., J. Plasma Phys. 86 (5), 865860502 (2020)]. While these results will be an achievement in their own right, the true aim of SPARC is to validate the physics and technology necessary to build a commercial fusion power plant, currently embodied in the ARC design. In addition to demonstrating high gain plasmas that are non-dimensionally similar to ARC plasmas, SPARC will also compare conventional and advanced divertor configurations at relevant heat and particles fluxes to inform ARC design. SPARC operations will also inform the plasma control and diagnostics for ARC. A rapid ramp up in SPARC capabilities enables the design and operation of ARC on a time frame consistent with combatting climate change and consistent with the goals laid out in recent FESAC and NASEM reports. |
Tuesday, November 9, 2021 2:24PM - 2:36PM |
JO07.00003: Physics exploration of scenarios towards breakeven and burning plasmas in the SPARC tokamak Pablo Rodriguez-Fernandez, Nathan T Howard, Alexander J Creely, Benjamin F Spector, Martin J Greenwald, Jerry W Hughes This talk will cover modeling and predictions of SPARC scenarios beyond the previously described, high fusion gain primary reference discharge (PRD). Most physics studies of SPARC scenarios to inform the engineering design of the machine have been focused on the PRD, as it is often the plasma scenario with the most demanding requirements on tokamak systems. The PRD scenario is predicted both empirically and with physics-based models to produce over 100MW of fusion power with a gain of Q=9-11 [1]. However, while tokamak operators and scientists build expertise towards this landmark discharge, a number of other plasma regimes will necessarily be explored and thus it becomes increasingly important to understand the parameter space that SPARC is predicted to operate at. This work will review the PRD and modeling assumptions used, and will present predictions and simulations of scenarios expected to be accessed in the path to breakeven (Q>1), SPARC’s core mission (Q>2), and the burning-plasma regime (Q>5). Parameters such as confinement regime, plasma shaping, ICRF input power, minority scheme, and average density will be examined to inform SPARC’s research program and determine the safest and fastest path towards net fusion energy. |
Tuesday, November 9, 2021 2:36PM - 2:48PM |
JO07.00004: Common physics basis for the SPARC and COMPASS-U 3D coil designs Nikolas C Logan, Ryan Sweeney, Tomas Markovic, Carlos Paz-Soldan, Jong-Kyu Park, SeongMoo Yang, Qiming Hu, Caoxiang Zhu Non-axisymmetric field coils have been designed to provide efficient error field compensation and suppress edge localized modes in two new high field tokamaks, SPARC and COMPASS-U. These designs utilize the Generalized Perturbed Equilibrium Code's (GPEC's) representation of the multimodal, non-axisymmetric plasma response to optimize the geometric coupling between 3D coil arrays and the desired core or edge plasma response. Error field compensation coils are designed to couple to the plasma-amplified kink that dominates the drive of core resonances. The maximum allowable error field is projected to these new machines using an empirical scaling supported by nonlinear MHD modeling. Asymmetric construction and assembly tolerances are then balanced against the corresponding kA-turns needed for compensation to levels below the allowable limit. The coupling to edge resonant magnetic perturbation (RMPs) is similarly optimized for edge localized mode (ELM) suppression, with kA-turn requirements set by thresholds projected using a combination of reduced theory and nonlinear modeling. Edge perturbations are further decoupled from the core response as much as possible using higher toroidal mode numbers as well as by careful tuning of the poloidal spectrum. These physics-driven coil designs thus provide confidence in our ability to realize previously demonstrated 3D field benefits in new high field reactor regimes. |
Tuesday, November 9, 2021 2:48PM - 3:00PM |
JO07.00005: Disruption physics driving the SPARC design Ryan Sweeney, Valeria Riccardo, Darren T Garnier, Robert S Granetz, Martin J Greenwald, Andrew Maris, Cristina Rea, Benjamin Stein-Lubrano, Roy A Tinguely, Jinxiang Zhu, Matthew L Reinke, Valerie Izzo SPARC is expected to produce up to 140 MW of fusion power with a low normalized plasma beta (βN=1.0), low Greenwald fraction (nG=0.37), and a moderate safety factor (q95=3.4), providing margin to disruption limits. The SPARC tokamak [1] is a high field compact fusion experiment to demonstrate Q>2 and is expected to reach Q=9-11 [1,2]. Disruption frequencies are derived from the C-Mod, JET, JT-60U, and DIII-D tokamaks, and realistic disruption prediction performances and learning rates are assumed to estimate mitigated disruption frequencies. Current quench duration distributions and halo current fractions are derived from the ITPA studies [3]. The maximum vertical force is bounded by the quadrupolar field and geometric considerations of the plasma and first wall. The sideways force is conservatively bounded by the highest prediction of independent theoretical models. Fluid simulations coupled with a kinetic solver show that runaway electron currents of many mega-Ampere are possible [4]. A novel passive non-axisymmetric coil to expel runaway seeds is under design and integrated simulations predict complete prevention of runaway beams [4]. Simultaneous massive gas injectors are planned for thermal and current quench mitigation. |
Tuesday, November 9, 2021 3:00PM - 3:12PM |
JO07.00006: Tungsten and Disruptions in SPARC Robert S Granetz, Benjamin Stein-Lubrano, Matthew L Reinke, Ryan Sweeney, Adam Q Kuang The divertor surfaces and first wall in SPARC will be bulk tungsten or tungsten heavy alloy. Based on empirical evidence from Alcator C-Mod, there is concern that this could result in a significant fraction of disruptions triggered by sudden injections of tiny amounts of tungsten from overheated tile corners/edges (UFO's). C-Mod's first wall and divertor was bulk molybdenum, a high-Z metal similar to tungsten. About 25% of disruptions on C-Mod occurred with no detectable change of any measured plasma parameters more than 2-3 ms before the thermal quench. This is not enough warning time for any existing mitigation system to be effective. Since SPARC, like C-Mod, will have high plasma thermal density, high current density, high strikepoint heat flux, and a high-Z metallic first wall, there is concern that it too could have a significant number of UFO-triggered, unmitigatable disruptions. The ITPA MHD group initiated a joint study of UFO-triggered disruptions in tokamaks with high-Z first walls, and one conclusion is that careful engineering design to eliminate gaps between tiles and leading edges, and to ensure mechanical alignment even after repeated disruptions, could greatly reduce the likelihood of UFO-triggered disruptions, even in machines with high strikepoint power flux. |
Tuesday, November 9, 2021 3:12PM - 3:24PM |
JO07.00007: Mitigating disruptions with MHD in a multiphase vacuum vessel technology Ethan E Peterson, Sara Ferry, Jeffrey P Freidberg, Joseph Jerkins, Myles Stapelberg, Dennis G Whyte A novel multiphase vacuum vessel technology is introduced which combines a liquid lead MHD force disperser with recent advances in silicon carbide (SiC) manufacturing to mitigate disruptions and provide a number of other reactor benefits over conventional RAFM steel vacuum vessels. The multi-layered silicon carbide vacuum vessel structure comprises a thin plasma facing layer, molten salt cooling channels, thick silicon carbide walls, and helical liquid lead channels. The SiC plasma facing layer provides a thin, low-Z material to facilitate heat transfer to the molten salt cooling channels while minimally impacting plasma performance. The helical liquid lead channels force the disruption induced currents to travel largely parallel to the magnetic field in the liquid lead rather than in the SiC structure since lead is much more conductive than SiC. This both reduces the magnitude of the forces and converts any shear stress in the lead into fluid motion and pressure gradients leading to purely compressive or tensile stresses in the thick SiC walls. Other benefits to reactor operations including the possibility of increased operating temperatures and thermal performance as well as drastically reduced radioactive waste are discussed. The modeling efforts to support the design and optimization of this vacuum vessel technology are presented as well. |
Tuesday, November 9, 2021 3:24PM - 3:36PM |
JO07.00008: A pulsed, radiative L-mode scenario for application in an ARC-class device Conor J Perks, Samuel Frank, Andrew O Nelson, Amelia Cavallaro, Adam Rutkowski, Tony Qian, Dennis G Whyte, Pablo Rodriguez-Fernandez, Allan Reiman A highly-radiative, pulsed, burning plasma L-mode regime is developed and evaluated as a candidate for the base scenario on future tokamak reactors. In this configuration, the fusion power density can be absolutely maximized before reaching material or administrative limits, allowing for the optimal gain from plasma physics in robust scenarios with 1500 MW of fusion power. To establish this operating scenario, high-temperature superconducting magnets are employed to maximize both plasma currents and plasma densities at moderate reactor scale. Pulsed operation alleviates the stringent current drive requirements of steady-state reactors, while high fusion power densities allow for economic benefits even with significant reactor downtime. Operation in L-mode affords access to an ELM-free highly-radiative mantle regime with ∼85% core radiation fractions, significantly reducing the power load requirements on plasma-facing components. Reactor conditions are calculated self-consistently using ACCOME, TGYRO, GENRAY, CQL3D, and a 1-D open-source POPCON code, demonstrating a robust operating point. |
Tuesday, November 9, 2021 3:36PM - 3:48PM |
JO07.00009: Investigations of tungsten heavy alloy (WHA) for use on SPARC Adam Q Kuang, Dina Yuryev, Travis K Gray, Matthew L Reinke, Trey Henderson, Petr Kolonin, Michael Lagieski, Ryan L Romesberg, Andrew Seltzman, Guy S Showers, Deepthi Tammana, Dennis Youchison, Douglas E Wolfe, Dan Brunner, Martin J Greenwald Tungsten based materials have been chosen for SPARC’s plasma facing components (PFCs). However, the brittle nature of pure tungsten makes it difficult to work with in a tokamak environment. This has motivated an evaluation of WHA for use in SPARC. WHA remains ductile even at room temperature and has been successfully deployed in a tokamak environment on ASDEX-Upgrade. However, there is a current gap in literature on the compression structural properties of WHA and the behavior of WHA when subjected to heat flux loading at magnitudes expected for SPARC transients. Previous tests have shown that the WHA, compared with pure tungsten, has a lower temperature failure mode due to the use of Ni and Fe binder materials. Testing indicates that WHA behaves in a ductile manner under compression loading. In addition, high heat flux testing carried out at the Applied Research Laboratory electron beam facility indicates a significant ejection of material when exposed to moderate heat flux levels (>100 MW/m2 for durations of <100 ms). Though, when exposed to higher heat flux levels (~500 MW/m2 for durations of <30 ms), surface melting in both WHA and pure tungsten is observed although with some distinguishing behavior. |
Tuesday, November 9, 2021 3:48PM - 4:00PM |
JO07.00010: Dependence of the Heat Flux Width on Average Pressure in Alcator C-Mod Sean B Ballinger, Dan Brunner, Amanda E Hubbard, Jerry W Hughes, Adam Q Kuang, Brian LaBombard, James L Terry, Anne E White The Alcator C-Mod heat flux width (λq) database suggests a strong correlation with the inverse square root of the volume-averaged plasma pressure, unified across L-, I-, and H-mode plasmas [1]. This implies that a high performance core at high pressure will lead to challenging heat and particle exhaust at small λq. This concern motivates our work, as does research from ASDEX Upgrade, where λq was found to be more strongly correlated with the edge electron pressure and its gradient length than with the average pressure [2]. This study adds density, temperature, and pressure profile data from the Thomson scattering and electron cyclotron emission diagnostics to the C-Mod database that was used to develop the pave-0.5 scaling. The profiles are fit using Gaussian process regression. Within the extended database, the average pressure still shows a stronger correlation with λq than any localized value of the profiles or gradient lengths. Pressure values in the region of ψnorm < 0.8 are found to be better correlated with λq than pressure values further out. Finally, scalings of λq with localized profile values in C-Mod are found to differ across confinement regimes. |
Tuesday, November 9, 2021 4:00PM - 4:12PM |
JO07.00011: Transport model validation in support of burning plasma physics Anne E White, Rachel Bielajew, William Burke, Xiang Chen, Garrard D Conway, Calvin Cummings, Nathan T Howard, Jerry W Hughes, Abhilash Mathews, Pedro A Molina Cabrera, Pablo Rodriguez-Fernandez, Aaron M Rosenthal, Francesco Sciortino, Christian Yoo Improving confidence in the predictions for performance in burning plasmas is of great importance for accelerating the development of fusion energy. State-of-the-art predictions of pressure profiles and rotation profiles in standard aspect ratio tokamaks like ITER and SPARC, as well as spherical tokamaks, are performed using physics-based, but reduced, models of turbulent transport. The details of the turbulence can be simulated in burning plasmas using high-fidelity nonlinear gyrokinetic simulation. Confidence in reduced models and in predictions from nonlinear gyrokinetic simulation relies on rigorous validation against experimental measurements. Validation studies carried out on major tokamak facilities around the world are driven by steady investment aimed at advancing the frontiers of plasma diagnostics, data analysis techniques, plasma theory, and computational science. This talk will summarize research carried out at the MIT PSFC on transport model validation in support of burning plasma physics, including new diagnostic development, fluctuation measurement databases, and integrated modeling approaches that leverage AI and machine learning. O. Meneghini et al. 2021 Nucl. Fusion 61 026006; C. Chrystal et al 2020 Nucl. Fusion 60 036003; P. Rodriguez-Fernandez et al., J. Plasma Phys. 86, 865860503 (2020); B. Patel, et al. 46th EPS Conference on Plasma Physics, EPS 2019; G. Staebler et al. 2017. Nucl. Fusion 57 (6), 066046; N. Howard et al. Physics of Plasmas 28, 072502 (2021). |
Tuesday, November 9, 2021 4:12PM - 4:24PM |
JO07.00012: Design of a flexible divertor for the DTT facility Paolo Innocente, Roberto Ambrosino, Flavio Crisanti, Selanna Roccella The standard Single Null Divertor (SND) configuration with tungsten monoblocks targets can face some difficulties in providing a solution scalable towards the realization of the fusion reactor based on H-mode tokamak configuration. To provide a safer solution the adoption of alternative divertor magnetic configurations (ADCs) or liquid metallic targets have been considered to mitigate the stationary heat load on targets. Additionally different plasma scenarios have been considered to avoid the huge transient energy release due to the type-I ELMs of the high confinement H-mode tokamak configuration. |
Tuesday, November 9, 2021 4:24PM - 4:36PM |
JO07.00013: Application of edge codes to the new high-field tokamaks Michael R Wigram, Martin J Greenwald, Brian LaBombard, Carlo Meineri, Matteo Moscheni, Fabio Subba, Hoasheng Wu, Paolo Innocente, Claudio Carati A new generation of high-field tokamaks are being designed for the near future. Such devices present potential benefits but also challenges - narrow heat flux widths associated with high poloidal fields make the divertor heat flux problem particularly difficult. The Divertor Tokamak Test (DTT) facility – a high-field device compared to current tokamaks, reaching 6T on-axis – will be an important step towards finding a solution. Simulation codes provide useful tools to design and assess divertor scenarios, but will encounter their own challenges in high-field devices, with narrow particle and energy channels and steep gradients. In this work, to improve the DTT divertor modelling capability, a comparison study is performed between three edge-plasma codes: SOLPS-ITER, UEDGE and SOLEDGE2D, to assess and understand the differences in code predictions, including a model validation study for a high-field, narrow SOL width discharge in Alcator C-Mod. Modelling will then move on to examine advanced long-leg configurations. Results find limitations in fluid neutral models opposed to kinetic solvers, the importance of charge-exchange models for molecular hydrogen, as well as mesh extension effects – all showing notable impact on predictions in the high-field, high fusion power context. |
Tuesday, November 9, 2021 4:36PM - 4:48PM |
JO07.00014: Comparisons of neutral density profile predictions using SOLPS-ITER and KN1D with experimental neutral profiles in Alcator C-Mod Richard M Reksoatmodjo, Francesco Sciortino, Saskia Mordijck, Jerry W Hughes, Jeremy D Lore, Xavier Bonnin, Matthew L Reinke
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Tuesday, November 9, 2021 4:48PM - 5:00PM |
JO07.00015: An integrated design study for the EXhaust and Confinement Integration Tokamak Experiment David B Weisberg, Tyler Abrams, Jayson L Barr, Michael W Brookman, James Leuer, Brendan C Lyons, Ross MacDonald, Joseph Mcclenaghan, Orso M Meneghini, Craig C Petty, Robert I Pinsker, Gregory Sinclair, Wayne M Solomon, Tony S Taylor, Dan M Thomas, Bart G Van Compernolle, Michael A Van Zeeland, Will Wehner, Wen Wu, Jonathan H Yu, Kurt Zeller A high-level design study for a new experimental tokamak shows that advances in fusion science and engineering can be leveraged to close the remaining gaps between present devices and a future fusion pilot plant (FPP). This new US facility will access an operational space close to the projected FPP performance regime via a compact, high field, high power-density approach that utilizes advanced tokamak scenarios and high temperature superconductor magnets. Full-device optimization via system code calculations, physics-based core-edge modeling, plasma control simulations, and finite-element structural & thermal analysis has converged on a BT=6T, IP=5MA, R0=1.5m, A=3 D-D tokamak with strong plasma shaping, long-legged divertors, and 50MW of auxiliary power. Such a device will match several absolute FPP parameters: pressure, exhaust heat flux, and bootstrap fraction. It will also narrow or close the gap in key dimensionless parameters: toroidal beta, normalized gyroradius, collisionality, and edge neutral opacity. Integrated neutron shielding preserves personnel access by limiting nuclear activation and maximizes experimental run time by reducing site radiation. In addition to design study results and optimization details, parameter sensitivities and uncertainties are also discussed. |
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