Bulletin of the American Physical Society
61st Annual Meeting of the APS Division of Plasma Physics
Volume 64, Number 11
Monday–Friday, October 21–25, 2019; Fort Lauderdale, Florida
Session PO5: MF: Research in Support of ITER |
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Chair: Devon Battaglia, PPPL Room: Grand B |
Wednesday, October 23, 2019 2:00PM - 2:12PM |
PO5.00001: Synthetic diagnostics for ITER First Plasma Operation Joyeeta Sinha, Peter De Vries, Luca Zabeo, Joseph Snipes, Simon Pinches, Yuri Gribov ITER First Plasma (FP) operation aims to produce a plasma with current higher than 100kA for a duration longer than 100ms, which corresponds to the plasma initiation phase of ITER operation. However, the low values of plasma density and temperature associated with the initiation phase make it difficult to diagnose the plasma accurately. Therefore, it is essential to develop models for the available diagnostics for ITER FP operation to determine the necessary measurement ranges for plasma initiation and use them as inputs for controller development and assessment within the Plasma Control System. The use of accurate diagnostic models helps to optimally prepare for and analyze ITER FP operation. A model developed for the H-alpha spectrometer for FP shows that it is possible to measure H-alpha emission soon after breakdown for plasma temperatures higher than 3eV and densities higher than 5e17m$^{\mathrm{-3}}$. The diagnostic model for the interferometer shows that for the proposed ITER FP scenarios, accurate measurements are possible for plasma densities higher than 5e17m$^{\mathrm{-3}}$. Improved diagnosis of the plasma state is possible by combining the measurements from the different diagnostics for ITER FP operation. [Preview Abstract] |
Wednesday, October 23, 2019 2:12PM - 2:24PM |
PO5.00002: Helium L-H transition threshold studies in JET-ILW Emilia R. Solano, M Maslov, E Delabie, G Birkenmeier, I Jepu, A Shaw, J Hillesheim For ITER, it is important to establish the L-H transition power threshold (P$_{\mathrm{LH}})$, so H-modes can be investigated in its non-active phase, be it in Hydrogen, Helium, or suitable mixtures. In JET-C He and Deuterium appeared to have similar P$_{\mathrm{LH}}$ [1]. In JET-ILW we have shown that adding Helium to a Hydrogen plasma can reduce P$_{\mathrm{LH}}$ [2]. Here we report on recent L-H transition experiments with $^{\mathrm{4}}$Helium plasmas at JET-ILW, heated either with D-NBI or ICRH of H minority, with broad density scans. In all cases n$_{\mathrm{D+H}}$/n$_{\mathrm{e}}$\textless 5{\%} at the time of the L-H transition. The He and H$+$D concentrations were measured spectroscopically using the ratio of $^{\mathrm{4}}$He and D lines in an Optical Penning gauge in the subdivertor. Results at 1.8 T, 1.7 MA and 2.4 T, 2 MA reveal that the electron density at which the power threshold is minimum, n$_{\mathrm{e,min}}$, is 60-70{\%} higher in Helium plasmas than in Deuterium plasmas, while it is only 35{\%} higher for Hydrogen at 1.8 T. [1] D. McDonald, PPCF 46, p. 519 (2004). [2] J. Hillesheim et al, 27$^{\mathrm{th}}$ IAEA Fusion Energy Conference, Ahmedabad, India (2018). [Preview Abstract] |
Wednesday, October 23, 2019 2:24PM - 2:36PM |
PO5.00003: Multi-machine Scalings of n$=$1 and n$=$2 Error Field Correction N.C. Logan, J.-K. Park, Q. Hu, S. Yang, C. Paz-Soldan, E.J. Strait, S. Munaretto, Y. In, H. Wang, T. Markovic, M. Maraschek, L. Piron, P. Piovesan, Y. Gribov New power law scalings of the error field (EF) penetration thresholds across a wide range of tokamaks have been developed for toroidal mode numbers n$=$1 and 2. These scale the threshold in the EF aligned with the dominant mode of core resonant drive as calculated by GPEC and project values on the order of 1-10G for ITER (criteria the ITER tolerances and correction coils easily satisfy). Within the typical regimes of DIII-D, the nonlinear MHD code TM1 reproduces the primary scaling behaviors observed in experiments. However, not all experiments conform to the simple power scaling framework. Observations at high density in KSTAR and DIII-D show a regime transition that drastically alters the density scaling. Other anomalies are observable in the impact of nonresonant error fields, which can alter the threshold through rotation braking or impact on the L-H transition dynamics. These anomalies necessitate modification of the simple resonant scalings to better describe the EF correction challenge in ITER and existing tokamaks. [Preview Abstract] |
Wednesday, October 23, 2019 2:36PM - 2:48PM |
PO5.00004: ELM solutions developed in EAST towards ITER steady-state scenario Baonian Wan A stationary grassy ELM regime has been achieved in the EAST tokamak with metal wall and low rotation in the parameter range $q_{\mathrm{95}} \quad \ge $ 5.2 and $\beta_{\mathrm{p}} \quad \ge $ 1.1 with a confinement improvement factor $H_{\mathrm{98y2}}$ up to 1.4, $\beta_{\mathrm{N}}$ up to 2 and ELM frequency of 0.5-3 kHz. This regime exhibits good compatibility with high bootstrap current fraction and fully non-inductive operation, accessible in a broad density range, $n_{\mathrm{el}}$/$n_{\mathrm{GW}} \quad =$ 0.4-1.1, with line-averaged density, $n_{\mathrm{el}}$, up to 6.4\texttimes 10$^{\mathrm{19}}$m$^{\mathrm{-3}}$. Particle transport carried by the grassy ELMs provides strong impurity exhaust and good density control. High separatrix density makes this regime especially suitable for operation with divertor detachment. A new detachment feedback control scheme has been demonstrated, which combines divertor Langmuir probe and radiation signals to achieve sustained detachment without confinement degradation suitable for long-pulse operation of high-performance grassy ELM plasmas. ELM suppression has been achieved using Boron powder injection and high n RMP in EAST. Sustained ELM suppression in high-performance H-mode plasmas has been achieved with Boron powder injection with nearly constant plasma stored energy and density. The ELM suppression achieved by Boron powder injection appears to be more effective and robust than Lithium powder/granule injection, and has been applied to a broader parameter range. In addition, full ELM suppression without obvious drop of plasma confinement using n$=$4 RMP has been demonstrated for the first time in EAST. [Preview Abstract] |
Wednesday, October 23, 2019 2:48PM - 3:00PM |
PO5.00005: Favorable Transport Properties of the Wide Pedestal QH-Mode Regime for ITER Operation D. R. Ernst, K. H. Burrell, C.C. Petty, K. Barada, T. L. Rhodes, G. Wang, S. Haskey, C. Chrystal, B. A. Grierson, T. Odstr\v{c}il, T. Wilks, S. Houshmandyar Recent DIII-D experiments in the ELM-stable Wide Pedestal Quiescent H-mode regime [Ernst IAEA EX/2-2 (2018)] show confinement improves when electron cyclotron heating (ECH) replaces neutral beam power (NBI) (so far up to 77\% ECH)-- promising for burning plasma operation where $\alpha$-particles heat electrons. An Internal Transport Barrier (ITB) forms with on-axis ECH due to a strong inward electron thermal pinch, producing Te0$>$12 keV. Gyrokinetic simulations with GENE show ETG modes are stable and TEMs dominate. Confinement increases 60\% with 1/3 off-axis ECH (no ITB), with ion channel improvement evident in the core and pedestal. Wide Pedestal QH-Mode has been demonstrated with zero injected NBI torque throughout [K. H. Burrell APS 2018]. Separate measurements of the intrinsic torque show it balances the local beam orbit loss torque. The regime has been sustained in ITER shape [T. Wilks APS 2018] where impurity content is reduced relative to double null. Further, confinement does not degrade with NBI power in this regime. Projections indicate ITER Baseline Scenario Parameters with fusion gain 0.4 are attainable with $q_{95}$=3.3 and 6.6 MW NBI at constant power per particle, without the benefit of ECH. [Preview Abstract] |
Wednesday, October 23, 2019 3:00PM - 3:12PM |
PO5.00006: \textbf{On the Divertor Heat Flux Width Scaling} X.Q. Xu, N.M. Li, Zeyu Li, X.Y. Wang, T.Y. Xia, Ben Zhu, V.S. Chan The BOUT$++$ code has been used to simulate edge plasma electromagnetic (EM) turbulence and transport, and to study the role of EM turbulence in setting the scrape-off layer (SOL) heat flux width \begin{figure}[htbp] \centerline{\includegraphics[width=0.17in,height=0.22in]{020720191.eps}} \label{fig1} \end{figure} . More than a dozen tokamak discharges from C-Mod, DIII-D, EAST, ITER and CFETR have been simulated \quad with encouraging success. The parallel electron heat fluxes onto the target from the BOUT$++$ simulations of C-Mod, DIII-D, and EAST follow the experimental heat flux width scaling of the inverse dependence on the poloidal magnetic field. Transport simulations show two distinct regimes: drift dominant regime and turbulence dominant regime. The simulations for ITER and CFETR indicate that divertor heat flux width $\lambda_{\mathrm{q}}$ of the future large machines may no longer follows the 1/B$_{\mathrm{pol,MP\thinspace }}$experimental Eich scaling and the Goldston's HD model gives a pessimistic limit of divertor heat flux width. The simulation results show a transition from a drift dominant regime to a turbulence dominant regime from current machines to future machines such as ITER and CFETR for two reasons. (1) The magnetic drift-based radial transport decreases due to large CFETR and ITER machine sizes. (2) the SOL turbulence thermal diffusivity increases due to larger turbulent fluxes ejected from the pedestal into the SOL when operating in a different pedestal structure, from an ELM-free H-mode pedestal regime to a small/grassy ELM regime. [Preview Abstract] |
Wednesday, October 23, 2019 3:12PM - 3:24PM |
PO5.00007: Interpretable Disruption Prediction Using Random Forest on EAST and DIII-D. Cristina Rea, Kevin Montes, Robert Granetz, Brian Sammuli, Keith Erickson, QiPing Yuan, Dalong Chen, Biao Shen, Wenhui Hu, Bingjia Xiao This contribution presents the Machine Learning-based disruption predictors working in real-time and near real-time on DIII-D and EAST tokamaks. The interpretable predictive algorithm used provides predictions of impending disruptions as well as which input features (i.e., plasma signals) are responsible for the prediction. This enables intelligent prioritization of the available control actuators in the Plasma Control System (PCS). Adopting explainable predictions is necessary to validate data-driven models for extrapolation to ITER and future fusion reactors. This Disruption Predictor using Random Forest (DPRF) algorithm operated in real-time on DIII-D during the 2018 experimental campaign [Rea et al., NF 2019], and was also interfaced with EAST PCS. For both these devices, DPRF's accumulated fraction of predicted disruptions is between 80-90{\%} when optimized to simulate PCS alarms in offline testing [Montes et al., NF 2019]. Thanks to the feature contribution analysis, it is possible to identify and discriminate among different types of disruptions. Interpretable disruption prediction algorithms are mandatory to seamlessly integrate data-driven models and PCS actuators, thus supporting viable solutions for disruption avoidance on ITER. [Preview Abstract] |
Wednesday, October 23, 2019 3:24PM - 3:36PM |
PO5.00008: Measurement and modeling of shattered pellet assimilation in DIII-D D. Shiraki, J.L. Herfindal, L.R. Baylor, E.M. Hollmann, R.A. Moyer, I. Bykov, C.J. Lasnier, N.W. Eidietis, R.M. Sweeney, R. Raman The particle assimilation during shattered pellet injection (SPI) in DIII-D plasmas has been measured in a wide range of plasma conditions, allowing empirical scalings based on plasma parameters to be derived. The initial electron temperature is found to be the dominant quantity determining the net assimilation, while plasma density is found to have little or no impact. Later in the current quench (CQ), Ohmic dissipation of the poloidal magnetic energy becomes an important energy source for sustaining ionization. The measured densities following SPI are found to be in good agreement with predictions from reduced 0D simulations accounting for ablation shielding of pellet fragments and non-coronal radiation rates. Simulations of neon and deuterium pellets suggest that added deuterium further raises the electron density, but the dilution cooling can reduce the assimilation of the primary radiating impurity neon. The 0D model is applied to ITER plasmas, where peak densities of several 10$^{\mathrm{21}}$ m$^{\mathrm{-3}}$ are predicted, with CQ rates remaining within allowable values. [Preview Abstract] |
Wednesday, October 23, 2019 3:36PM - 3:48PM |
PO5.00009: Runaway electron beam dynamics in DIII-D: energy distribution, current profile, and RE-driven instabilities A. Lvovskiy, C. Paz-Soldan, N.W. Eidietis, Y.Q. Liu, K. Thome, P. Aleynikov, A. Dal Molin, M. Nocente, W.W. Heidbrink, R. Moyer Novel measurements of the post-disruption runaway electron (RE) beam energy distribution function show potential for kinetic instabilities and provide a constraint on the internal current profile. The RE energy distribution measured by inverting bremsstrahlung spectra has a non-monotonic feature at 5 MeV supporting the presence of free energy to drive kinetic instabilities. The RE current profile is reconstructed for the first time using the spatially resolved RE energy distribution. It is found to be more peaked than the pre-disruption current, with higher internal inductance, suggesting preferential formation of REs in the core plasma or potentially a radially inward motion of REs. The accessed relatively low-current (180 kA) RE beam is found to be stable, likely due to its elevated safety factor profile. From this base stable equilibrium, instability is accessed by applying large loop voltage. Under an accelerating loop voltage, an internal kink mode is observed. It leads to sawtooth-like relaxation of the RE current profile, but drives no RE loss. Under a decelerating voltage, frequency chirping correlated with RE loss is detected. This work yields unique opportunities for validating models that predict RE physics in ITER. [Preview Abstract] |
Wednesday, October 23, 2019 3:48PM - 4:00PM |
PO5.00010: First long pulse experiments with the actively cooled W-divertor in WEST A Ekedahl, C Bourdelle, J Bucalossi, N Fedorczak, T Loarer, P Moreau, E Tsitrone, J-F Artaud, L Delpech, C Desgranges, P Devynck, T Dittmar, R Dumont, J Gaspar, C Gil, M Goniche, J P Gunn, C C Klepper, P Maget, J Morales, R Nouailletas, Y Peysson, X Regal-Mezin, C Reux, D Vezinet WEST is a full tungsten (W) superconducting tokamak with large aspect ratio (\textasciitilde 5), whose mission is to assess power exhaust with ITER actively cooled divertor technology and to master long pulse operation in a metallic environment. Repetitive and reliable long L-mode X-point discharges (\textasciitilde 30 s) have been achieved in WEST, accumulating \textasciitilde 20 minutes of plasma over two days. They were performed in upper single null configuration on the actively cooled W-divertor, using 2.7 MW Lower Hybrid (LH) power and nitrogen seeding in some discharges. The plasma radiation and density remained constant during the long pulses, indicating that no W-accumulation was taking place. In short pulses, 5.0 MW of LH power has been coupled in L-mode discharges. The central electron temperature increases linearly with the input power and reaches 5 keV with 5.0 MW of LH power. Loop voltage \textless 0.15 V is achieved for line average density$_{\mathrm{\thinspace }}$\textless 4*10$^{\mathrm{19}}$ m$^{\mathrm{-3}}$, with no sign of loss of LH current drive efficiency at the highest density. In the experiments, the heat load pattern on the divertor target is monitored with a unique set of PFC diagnostics (infrared systems, Langmuir probes, thermocouples, Fiber Bragg gratings). [Preview Abstract] |
Wednesday, October 23, 2019 4:00PM - 4:12PM |
PO5.00011: Tungsten in the divertor of DIII-D: effect of material choice on intrinsic fuel source on ELM time scale I Bykov, E Hollmann, M Groth, A Pigarov, J Guterl, D Rudakov, H Wang, J Watkins, Ch Lasnier, A McLean, T Abrams, D Thomas Significant advances have been made in understanding fuel recycling and sourcing from a W-coated divertor in DIII-D during the Metal Rings Campaign (MRC). Simultaneous measurements of D atoms and D$_{\mathrm{2}}$ molecules recycling at the Outer Strike Point (OSP) enabled the quantification of the relative contribution ($F)$ of D atoms to the total recycling flux on tungsten. Between ELMs, $F$\textasciitilde 40{\%}, consistent with expectations if all atomic recycling is due to reflections. In an opaque SOL of a larger tokamak such as ITER, the fast reflected D may dominate the intrinsic fueling of the pedestal because the low energy neutrals will be screened in the divertor. During ELMs, $F$ increased to 60{\%}. This effect was studied in a DiMES experiment with a variety of metal samples (Mo, W, W fuzz, and Ti). In L-mode DiMES was biased to vary the ion impact energy, E$_{\mathrm{i}}$, to simulate the effect of ELMs in controlled conditions. On all samples an increase of E$_{\mathrm{i}}$ led to a transient increase of the recycling fraction, similar to the MRC results.$\backslash $pardThe fueling efficiency of the neutral D source in the divertor depends on the flux and the energy of the D reflected from the target. Therefore, the properties of the target material are crucial in controlling the source of divertor D available for edge fueling. During the MRC, a relatively small divertor target area (0.6{\%} of the total wall area) covered in W led to an 8{\%} edge n$_{\mathrm{e}}$ increase with OSP placed on W in L-mode, qualitatively consistent with EDGE2D-EIRENE. However, the effect of W on H-mode edge profiles in DIII-D was small. [Preview Abstract] |
Wednesday, October 23, 2019 4:12PM - 4:24PM |
PO5.00012: Assessment of ITER W divertor performance during early operation phases Jae-Sun Park, Xavier Bonnin, Richard Pitts Until now, the focus of plasma boundary simulations at ITER has been on tungsten divertor performance under the most challenging conditions of H-mode, burning plasmas. However, operations will begin at low power in L-mode with H fuel only. Here we use the SOLPS-ITER code to assess the divertor performance in this first phase at q95$=$3 with SOL power up to PSOL$=$20 MW. The focus will be on a study of the impact on detachment behaviour of gas fuelling location (main chamber (MC), divertor) and possible Be coating of the divertor surfaces driven by MC erosion and migration. Throughput scans show that the divertor material does impact the recycling balance between H atoms and molecules and hence the relative contributions to momentum and power losses, but the overall effect on the detachment behaviour is marginal. Similarly, the influence of gas puffing location is small, but does impact neutral penetration in the MC SOL. Even at this low PSOL, full detachment is not observed in the code at the maximum operational fuel throughput. For divertor puffing, the upstream separatrix density saturates after the initial low recycling phase, behaviour found also at high performance. Analysis shows that this is a consequence of the vertical target configuration and physical size of the divertor. [Preview Abstract] |
Wednesday, October 23, 2019 4:24PM - 4:36PM |
PO5.00013: SOLPS modeling of neutral effects on pedestal structure during pellet fueling Aaron Sontag, M.W. Shafer, D. Shiraki, F. Laggner, A.O. Nelson SOLPS-ITER has been used to model a set of discharges with the neutral particle source location varied from being dominantly external with an external gas puff for fueling and ECH for particle-free core heating, to being dominantly internally sourced with pellet fueling and NBI heating. Pellet fueling is modeled as a constant neutral source in the pedestal region where pellet ablation is observed. The discharges with core pellet and NBI fueling show increased pedestal density and an increase of over 30\% in the ratio of the pedestal to separatrix density. Modeling shows that all cases have similar particle confinement in the steep-gradient region of the pedestal, but moderate pellet fueling increases particle diffusion from the no-fueling case and decreases electron thermal transport inside $\psi_{N}$ of 0.9. Higher fueling leads to increased electron thermal transport in the steep gradient region for all cases due to increased convective flow. Ion charge exchange momentum loss is increased by a factor of 4 across the entire pedestal with external fueling as compared to core fueling. The inclusion of $\nabla$B drift effects results in an inward flux of particles for all cases, significantly altering the neutral deposition profile. [Preview Abstract] |
Wednesday, October 23, 2019 4:36PM - 4:48PM |
PO5.00014: Overview on Density Pedestal Structure: Role of Fueling versus Transport Saskia Mordijck, Jerry Hughes, Rich Groebner, Ahmed Diallo In this paper we will give an overview of the results from the 2019 Joint Research Target on what sets the pedestal density structure based on experimental observations from DIII-D, C-Mod and NSTX as well as theoretical models. In ITER and other future burning plasma magnetic confinement devices, the ionization source will be pushed further out into the Scrape-Off Layer. This motivates the question, how much of the pedestal density structure is governed by this edge ionization source versus plasma transport effects? Theoretically, several turbulent modes have been identified which could provide a ‘pinch’ like up-gradient transport mechanism. In a source-free region, a peaked density profile relies on this non-diffusive inward pinch component. In an opaque edge, without the existence of such a pinch, the pedestal density structure on closed flux surfaces would eventually disappear. The experiments and modeling as part of the JRT will address the effects of opaqueness, divertor closure and fueling locations and their role in determining the density pedestal structure. Preliminary results indicate that details of the pedestal ionization source and of peeling-ballooning stability have important effects on the density pedestal structure. [Preview Abstract] |
Wednesday, October 23, 2019 4:48PM - 5:00PM |
PO5.00015: Self-consistent modeling investigation of density fueling needs on ITER and future devices Joseph McClenaghan, Jie Zhang, Lang Lao, Orso Meneghini, Paul Parks, Sterling Smith, Wen Wu Self-consistent modeling using the STEP workflow in OMFIT (predicting pedestal with EPED, core profiles with TGYRO, current profile with ONETWO, and EFIT for equilibrium) suggests ITER and future devices such as CFETR will benefit from high-density operation (Greenwald limit fraction $f_{gw}\sim 0.7-1.3$). Regimes with operational $n_e$ near the Greenwald limit will likely need peaked $n_e$ profiles so that the $n_e$ pedestal remains below the Greenwald limit. Peaked $n_e$ profiles can be achieved with the help of pellet injection. Furthermore, the primary source of tritium in ITER will be provided via pellet injection. The Pellet Ablation Module (PAM), which predicts the density source of an ablated pellet based on the PELLET module, has been developed for predicting pellet fueling for transport studies, and has been incorporated into the STEP workflow for predictive modeling. On ITER the effect of pellet fueling is examined on two high-density scenarios: the super-H mode inductive scenario and the steady-state high $\beta_p$ scenario. On CFETR, with a mid-radius density source, an average of $10^{22}$ particles/sec are required to predict $n_e$ and $T_i$ necessary for the 1000 MW advanced scenario. [Preview Abstract] |
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