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 NO08: MFE: Research in Support of ITEROn Demand
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Chair: Alex Creely, Commonwealth Fusion Systems Room: Rooms 317-318 |
Wednesday, November 10, 2021 9:30AM - 9:42AM |
NO08.00001: Status of the US contributions to the ITER Project Kathryn McCarthy As the international ITER project prepares for tokamak assembly, US ITER continues to deliver, fabricate and design essential ITER hardware systems. As of April 2021, planned US hardware deliveries are 40% complete, with two systems already fully delivered. In 2021, a major milestone was achieved: the successful completion of testing plus shipment of two central solenoid magnet modules to the ITER site. The central solenoid is the only ITER magnet to undergo post-production testing that simulates the ITER operational environment, with cryogenic temperatures of 4.5 degrees Kelvin and powering to 40,000 amperes. Disruption mitigation testing continued on JET tokamak, using US ITER and ORNL-developed hardware. Looking forward to 2022, US ITER is continuing to complete essential deliveries for the central solenoid support structures, tokamak cooling water system, and vacuum system, and is advancing post-first plasma design to prepare for fabrication and deliveries that are needed after initial operations. ORNL, which manages US ITER for DOE Office of Science, is also engaging with partners across the US to prepare for ITER science operations. |
Wednesday, November 10, 2021 9:42AM - 9:54AM |
NO08.00002: Correlation of the scrape-off layer (SOL) power width with edge electron pressure gradients at ASDEX Upgrade Davide Silvagni, Thomas Eich, Michael Faitsch, Tim Happel, Bernhard Sieglin, Pierre David, Dirk Nille, Luis Gil, Ulrich Stroth Understanding how the scrape-off layer (SOL) power decay length λq scales with plasma parameters is essential for designing fusion reactors and developing a power exhaust solution. Previously, several scaling laws have been proposed to predict λq values in ITER [1, 2]. They are based on engineering or global plasma parameters, which on the one hand make the λq prediction easier, but on the other hand could hide the underlying physics mechanism setting λq. To shed additional light on this, a cross-regime (L-mode, I-mode and H-mode) database combining λq measurements from infrared cameras and upstream SOL and pedestal gradient lengths has been assembled at ASDEX Upgrade. It is found that λq exhibits a clear correlation with both SOL and pedestal electron pressure gradient lengths, which in turn are remarkably well correlated to each other. Consequently, as the pedestal width is mainly constant in the analyzed discharges, a cross-regime λq scaling is well described by the edge electron pressure evaluated at ρpol = 0.95. |
Wednesday, November 10, 2021 9:54AM - 10:06AM |
NO08.00003: Impurity leakage and radiative cooling in the first nitrogen and neon seeding study in the slot divertor at DIII-D Livia Casali, David Eldon, Thomas Osborne, Anthony W Leonard, Adam G McLean, Brian A Grierson Impurity seeding experiments were performed in the SAS divertor at DIII-D using N and Ne as radiative species showing a strong dependence of leakage behavior on impurity species [1]. The experimental results indicate that Ne radiates more upstream than N reducing compression in the divertor while increasing the pedestal pressure profile. The different dissipative behavior between the two radiative species is confirmed by SOLPS-ITER modelling which for the first time at DIII-D includes multiple impurity species and a treatment of full drifts, SOL currents and n-n collisions. The impurity transport in the SOL is studied in terms of the parallel momentum balance showing that N is mostly retained in the divertor whereas Ne leaks out of the divertor consistent with its higher ionization potential and longer mean free path producing the lower divertor enrichment found for Ne. The strong ionization source and particle drifts characterizing the SAS contribute significantly in the shift of the stagnation point, thus affecting impurity leakage. The results demonstrate the impact of flow reversal and drifts in closed divertor structure with important consequences on SOL impurity transport and pedestal performance. [1] L. Casali et al PoP 2020. |
Wednesday, November 10, 2021 10:06AM - 10:18AM |
NO08.00004: Research in Support of the Development of a Plasma Pulse-Resolving, Fuel-Cycle Exhaust-Gas Analyzer for the ITER Divertor C.Christopher Klepper, Ephrem Delabie, Ionut Jepu, Chris Marcus, Georg Schlisio, Stephane Vartanian, Kurt G Vetter, Theodore M Biewer, Jeffrey H Harris, Uron Kruezi, Fabio A Ravelli The ITER Diagnostic Residual Gas Analyzer (DRGA) will measure the distribution of gas species, i.e., H, D, T, and impurities, in the divertor exhaust stream and in the plasma periphery. Recently restarted, final design activities are strongly benefiting from testing of prototypical DRGA components and methods on current fusion devices under international collaboration programs. Advances include recent demonstration of simultaneous H/D/T and 3He/4He absolute concentration measurements on JET with an upgraded, DRGA-like optical gas analyzer and with detection down to 0.1% levels [1]. Experience with this measurement is currently extending to T containing plasmas and has direct implications for operating ICRH in ITER PFPO-2 [2]. A US-ITER developed, prototype quadrupole mass spectrometer, using custom compensation circuitry to extend the cable length by over 10x the state-of-the-art, has also been successfully demonstrated in JET pre-DTE2 plasma operations [3]. The Molfow+ MC simulation code, applied to the divertor DRGA configuration, has confirmed previously estimated response times [4]. Comparison of this simulation to measurements from a prototype ITER DRGA analysis station on W7X, operated in the OP1.2b campaign, revealed a backflow issue for light gas species, which can limit fuel cycle gas isotopic detection limits and whose mitigation is now included in the ongoing final design plan for ITER. |
Wednesday, November 10, 2021 10:18AM - 10:30AM |
NO08.00005: Fluid turbulence simulations of ELM characteristics and divertor heat load for ITER scenarios using BOUT++ Xueqiao Xu, Xueyun Wang, Xiaoxue He, Zeyu Li, Philip B Snyder, Ben Zhu The BOUT++ code is used to simulate the ITER 11.5MA hybrid scenario and comparisons are made among ITER baseline, hybrid and steady-state operation (SSO) scenarios. Peeling-ballooning instabilities with different toroidal mode numbers dominate in different scenarios and consequently yield different types of ELMs. The energy loss fractions (DW/Wped) in the baseline and hybrid scenarios are large (~10-20%) while the one in the SSO scenario is dramatically smaller (~1%), which are consistent with the features of type-I ELMs and grassy ELMs. The divertor heat flux width in the 3 scenarios given by the simulations is larger than the estimations based on the HD model and Eich scaling. The toroidal gap edge melting limit of tungsten monoblocks of divertor targets imposes constraints on ELM energy loss, giving that the ELM energy loss fraction should be smaller than 0.4%, 1.0%, and 1.2% for ITER baseline, hybrid and SSO scenarios. The simulations show that only the SSO scenario with grassy ELMs may satisfy the constraint. From transport simulations, we obtain the critical diffusivity χc=0.5m2/s in 5MA/1.77T PFPO-1 scenario and χc=0.3m2/s in 7.5MA/2.65T PFPO-2 scenario for a transition from a drift to turbulence dominant regime. |
Wednesday, November 10, 2021 10:30AM - 10:42AM |
NO08.00006: Nonlinear two-fluid modeling of plasma response for the RMPs ELMs control in ITER baseline scenarios Qiming Hu, Jong-Kyu Park, Raffi M Nazikian, Nikolas C Logan, SeongMoo Yang, Brian A Grierson, Qingquan Yu The plasma response of resonant magnetic perturbations (RMPs) for controlling edge-localized modes (ELMs) in ITER baseline scenarios is investigated by combining the toroidal ideal MHD code GPEC and the nonlinear two-fluid MHD code TM1. A number of issues relative to RMP ELM control are investigated, including RMP coils configuration optimization, predicting the density pump-out and q95 windows of ELM suppression. GPEC 2D scans of the relative coil current phasing reveal the optimal phasing for n = 1 to 5. TM1 nonlinear simulations show RMP penetration at the pedestal-foot, which leads to density pump-out with its magnitude scaling as IRMP0.5, and it becomes weaker for higher-n RMP. The TM1 simulations also show field penetration at the pedestal top. Simulations by 2D scans of RMP coil current and q95 reveal the accessing q95 windows of ELM suppression for both n = 3 and 4. The predicted q95 windows of ELM suppression for ITER are very similar to the observations in recent tokamaks and the required RMP coil current is within the designed capability. In addition, the simulations indicate that wide q95 windows of ELM suppression may be accessible in ITER by operating n = 4 RMPs. |
Wednesday, November 10, 2021 10:42AM - 10:54AM |
NO08.00007: Optimal strategy for RF stabilization of NTMs in ITER Richard Nies, Allan Reiman, Nathaniel J Fisch Recent findings concerning the effect of the test blanket modules on NTM locking in ITER, as well as recent projections for the broadening of the RF deposition profile due to edge density fluctuations, call into question the planned strategy of RF stabilization of NTMs while the islands are rotating. We find that, given these recent projections, it becomes advantageous to opt instead for a locked mode stabilization strategy, where the mode is made to lock at small width in front of the RF wave launcher and can then be stabilized efficiently. This strategy is made possible by the recent recognition that the H-mode is preserved for a momentum confinement time after locking, and that RF stabilization on a shorter time scale than this preserves the H-mode. Our findings emerge from a model for the temporal evolution of the magnetic island width and rotation frequency that includes the effects of the bootstrap, RF-driven and polarisation currents, the viscous drag and the effects of both the error field and the resistive wall on the island's growth and rotation. These effects all play an important role in the evolution of magnetic islands, making their inclusion critical to evaluate and compare the locked mode and rotating mode stabilisation strategies. This alternative strategy requires no changes to ITER's design, and would allow for more robust stabilization of NTMs with lower peak and averaged power requirements. These would directly translate to increased fusion gain, more availability of RF power for other needs, and lower disruptivity. |
Wednesday, November 10, 2021 10:54AM - 11:06AM |
NO08.00008: Alfvén Eigenmode stability measurements in recent JET H, D, T, and DT plasmas Roy A Tinguely, Miklos Porkolab, Paulo Puglia, Ambrogio F Fasoli, Nicolas Fil, Stuart Dowson, Michael Fitzgerald, David Keeling, Sergei Sharapov, Remi Dumont, Javier Gonzalez Martin, Zhihong Lin, Yevgen Kazakov, Jef Ongena, Massimo Nocente, Mario L Podesta, Anna Teplukhina, Ziga Stancar We present Alfvén Eigenmode stability measurements (frequencies, damping rates γ<0, and toroidal mode numbers) from hundreds of H,D,T, and DT plasmas in the 2019-21 JET campaigns. Data are collected by eight in-vessel, toroidally spaced antennas [Panis NF 10] independently powered/phased [Puglia NF 16] to resonate with stable AEs [Fasoli PRL 95]. The system recently demonstrated improved performance during high-power heating [Tinguely NF 21a] and X-point configuration [Tinguely NF 21b] compared to the previous system's operation during early JET campaigns, including 1997 DT [Fasoli PoP 00]. Contributions to γ from various drive/damping mechanisms (especially from alphas) are assessed through MHD, kinetic, and gyrokinetic simulations, allowing extrapolation to future fusion devices. Importantly, knowledge gained from a range of isotope mixes, plus significant He fractions, can be transferred to pre-fusion and DD/DT ITER operations. |
Wednesday, November 10, 2021 11:06AM - 11:18AM |
NO08.00009: Dependence of the L-H power threshold on the alignment of externally applied non-axisymmetric MPs Matthias Willensdorfer, Ulrike Plank, Marco Cavedon, Garrard D Conway, David Ryan, Wolfgang Suttrop, Rico Buchholz, Mike Dunne, Michael Griener, Jörg Hobrik, Andrew Kirk, Rachael M McDermott, Thomas Pütterich, Giovanni Tardini, Rainer Fischer, Qingquan Yu, Dominik Brida, Sergei Kasilov A series of experiments at ASDEX Upgrade (AUG) has been conducted to study the impact of the alignment of external magnetic perturbations (MPs) with n = 2 toroidal mode symmetry on the power threshold of the transition from L-mode to H-mode (L-H transition). This is interesting for ITER, since its operation will rely on an MP-field configuration that is expected to solidly suppress edge localised modes (ELMs) while avoiding an increase in the L-H power threshold (PLH). |
Wednesday, November 10, 2021 11:18AM - 11:30AM |
NO08.00010: Thermal transport in hydrogen plasmas at high Te/Ti in DIII-D relevant to ITER PFPO-1 Brian A Grierson, Shaun R Haskey, Gary M Staebler, Arash Ashourvan, Terry L Rhodes, Francesca M Poli, George R McKee, Tom H Osborne A recent L-mode isotope experiment for ITER pre-fusion power operation achieved high Te/Ti in hydrogen plasmas for transport model validation and predictions for ITER PFPO-1. Using recent improvements to TGLF we show the ability to predict the Te, Ti and ne profiles approaching very close to the separatrix ρ~0.95, previously limited to ρ~0.8. The ITER research plan at ⅓ field will use 3rd harmonic electron cyclotron heating (ECH) that is sensitive to Te for absorption and rely on electron-ion coupling to access H-mode. DIII-D experiments with high hydrogen purity (~90%) have varied the plasma density to change Te/Ti through electron-ion exchange, and added up to 1.6 MW of central ECH achieving core Te/Ti~1-3.5 and the same power per area as ITER with 20-30 MW ECH. TRANSP and new H/D measurement using main-ion CER are used to determine ion composition and power flows. Simulations using a new version of TGLF that incorporates improved geometry, collisions and calibrated against CGYRO shows TGLF captures the core temperature profiles well and much closer to the plasma separatrix than the original saturation rule in TGLF due to higher stiffness in the new model. |
Wednesday, November 10, 2021 11:30AM - 11:42AM |
NO08.00011: Reduced L-H Transition Power Threshold in ITER-Similar-Shape DIII-D Hydrogen Plasmas* Lothar W Schmitz, K. Callahan, T. L Rhodes, L. Zeng, Z. Yan, G. R McKee, Y. Q Liu, T. Osborne, C. Chrystal, P. Gohil, B. C Lyons, C. C Petty, R. S Wilcox, D Shiraki, F. Laggner, S. R Haskey, B. A Grierson Recent DIII-D experiments in low-torque, ITER-similar-shape (ISS) hydrogen plasmas (q95 ~3.6) show that the L-H transition power threshold PLH can be reduced substantially (~ 25-30%) via moderate (≤ 20%) Helium admixtures. PLH has also been effectively reduced at low ion edge collisionality via applied n=3 Non-Resonant Magnetic Perturbations (NRMF), producing local edge counter-current torque via the Neoclassical Toroidal Viscosity (NTV). CER measurements confirm a significant increase in toroidal (Carbon) edge rotation shear with applied NTV before the L-H transition. Techniques for reducing PLH are very important for ITER, in particular for hydrogen plasma operations during the PFPO-1 campaign with marginal auxiliary heating (20-30 MW of ECH). Control of L-mode E×B shear via Helium seeding, or applied NRMF/NTV can open up a path for reducing PLH in burning plasma experiments. Linear plasma response simulations for ITER indicate that the 3-D internal coil set (with n=3) can be used to generate large edge NTV, favored by the low collisionality expected in the ITER L-mode edge. *This work was supported by the US Department of Energy under DE-SC0020287, DE-FG02-08ER54984, DE-AC05-00OR22725, DE-FG02-08ER54999, DE-FC02-04ER54698, DE-FG02-07ER54917, and DE-AC02-09CH11466. |
Wednesday, November 10, 2021 11:42AM - 11:54AM |
NO08.00012: Modeling the impact of disruption induced heat loads on ITER plasma-facing components Lei Chen, Richard Pitts, Michael Lehnen, Jonathan Coburn, Gregor Simic, Matic Brank, Svetlana Ratynskaia In ITER, transient heat loads during the current quench (CQ) phase of unmitigated disruptions are expected to impact strongly some regions of the actively cooled beryllium first wall (FW) and tungsten (W) divertor plasma-facing components (PFC). The very strong thermionic emission of W at melting temperature, or the halo currents entering the FW provide the driving (J X B ) force for melt motion. Additionally, relativistic runaway electrons (RE) carrying currents of up to 10 MA and kinetic energy up to 20 MJ could be formed during the CQ phase. The power width of this runaway beam is extremely small (~mm) and may thus induce extremely high volumetric heat loads on PFCs in very localized areas. This paper presents analysis of the CQ and RE-induced heat loads and the subsequent melt dynamics using the DINA-SMITER-GEANT4-MEMOS-U workflow developed at the ITER Organization. Emphasis is on scoping simulations to examine the thresholds for CQ heat flux and RE volumetric heat deposition before disruption mitigation will be mandatory to ensure PFC integrity and lifetime consistent with the ITER Research Plan. |
Wednesday, November 10, 2021 11:54AM - 12:06PM |
NO08.00013: Disruption mitigation by multiple shattered pellet injections in KSTAR Jayhyun Kim, Larry R Baylor, Michael Lehnen, Nicholas Eidietis, Soohwan Park, Daisuke Shiraki, Jaewook Kim, Jeongwon Yoo, Juhyeok Jang, Jun-Gyo Bak, Kwangpyo Kim, Kwanchul Lee, Young-chul Ghim, Gunsu S Yun, Kunsu Yi, Junewoo Juhn, Donggeun Lee, Min Uk Lee, Sehyun Bae, Shekar Thatipamula, Jeffery Herfindal For disruption mitigation, ITER has adapted to use multiple SPIs simultaneously or sequentially to safely dissipate stored energies. However, sufficient experimental assessment has not been made on the disruption mitigation by multiple SPIs. Before full-scale operation of ITER, it is desirable to optimize the usage of the multiple SPIs. In KSTAR, two identical SPIs were installed symmetrically 180 degrees apart toroidally to verify the effectiveness of multiple SPIs. The KSTAR SPI can inject up to six pellets simultaneously or with specific time delays. The experiments have been focused on characterizing the impact of delays in the arrival of the pellet fragments when performing multiple injections either from different toroidal locations or from different barrels at the same location. The experiments demonstrated that synchronized pellet injection into H-mode plasmas is more effective compared to delayed injections. These findings provided important input to key design requirements such as the jitter in pellet release and spread in pellet velocity, both related to the pellet launching technique. Furthermore, we are developing an optimal injection scheme to achieve both high density and effective radiation by injecting pure deuterium pellets and neon-doped pellets sequentially. |
Wednesday, November 10, 2021 12:06PM - 12:18PM |
NO08.00014: Deuterium shattered pellet assimilation in DIII-D Daisuke Shiraki, Jeffrey L Herfindal, Larry R Baylor, Eric M Hollmann, Zana Popovic, Claudio Marini, Jose A Boedo, Dmitry L Rudakov, Renato Perillo, Thomas B O'Gorman, Nicholas Eidietis, Andrey Lvovskiy, Charles J Lasnier In DIII-D, radial transport, MHD growth, and intrinsic impurity species radiation are found to be important in determining the disruption dynamics following D2 shattered pellet injection (SPI). SPI with hydrogen species is an essential capability for disruption mitigation in ITER, to reduce runaway electron generation by maximizing particle assimilation and reducing hot-tail seed generation. Particle flux and profile measurements indicate significant radial particle transport occurs during the pre-thermal-quench (TQ) and current quench (CQ) of D2 SPI shutdowns. Radial penetration is deeper for bigger pellets, significantly shortening the pre-TQ. When assimilation rates are insufficient to directly initiate a CQ, the pre-TQ timescales are found to be determined by the growth of n=1 MHD instabilities. TQ onset, and the variability in this timescale, is consistent with existing thresholds for non-SPI disruptions due to locked modes [1]. Subsequent CQ rates are dependent on carbon impurity levels, suggesting that sputtering sources may become important if additional radiating impurities are not injected. |
Wednesday, November 10, 2021 12:18PM - 12:30PM |
NO08.00015: Extrapolation of the Runaway Electron Benign Termination Scenario to ITER Carlos A Paz-Soldan, Ksenia Aleynikova, Pavel Aleynikov, Matthew T Beidler, Diego del-Castillo Negrete, Cedric Reux, Yueqiang Liu, Nicholas Eidietis, Eric M Hollmann D2 injection into mature runaway electron (RE) beams is found to enable access to a benign termination scenario that can mitigate MA-level RE currents without measurable wall heating. This result is enabled by the excitation of large and sudden MHD events (dB/B ~ 5%) that are found to promptly disperse the entire RE population over a large wetted area, with MHD accelerated by a recombined background plasma [1,2]. Fast RE loss timescales (<< ms) also prevent magnetic to kinetic energy conversion. We review benign termination phenomenology with published data from existing devices and focus on extrapolation to ITER, specifically: 1) vertical displacement event evolution and MHD instability access; 2) the required wetted area enhancement to disperse the kinetic energy; 3) the impact of the increased avalanche gain. Using the DINA code, we find that high current ITER RE beams should robustly access edge q of 3 & 2, where instability is expected. Using the MARS-F code, we find that the large-scale dispersal of RE kinetic energy is expected if dB/B is high. The large avalanche gain in ITER poses a severe challenge, likely requiring multiple cycles of the benign loss to fully terminate a high current RE beam. [1] Paz-Soldan et al PPCF 2019 & NF 2021 [2] Reux et al PRL 2021. |
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