Bulletin of the American Physical Society
64th Annual Meeting of the APS Division of Plasma Physics
Volume 67, Number 15
Monday–Friday, October 17–21, 2022; Spokane, Washington
Session YO03: Research in Support of ITER Burning Plasma PhysicsLive Streamed
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Chair: Florian M. Laggner, North Carolina State University Room: Ballroom 100 C |
Friday, October 21, 2022 9:30AM - 9:42AM |
YO03.00001: Evolution of Density and Temperature Full Profiles after Pure Ne and D2 Shattered Pellet Injections on DIII-D Andrey Lvovskiy, Nicholas W Eidietis, Thomas B O'Gorman, Daisuke Shiraki, Akinobu Matsuyama, Eric M Hollmann, Jeffery L Herfindal, Michael LEHNEN, Rejean Boivin We report detailed measurements of the electron density (ne) and temperature (Te) full profile evolution during pure neon (Ne) and deuterium (D2) shattered pellet injection (SPI) shutdowns on DIII-D. In order to mitigate thermal and electromagnetic loads and suppress runaway electron generation in ITER it is planned to perform massive injection of high-Z and low-Z species respectively. To study temporal and spatial dynamics of such injections, pure Ne and D2 were injected in the form of shattered pellets into DIII-D plasmas. The evolution of ne and Te profiles during the thermal (TQ) and current quench was measured using the Thomson scattering system recently upgraded with new capabilities to obtain low-Te measurements as well as to trigger based on a pre-defined level of ablation light. It was found that after Ne SPI the ne peak is located outside the q=2 flux surface and significant impurity mixing with the core plasma is predominantly observed only during and after the TQ when MHD activity is predicted to increase the radial transport. Similar accumulation of the impurity at the plasma edge is observed after D2 SPI, however, low mixing with the core plasma is seen even immediately after the TQ in the most of cases: the ne peak remains overwhelmingly in the outer half of the plasma radius. Initial modeling shows that this reduced mixing is due to the drift of a plasmoid formed around the ablation fragments. |
Friday, October 21, 2022 9:42AM - 9:54AM |
YO03.00002: The Electromagnetic Particle Injector as a Fast Time Response Disruption Mitigation System Roger Raman, Robert A Lunsford, Cesar Clauser Predicting and controlling disruptions is an important and urgent issue for ITER. Some disruptions with a short warning time may be unavoidable. For these cases, a fast time response disruption mitigation method is essential. A novel rapid time-response disruption mitigation system (DMS), referred to as the Electromagnetic Particle Injector (EPI), is being developed for tokamak-based reactors and as a backup option for ITER. Experimental tests on a prototype system have been able to verify the primary advantages of the concept. These are its ability to meet short warning time scales of <10 ms while attaining the projected high velocities for deep radiative payload penetration in reactor-scale plasmas. The EPI relies on an electromagnetic propulsion system. A metallic sabot is accelerated electromagnetically to the required velocities (> 1 km/s) within 2 ms, at which point it releases a radiative payload consisting of a shell pellet or well-defined microspheres. Initial experimental tests from the prototype system show attainment of over 600 m/s in about 1 ms. Essential aspects of payload separation from the sabot and sabot capture have also been demonstrated at 200 m/s, and the method can be extended to over 2 km/s. Results from the operation of the EPI-2 system that uses 3 T boost magnetic fields and EPI design projections to ITER will be described [Raman et al., Nuclear Fusion 61 (2021) 126034]. |
Friday, October 21, 2022 9:54AM - 10:06AM |
YO03.00003: Non-axisymmetric Radiation Modeling of JET SPI Discharges Using Emis3D Benjamin Stein-Lubrano, Ryan M Sweeney, Daniele Bonfiglio, Pedro Carvalho, Jack Lovell, Larry R BAYLOR, Robert S Granetz, Stefan Jachmich, Emmanuel Joffrin, Michael LEHNEN, Costanza F Maggi, Earl S Marmar, Eric Nardon, Umar A Sheikh, Daisuke Shiraki, Scott Silburn Precise radiated power values in tokamak disruptions are required to ensure stored energy in net energy tokamaks like ITER and SPARC is well mitigated. In mitigated disruptions, 3D MHD modes and localized impurity sources break toroidal symmetry. To incorporate asymmetric radiation structures, the Emis3D code adopts a physics motivated guess-and-check approach. Virtual radiation structures are observed with the Cherab modeling framework [M. Carr EPS 2017] and a best fit chosen using a reduced χ2 statistic. 2D axisymmetric inversions, helical structures, and 3D MHD simulated distributions from JOREK [Huysmans NF 2007] are tested. Two JET discharges terminated by shattered neon pellets are analyzed. 2D inversions are the best fit to the current quench, but helical structures are within uncertainty. Helicals fit the pre-thermal quench (pre-TQ) best and exhibit a parallel flow towards the high-field side. JOREK agreement to experiment in the pre-TQ is improved by increasing fidelity of the impurity source. Radiated fractions of 0.95 +0.05/-0.31 and 0.98 +0.02/-0.26 are found, suggesting that the stored energy may have been fully mitigated. |
Friday, October 21, 2022 10:06AM - 10:18AM |
YO03.00004: BOUT++ electromagnetic turbulence simulations of edge plasma dynamics during thermal quench Ben Zhu, Xueqiao Xu, Xianzhu Tang Damage of plasma facing components during tokamak disruption is a major concern for ITER and other future devices; and yet the underlying physics of disruption hasn’t been fully understood. In this study, BOUT++ six-field turbulence model with flux-driven capability is applied to investigate plasma turbulence and transport dynamics at the tokamak edge region, as well as the divertor power loads during the thermal quench phase of disruption. We find that with excessive particle and power are applied at the pedestal region for a short period of time (~10-20% of stored thermal energy within 0.1-1ms) to mimic the intensive particle and energy outflow from the core during the onset of thermal quench, two transport mechanisms - ExB edge turbulence and stochastic parallel diffusion, play important roles due to giant ELM-like instabilities as a result of rapid pedestal build-up. Enhanced edge turbulence is responsible for the surging divertor heat load and broadened heat flux at the early stage of thermal quench. At the late stage, turbulence induced magnetic fluctuation becomes large enough (>10−3) to completely break flux surfaces such that stochastic field-lines can directly connect pedestal top plasma to the divertor target plates or first wall, further impacting divertor heat load. |
Friday, October 21, 2022 10:18AM - 10:30AM |
YO03.00005: Zeff dependence of the L-H power threshold in ITER-similar-shape hydrogen and deuterium plasmas, and implications for reducing PLH in ITER Kyle Callahan, Lothar Schmitz, Troy Carter, Emily A Belli, Shaun R Haskey, Colin Chrystal, Brian A Grierson, Quinn Pratt, Christopher G Holland, Sterling P Smith, Kathreen E Thome, Gary M Staebler, George R McKee Review of a database of >500 DIII-D L-H transitions near ITER Pre-Fusion Power-Operation (PFPO-1) conditions (ITER-similar shape (ISS), low collisionality, near-balanced beam torque) has found a strong reduction of the L-H transition power threshold with increasing effective ion charge (Zeff) in both hydrogen and deuterium plasmas. This result suggests that up to 50% reduction of the L-H power threshold may be possible in ITER hydrogen plasmas by modest seeding of light impurities (Zeff~1.7). This trend was found after investigating different physics mechanisms (electron non-adiabaticity, ExB shear, collisionality) that can contribute to increasing thermal transport with smaller isotope mass [1]. Comparison of thermal fluxes to predictions from TGLF quasi-linear gyro-fluid simulations and CGYRO gyro fluid modeling in hydrogen and deuterium indicate the difference in heat flux is attributed to the different carbon content, with increased Zeff in D stabilizing ITG/TEM turbulence by increasing the critical ion temperature gradient [2,3] via main ion dilution and Landau damping by impurities. |
Friday, October 21, 2022 10:30AM - 10:42AM |
YO03.00006: Impact of divertor closure and X‑point height on the L‑H transition in DIII‑D Zheng Yan, George R McKee, Lothar Schmitz, Shaun R Haskey, John L Watkins, Huiqian Wang, Punit Gohil, Robert S Wilcox, Yasmin Andrew, Eun-Jin Kim The L‑H transition power threshold (PLH) in a partially closed divertor and with the small angle slot divertor (SAS) on DIII‑D is found to be about half of the PLH in open divertor plasmas at densities below 2×1013 cm‑3, and 30% lower at higher densities with otherwise similar parameters. PLH also reduces as the X‑point moves closer to the target plate, i.e. at shorter divertor leg length (DLL). Comprehensive 2D turbulence and flow measurements in the plasma edge reveal that turbulence characteristics, flow dynamics and EΧB drift near divertor are closely related to the PLH dependencies. As the DLL is reduced, PLH is observed to reduce by another 30%, while turbulence amplitude correspondingly increases, potentially providing a higher Reynolds stress drive for shear flow. The turbulence poloidal advection velocity from BES measurements increases across the separatrix into the SOL prior to the transition with shorter DLL. Ion saturation current (Jsat) measured by Langmuir probes inside the divertor region shows a significant reduction when the DLL is reduced, suggesting that EΧB drifts near the divertor region may play an important role. |
Friday, October 21, 2022 10:42AM - 10:54AM |
YO03.00007: Tungsten erosion and leakage in the new DIII-D V-shaped small angle slot divertor Gregory Sinclair, Tyler Abrams, Matthew S Parsons, Roberto Maurizio, Xinxing Ma, Dan M Thomas, John D Elder, Seth H Messer, Shawn A Zamperini, Jake H Nichols, Peter C Stangeby Validated impurity sourcing and transport simulations of the new tungsten-clad, V-shaped Small Angle Slot (SAS-VW) divertor in the DIII-D tokamak indicate the efficacy of closed, slot-like divertors in containing eroded high-Z impurities. Recent experiments measured a gross tungsten (W) erosion rate halfway between the entrance and bottom of the slot (known as the progressive angle, PA) of 1.3 × 1020 m-2 s-1 (Te ~16 eV, ne ~2.5 × 1019 m-3) with the outer strike point (OSP) at the PA, which decreased by a factor of 3.6 (30% drop in Te) when the OSP was moved to the slot bottom (ion B × ▽B drift towards the divertor). The observed trend is consistent with DIVIMP impurity transport simulations performed using a SOLPS-ITER plasma background (including particle drift effects), where the average W gross erosion rate calculated at the PA is 8.2 × 1018 m-2 s-1 (Te ~21 eV, ne ~1.5 × 1019 m-3) when the OSP is on the PA, and decreases by a factor of 54 (71% drop in Te) when the OSP is moved to the slot bottom. The probability that eroded W leaks into the core is predicted to decrease if sourced closer to the slot bottom, due to an increase in the parallel friction force. Benchmarked modeling of W transport provides valuable guidance for the design of reactor-scale, dissipative divertors. |
Friday, October 21, 2022 10:54AM - 11:06AM |
YO03.00008: Heuristic predictions of RMP configurations for ELM suppression in KSTAR and ITER, and their implications for divertor parameters Heinke G Frerichs, Jonathan M Van Blarcum, Oliver Schmitz, Jong-Kyu Park, SeongMoo Yang, Li Li, Yueqiang Liu, Alberto Loarte, Richard Pitts The three row setup of the in-vessel control coils (IVCC) in KSTAR offers many degrees of freedom for tailoring resonant magnetic perturbations (RMPs) for suppression of edge localized modes (ELMs), and ideal MHD based predictions of RMP configurations for ELM suppression based on empirical thresholds have been validated experimentally [J.-K. Park et al., Nature Physics 14, 1223 (2018)]. It is found that the heuristic ELM suppression window of these predictions can be reproduced based on kink amplitudes from drift-kinetic MHD plasma response (GPEC) calculations. For ITER, the ELM suppression threshold is taken from recent non-linear MHD (JOREK) simulations [M. Becoulet et al., Nucl. Fusion 62 (2022) 066022], and it is shown that edge stability proxies based on the kink amplitude and on the X-point displacement from linear resistive MHD plasma response (MARS-F) yield similar windows in RMP configuration space. |
Friday, October 21, 2022 11:06AM - 11:18AM |
YO03.00009: Impact of Resonant Magnetic Perturbations Scenarios on Detachment for ITER Jonathan M Van Blarcum, Heinke G Frerichs, Oliver Schmitz, Alberto Loarte, Richard Pitts, SeongMoo Yang, Jong-Kyu Park, Yueqiang Liu, Li Li A scan of gas injection into the scrape-off layer (SOL) conducted with EMC3-EIRENE for different resonant magnetic perturbation (RMP) scenarios shows how differences in the edge plasma magnetic geometry impacts detachment in ITER’s pre-fusion power operation (PFPO) configuration. ITER will suppress edge localized modes (ELM) by applying RMPs to provoke a response in the plasma edge. The distorted edge alters the flux to the divertor surface, raising concerns that the non-axisymmetric plasma surface interaction may inhibit detachment or induce intolerable localized heat flux. A previous study of how RMPs impact detachment for ITER PFPO showed that the region near the main strike point detaches more promptly in the density scan, but that the far SOL region remains attached [H. Frerichs et al., Nuclear Fusion (2021)]. This study compares n3, n4, and hybrid toroidal mode RMPs with phasing for optimized X-point displacement, as an ELM suppression metric. Detachment at the main strike point occurs most promptly for the largest magnetic footprint scenario but sustains high heat flux in the far SOL. These trends between the magnetic footprint and heat flux facilitate the selection of RMP configurations in an ELM suppressing subspace to optimize heat and particle loads on the divertor. |
Friday, October 21, 2022 11:18AM - 11:30AM |
YO03.00010: Progress on the ITER Toroidal Interferometer and Polarimeter (TIP) Michael Van Zeeland, Tsuyoshi Akiyama, Thomas N Carlstrom, Anthony Gattuso, Michael LeSher, Ryan Finden, Peter Trost, Daniel Finkenthal, Sebastian Miranda, Joseph Vincent, Marina Becoulet, Marc-Andre De Looz, Charlson C Kim, Christopher Watts The ITER Toroidal Interferometer and Polarimeter (TIP) is the primary diagnostic used for real-time feedback control of plasma density. In TIP, two-color vibration compensated interferometry is carried out at 10.59 μm and 4.6 μm using a CO2 and Quantum Cascade Laser respectively while a separate polarimetry measurement of the plasma-induced Faraday effect is made at 10.59 μm. Following tests of a full-scale TIP prototype on the DIII-D tokamak, which largely validated the measurement approach, work continues in advancing the engineering design and experimentally mitigating risks identified throughout the course of testing. Issues under investigation include atmospheric and window effects, beam refraction, and feedback alignment bandwidth. The TIP prototype has been upgraded with environmentally controlled "plasma" and "reference" leg chambers where humidity, pressure, temperature and species can be varied and the impact on measurements quantified. Numerical modeling suggests refraction is negligible for all but the highest density disruption mitigation scenarios; however, atmospheric and window effects, if not mitigated, will be issues and dominate TIP errors. |
Friday, October 21, 2022 11:30AM - 11:42AM |
YO03.00011: Optimizing the Microwave Performance of the ITER ECH Transmission Line System Michael C Kaufman, Gregory R Hanson The Electron Cyclotron Heating and Current Drive (ECH) system performs a variety of functions for ITER including plasma startup, instability suppression, current drive, and heating. The system consists of gyrotrons to produce the microwave power, launchers to inject it into the plasma, and transmission lines with their various components to transmit that power from the gyrotron to the launcher. |
Friday, October 21, 2022 11:42AM - 11:54AM |
YO03.00012: Radiation Shielding for ECH Launchers in ITER Michael J Loughlin Radiation transport analyses were carried out for the ITER electron cyclotron heating launchers. Elevated dose rates were predicted to the south of the tokamak due to activated water in the tokamak cooling system. These are among the most challenging radiation transport calculations conducted at ITER: they required unprecedented methods to accelerate the calculations as well as code development for the handling of very large models and the visualization of the results. Long (~25m) streaming paths defined by waveguides with a radius of less than 8 cm that cross several walls of the port cells and through the south wall into the assembly hall create a highly collimated beam of photons. |
Friday, October 21, 2022 11:54AM - 12:06PM |
YO03.00013: Investigating Divertor-safe Burning-Plasma Regimes in ITER using a Core-edge Model with SOLPS Scalings Vincent R Graber, Eugenio Schuster To maximize fusion power, ITER will need to access burn regimes that push against operational limits. These limits include saturation of actuators such as neutral beam injectors and pellet injectors. Furthermore, the heat load on the divertor target should remain below 10 MW/m2 to avoid melting. Using Plasma Operation Contour (POPCON) plots based on a model coupling the plasma’s core with the scrape-off-layer and divertor regions, the temperature-density space of ITER plasmas is investigated in this work to determine which operational limits are most restrictive towards achieving higher fusion powers. In this core-edge model, the core-plasma region is governed by nonlinear density and energy response models, while the edge-plasma regions are modeled with scalings. These scalings were generated from SOLPS4.3 simulation results [1] and depend on the core-plasma's power and particle outflows, the separatrix impurity concentration, and the gas injection rate. Outputs of the scalings such as the neutral influx, the separatrix temperatures, and the separatrix densities enter into the core-plasma response models. In addition, the scalings yield the target's heat load which is crucial to identify the high-fusion-power regimes compatible with safe divertor operation. |
Friday, October 21, 2022 12:06PM - 12:18PM |
YO03.00014: Nonlinear Alfvén instability analysis for ITER regimes Donald A Spong, Yashika Ghai, Jacobo Varela Rodriguez, Luis Garcia ITER will introduce new energetic particle (EP) physics regimes that have not been encountered of existing fusion experiments. The main parameter changes are in the EP gyro radius to machine size ratio and in the Alfvén Mach number (ratio of EP velocity to Alfvén speed); also, multiple fast ion sources and EP species will be present. This leads to dense, closely spaced Alfvén gap structures that lead to novel nonlinear coupling dynamics unlike those in current devices. For example, zonal flow generation/relaxation and successive EP profile flattening/avalanching can enhance EP transport and drive intermittency. To model these effects, large multi-mode simulations need to be followed in the saturated state for timescales long enough to capture the relevant dynamics. The nonlinear FAR3d gyro-Landau closure model has been developed for these goals and has been adapted to several ITER cases both in the pre-fusion power and the full fusion power regimes. The hybrid (MPI/OpenMP) parallelism of this model has allowed global nonlinear simulations including up to 20 coupled toroidal mode number families over a range of helicities (m/n) that cover ~90% of the minor radius. Features of EP driven turbulent transport in ITER’s fusion power regime will be described. |
Friday, October 21, 2022 12:18PM - 12:30PM |
YO03.00015: Measurements of Alfvén Eigenmode stability in the 2021 JET DT campaign Roy A Tinguely, Miklos Porkolab, Paulo Puglia, Nicolas Fil, Stuart Dowson, Michael Fitzgerald, David Keeling, James Oliver, Sergei Sharapov, Ziga Stancar, Remi Dumont, Jeronimo Garcia, Yevgen Kazakov, Phillip J Bonofiglo, Mario L Podesta, Ambrogio F Fasoli We present measurements of Alfvén Eigenmode (AE) stability from the 2021 JET DT campaign. The AE Active Diagnostic resonates with stable AEs and measures total damping rates γ < 0, eigenfrequencies, and toroidal mode numbers [Fasoli 1995 PRL]. Of 2000+ datapoints in 200+ DT discharges, we highlight measurements that can be compared to reference D or T plasmas: stable edge EAEs during modulations of the NBI power and alpha population; simultaneous un/stable AEs in the search for nonlinear coupling of fast ions, turbulence, and AEs; and first-ever stability measurements during JET energetic particle afterglow experiments. We use the kinetic-MHD code NOVA-K to identify eigenmodes, assess stability, and evaluate the contribution, if any, from alphas. |
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