28th IAEA Fusion Energy Conference (FEC 2020)
Virtual Event
-
-
-
The 2019-2020 scientific and technological programme exploits JET’s currently unique capabilities: Tritium handling and ITER-like wall (ILW: Be wall and W divertor). It is the culmination of years of concerted scientific and engineering work, with the ILW installation in 2010, improved diagnostic capabilities, now fully available, a major Neutral Beam Injection (NBI) upgrade providing record power in 2019 (P$_{NBI}$ up to 32MW), and the technical & procedural preparation for safe operation with tritium. Research along three complementary axes yielded a wealth of new results since last IAEA. Firstly, the JET plasma programme delivered scenarios suitable for high fusion power (P$_{FUS}$) and alpha particle ($\alpha$) physics in the coming D-T campaign (DTE2), with record sustained neutron rates, as well as plasmas for clarifying the impact of isotope mass on plasma core, edge and plasma-wall interactions, and for ITER pre-fusion power operation (e.g. L-H transition in He plasmas). The efficacy of the newly installed Shattered Pellet Injector (SPI) [ref1] for mitigating disruption forces and runaway electrons was demonstrated, informing ITER disruption management. Secondly, research on the consequences of long-term exposure to plasma in the ILW was completed, with emphasis on wall damage and fuel retention, and including analyses of wall materials and dust particles. This will help validate assumptions and codes for the design & operation of ITER and DEMO. Thirdly, the nuclear technology programme aiming to deliver the maximum technological return from operations with D, T and D-T [ref3] benefited from the high D-D neutron yield in 2019 (2.26x10$^{19}$n), securing new results for validating radiation transport and activation simulation codes, and nuclear data for ITER. Measuring systems are ready for collecting data in T and in D-T campaigns producing 14MeV neutrons.
Integrated scenarios preparation for high P$_{FUS}$ sustained for 5s (i.e. relevant to energy confinement times in JET) progressed significantly for the two routes investigated: ‘Baseline’ (q$_{95}$~3, I$_P$≥3MA, $\beta$$_N$<2) and ‘Hybrid’ (tailored q-profile, q$_{95}$~4.5, I$_P$ ≤2.7MA, $\beta$$_N$≥2.4). Peak neutron rate of 4.2x10$^{16}$n/s (2.7x10$^{16}$n/s averaged over 5s) are obtained simultaneously with tolerable divertor temperatures and controlled high/medium Z impurity for the full pulse duration in Baseline plasmas at 3.3T/3.5MA, with P$_{TOT}$=34MW (NBI and Ion Cyclotron Resonance Heating (ICRH)). Pellets help controlling the ELM frequency (f$_{ELM}$) needed for impurity flushing, with low total D$_2$ throughput for high confinement. Hybrid plasmas developed to 3.4T/2.3MA reached 4.8x10$^{16}$n/s but MHD avoidance and f$_{ELM}$ control must be optimised for improved, steady performance. The equivalent P$_{D-T}$ for these pulses is consistent with past predictions at same B$_T$, I$_P$, P$_{TOT}$ [ref4], giving confidence in the theory-based modelling. Further gains are likely with 40MW now reachable and higher I$_P$, with divertor heat loads controlled by strike-point sweeping, thus prospects for reaching the target (5x10$^{16}$n/s) are good. In these conditions P$_{D-T}$=11-16MW is predicted by theory-based physics models, with range due to uncertainties in the pedestal predictions and to whether isotope effects are included or not. Fast particle diagnostics, significantly improved since DTE1, can now detect small amount of $\alpha$’s, as shown in dedicated experiments making use of 3-ion ICRH scheme (D-(D$_{NBI}$)-3He) to create MeV range particles, with $\alpha$ (≈10$^{16}$s$^{-1}$) from D+$^3$He reactions. Simultaneous detection of He and hydrogen isotopes with an enhanced high resolution sub-divertor residual gas analyser, as planned for ITER, has been demonstrated.
Experiments and modelling preparing the T campaign. Observations that the impact of isotope on H-mode plasmas comes mainly from the pedestal [ref5] motivated recent gyrokinetic (GK) theoretical investigations of JET pedestals showing that the toroidal branch of the ETG instability can be driven at ion-scale poloidal wavelengths and may be responsible for significant inter-ELM pedestal heat transport. However, in some regimes, isotope effects on core plasma may also be important [ref4]. New experiments in D$_2$ and H$_2$ L-mode plasmas and related core GK modelling show that, in plasmas with a strong stabilizing effect of fast particles, differences in fast particle content with isotope mass may lead to strong deviations from the gyro-Bohm scaling of core transport. Recently developed ICRH-only H-mode plasmas (low input torque, dominant e-heating) show the same normalised confinement factor (H$_{98(y,2)}$) and T$_e$ profiles as their NBI-only counterpart at same PTOT, but n$_e$ profile for the NBI case is 50% higher due to NBI fuelling and possibly different particle transport. Work to clarify the impact of edge/divertor was performed. Experiments at 2MA/2.3T, low triangularity ($\delta$) demonstrated that changes in H$_{98(y,2)}$ and pedestal T$_e$ due to divertor configuration can be condensed into a single trend when mapped to the target T$_e$ as the main parameter governing recycling conditions rather than D$_2$ fuelling rate. A high performance neon seeded scenario (2.7T/2.5MA, high-δ shape) with edge conditions closer to ITER was developed. Neon seeding leads to significant increase (by ≈50%) in pedestal pressure and T$_e$ (from 0.4keV to 0.8keV) and in H$_{98(y,2)}$ (from 0.6 to 0.9) with mitigated divertor power loads. The well diagnosed discharges are used for validating physics-based SOL-edge modelling, increasing confidence in ITER divertor design basis and supporting deployment of neon over chemically reactive N$_2$ as seed gas in ITER.
JET disruption management programme is in two parts: 1) disruption avoidance based on improved termination techniques and on real-time detection of unhealthy plasmas with jump to controlled termination, causing significant reduction in disruption rate in baseline (60% to 20%) and hybrid plasmas (9%), and 2) disruption mitigation with SPI performed as part of a collaboration between ITER, US and Europe. After successful installation and commissioning, extensive experiments took place with the JET SPI demonstrating very good reliability. By varying the neon content in the SPI pellets, the disruption current quench time can be controlled efficiently in JET, scaling to the range required by ITER. High Z impurity SPI also demonstrated run-away electron suppression. Additionally, it was discovered that D$_2$ SPI applied to a high current run-away beam leads to benign impacts on the wall, suggesting a new potential solution for run-away electron control in ITER.
Plasma-facing components (PFC) long term exposure in ILW. Retrieval of PFC, wall probes (including test mirrors) and dust for ex-situ studies after three ILW campaigns provide deep insight into material erosion and deposition. Low mobilization of dust during in-vessel operations is shown. Emphasis was placed on material damage such as melting of the Be upper dump plates (UPD) and the identification of factors triggering this process. Comprehensive studies including imaging survey, morphology changes, mass loss and fuel inventory analysis on the most affected UDP tiles was performed. The undisputed reason for melting was unmitigated disruption events which tend to move the melt layers in the poloidal direction resulting in formation of upwards going waterfall-like structures of molten metal. The halo current is believed to provide the j x B force driving the melt layer motion. Global material migration results constitute a unique dataset for modelling and thus improved predictions for ITER.
[ref1] L. R. Baylor, et al., Nucl. Fusion 68 (2019) 211, [ref2] I. Jepu et al., Nucl. Fusion 59 (2019) 086009, [ref3] P. Batistoni et al, Fusion Engineering and Design Vol 109-11, 2016, [ref4] J. Garcia et al. Nucl. Fusion 59 (2019) 086047 [ref5] C. F. Maggi et al., Plasma Phys Control Fusion 60 (2018) 014045
Speaker: Dr Joelle Mailloux (UKAEA) -
DIII-D physics research addresses critical challenges for operation of ITER and the next generation of fusion energy devices through a focus on innovations to provide solutions for high performance long pulse operation, development of scenarios integrating high performance core and boundary plasmas, and fundamental plasma science and model validation. Substantial increases in off-axis current drive efficiency from an innovative top launch system for EC power (Fig. 1), and in pressure broadening for Alfven eigenmode control from a co-/counter-Ip steerable off-axis neutral beam, both improve the prospects for optimization of future long pulse/steady state high performance tokamak operation. A high beta-p optimized-core scenario with an internal transport barrier that projects nearly to Q=10 in ITER at 9 MA was coupled to a detached divertor, and a Super H-mode optimized-pedestal scenario with co-Ip beam injection (Fig. 2), was coupled to a radiative divertor. Fundamental studies into the evolution of the pedestal pressure profile, and electron vs. ion heat flux, measuring both density and magnetic field fluctuations, validate predictive models of pedestal recovery after ELMs (Fig. 3).
Link to High Resolution Figures 1, 2, 3
The achievement of more than double the off-axis ECCD efficiency using top launch geometry compared with conventional low field side (LFS) launch, as predicted by quasi-linear Fokker-Planck simulations, is due to the longer absorption path for the EC waves which also interact with higher v|| electrons that suffer fewer trapping effects than outside launch. In addition, the new unique co-/counter-Ip steerable off-axis neutral beam broadens the energetic particle (EP) pressure profile and reduces Alfven eigenmode (AE) drive in scenarios with both high toroidal rotation and those with net zero average input torque. New EP measurements show a beam current threshold for Compressional AEs, insensitivity of Beta-induced Acoustic AEs to fast beam ions, and resolution of phase-space flows caused by AEs, from first-of-a-kind Ion Cyclotron Emission (ICE) and Imaging Neutral Particle Analyzer (INPA) data.
Studies of high current runaway electron (RE) beams reveals excitation of current-driven (low safety factor) kink instabilities that promptly terminate the RE beam on an Alfvenic time-scale, offering an unexpected alternate pathway to RE beam mitigation without collisional dissipation. Newly developed real-time stability boundary proximity control and neural-net-based Vertical Displacement Event (VDE) growth-rate calculations are shown to prevent VDEs. The effectiveness of emergency shutdown and disruption prevention tools projects to at least 50% of ITER disruptions being delayed until normalized-Ip is at safe levels, and demonstration of a novel technique for healing flux surface with 3D fields shows promise for providing current quench (CQ) control. Single and multiple Shattered Pellet Injection particle assimilation rates and current quench (CQ) densities are shown to be predictable from 0-D simulations and empirical scaling laws.
Several core-edge integration scenarios demonstrate coupling of a high performance core and radiative divertor operation for target heat flux control. High density and stored energy plasmas with Super H-mode edge pedestals were made both in a lower single null shape accessible by JET and in a higher triangularity near double null shape coupled to a radiative divertor for target heat flux control using nitrogen injection in a core-edge integrated scenario. High-performance plasma with high poloidal beta, large Shafranov shift, and Te and Ti internal transport barriers coupled to a detached divertor with active feedback-controlled Nitrogen puffing also demonstrated integration of core-edge solutions. A high performance hybrid core demonstrated compatibility with radiative divertor operation using Neon or Argon gas injection. Core impurity peaking in the hybrid was substantially reduced using near-axis electron cyclotron heating.
The ability to predict the impurity seeding needed for divertor dissipation has advanced through new capability for measuring charge-state resolved densities of impurity species in the divertor. Also electric drifts in detached divertors with convection dominated heat transport lead to expanded radiative volume. Using these advances, SOLPS-ITER simulations show the synergy between SOL drifts and the SAS divertor geometry for achieving lower density detachment. Modeling of intra-ELM tungsten gross erosion with an analytic Free-Streaming plus Recycling Model is now validated in ITER-relevant mitigated-ELM regimes using pellet pacing and RMPs. SOL tungsten transport in plasmas with both BT directions is consistent with strong entrainment in SOL flows and ExB drift effects.
Advances in pedestal physics through new measurements of density and internal magnetic fluctuations suggest a possible role for micro-tearing and trapped electron modes in DIII-D pedestal transport. Main ion CER measurements indicate ion heat flux is anomalous at low collisionality and transitions to near neoclassical levels at high collisionality. Plasma rotation scans, and both new non-linear analytic theory and 2-fluid code simulations, confirm that ELM suppression by RMPs requires near zero ExB velocity at the top of the pedestal, and achieving suppression appears to be closely linked to a high field side plasma response. The wide pedestal QH-mode regime was obtained with zero input beam torque, electron heating, and LSN shape, consistent with requirements for ITER.
Recent fundamental research on L-H mode power threshold physics shows that turbulence driven shear flow through Reynolds stress and the coexistence of modes associated with various instabilities can lower the L-H power threshold across multiple parameters: eg. q95 and ion grad-B drift direction. Application of RMPs raises turbulence decorrelation rates and reduces Reynolds stress driven flow and flow shear, hence increasing the L-H power threshold. Finally, plasmas with negative triangularity show weak power degradation of H-mode level core confinement while maintaining an L-mode-like edge without ELMs.
In 2020 and beyond DIII-D will install additional tools for optimizing tokamak operation through current and heating profile control using a low field side 1 MW helicon high harmonic fast wave CD system, a unique high field side Lower Hybrid CD system, increased ECH power, and coupling to boundary advances using a new high power closed divertor and a wall insertion test station. Experiments will continue the optimized coupling of high performance core and high power density divertor solutions.
This work was supported in part by the US DOE under contracts DE-FC02-04ER54698 and DE-AC52-07NA27344Speaker: Max Fenstermacher (LLNL @ DIII-D) -
Since the last IAEA-FEC, the EAST research programme has been, in support of ITER and CFETR, focused on development of the long-pulse steady-state (fully non-inductive) high beta H-mode scenario with active control of stationary and transient divertor heat and particle fluxes $^{[1]}$. The operational domain of the steady-state H-mode plasma scenario on EAST has been significantly extended with the ITER-like configuration, plasma control and heating schemes. Several important milestones on the developments of the high beta H-mode scenario and its related key physics and technologies have been achieved.
-
A minute time scale long-pulse steady-state high beta H-mode discharge (shown in figure 1) with the major normalized plasma parameters similar to the design of the CFETR Phase-III 1GW fusion power operation scenario ($\beta_P$ ~ 2.0, $\beta_N$ ~ 1.6, $f_{bs}$ ~ 50%, $H_{98(y2)}$ > 1.3 at $q_{95}$ = 6.5~7.5) has been successfully established and sustained by pure RF heating on EAST with the ITER-like tungsten divertor $^{[2]}$, as shown in figure 1. The important feature of this high beta H-mode plasma scenario is that, due to the stabilization effect of the Shafranov shift on the plasma turbulence, a higher $\beta_P$ results in a better plasma confinement as shown in figure 2. Further simulation suggested that a high density gradient promotes the ITB formation in high $\beta_P$ plasmas, which might further benefit the development of this high beta plasma scenario towards a high density regime.
-
Active control of divertor radiation has been successfully integrated into the high beta H-mode ($\beta_P$ ~ 2.5, $\beta_N$ ~ 2.0, $f_{bs}$ ~ 50%) plasma scenario without a degradation of the plasma confinement ($H_{98(y2)}$ > 1.2) at high density ($n_e$/$n_{GW}$ ~ 0.7) and moderate edge safety factor ($q_{95}$ ~ 6.7) $^{[2]}$. The peak heat flux on the tungsten divertor was reduced by ~30% with active impurity seeding of a mixture of 50% neon and 50% deuterium. The high-Z impurity concentration in the plasma core has been well controlled in a low level by applying the on-axis ECRH and reducing the fast ion losses through beam energy optimization.
-
The grassy-ELM regime has been extended to the normalized parameter space designed for the CFETR 1GW fusion power operation scenario. This regime exhibits good compatibility with high $f_{bs}$ and fully non-inductive operation, being characterized by a low pedestal density gradient and a wide pedestal, which prevents large-ELM crashes due to an expansion of the peeling-ballooning boundary after the initial pedestal collapse, indicated by BOUT++ nonlinear simulations. High separatrix density makes this regime especially suitable for operation with divertor detachment. Several feedback control schemes have been developed to achieve sustained detachment with good core confinement $^{[3]}$. This includes control of total radiation power, target electron temperature, and particle flux measured by divertor Langmuir probes or a combination of the control of target electron temperature and AXUV radiation near the X point. Integration of these detachment feedback control schemes with the grassy-ELM regimes and the high $\beta_P$ scenario has been demonstrated with neon seeding, which provides an integrated high beta scenario applicable to long-pulse operation $^{[4]}$.
-
ELM suppression has been achieved using various different methods on EAST. Full suppression of ELMs has been demonstrated, for the first time, for ITER-like low torque injection plasmas by using n=4 resonant magnetic perturbations (RMPs) $^{[5]}$. A moderate reduction (~5%) of the energy confinement has been observed together with significant reduction of both the plasma density(20%)and the Tungsten concentration (a factor of 2) during ELM suppression. The ELM suppression window agrees well with the prediction by MARS-F modelling. Robust ELM suppression by Boron powder injection without confinement degradation or even with confinement improvement has been achieved in a wide parameter range. EHO-like edge coherent modes were excited during the ELM suppression phase by Boron powder injection. In addition, simulation results from BOUT++ confirm that both the helical current filaments (HCF) driven by lower-hybrid waves (LHWs) and RF sheath effects on the ICRF antenna contribute to ELM suppression.
-
A flowing liquid lithium (FLiLi) limiter plate has been successfully assembled and tested in EAST for the first time. This plate was designed based on the concept of liquid metal infused trenches (LiMIT), which is using thermoelectric MHD to drive liquid Li flow along the surface channels. The preliminary results show that with the increase of Li flow rate, the fuel particle recycling is gradually reduced, and the plasma performance is slightly improved. There was no obvious Li burst and limiter damage even at a high injection power of 5.5 MW. In addition, ELM mitigation was observed with FLiLi operation.
-
For the first time, EAST has been operated with helium to support the ITER needs. The H-mode power threshold in a helium plasma is found to be 1.2-2.2 times higher than the scaling law in deuterium plasma with pure RF-heating $^{[6]}$. The $C_{He}$ plays a crucial role in determining the energy confinement and pedestal characteristics in helium H-mode plasmas. Divertor detachment is more difficult to achieve in He than in D. The control of divertor heat loads and W sources is achieved by RMP along with neon impurity seeding. The inter-ELM W erosion rate in He is about 3 times that in D with similar divertor conditions, while the intra-ELM W sputtering source shows a strong positive correlation with the ELM frequency $^{[7]}$.
-
A new lower tungsten divertor with a higher closure has been designed and to be installed on EAST in 2020. Several key subsystems, including the heating & current drive, cryogenic, plasma control and diagnostics will be upgraded accordingly for achieving the next milestones i) >400s long-pulse H-mode operation with ~50% bootstrap current fraction, and ii) demonstration of power exhaust at ~10 MW power injection for >100s.
Reference:
1) B. N. Wan et al 2019 Nucl. Fusion 59 112003
2) X. Z. Gong et al 2020 this conference
3) L. Wang et al 2020 this conference
4) G. S. Xu et al 2020 this conference
5) Y. W. Sun et al 2020 this conference
6) B. Zhang et al 2020 this conference
7) R. Ding et al 2020 this conference
Speaker: Baonian Wan (Institute of Plasma Physics, Chinese Academy of Sciences) -
-
-
-
The ADITYA/ADITYA-U tokamaks are equipped with state-of-art spectroscopic diagnostics in the visible and vacuum ultraviolet (VUV) region of the spectra. These spectroscopic systems are used to study several physics problems in ADITYA tokamak as well as in ADITYA-U, which is an upgraded version of ADITYA, having capability of producing shaped plasmas. The physics studies addressed in this paper are studies of impurity behaviour, dynamics of neutrals, measurement of plasma rotation, a novel technique to estimate electron temperature in the edge region by simulating the temporal profile of H-alpha intensity observed at the time of hydrogen gas puff, estimation of particle confinement and ion temperature etc.
The ADITYA/ADITYA-U tokamak plasma is diagnosed mainly by detecting the radiation emanating from the plasma using spectrometers, filter-PMT combination and fast imaging cameras. Several spectroscopic diagnostics were developed for ADITYA, which was having a poloidal ring limiter, and its present upgraded version ADITYA-U tokamak with toroidal belt limiter. The line averaged ne and core Te of the plasmas are in the range of 1 - 6.0 x 1019 m-3 and 300-700 eV, respectively. The toroidal magnetic field was ~ 0.75 to 1.4 T in these studies.
The fast visible imaging camera was attached to tokamak using an imaging fiber bundle on a re-entrant view port to acquire images of plasma evolution at ~ 14 kHz frame rate. This system enabled to observe the thick toroidal filaments (figure 1a) during the disruptive phase of ADITYA tokamak plasma and those were explained in term of interchange modes 1. Further the boundary of the plasma was also detected by the imaging camera and was used to control plasma position. Furthermore, the fast visible imaging was used to study the plasma collapse during micron-sized particle injection through an electromagnetically driven payload for disruption mitigation studies. Photo multiplier tube (PMT) based system, which is normally used to monitor temporal evolution of spectral lines, such as H, from OII and CIII and visible continuum emission, was extensively employed to study the hydrogen recycling, impurities behaviour and Zeff [2]. It was found that the Zeff were reduced 1.5 to 3.0 for the analyzed discharges after the Li coating of plasma facing components of ADITYA tokamak compared to the Zeff values of 2.0 - 4.5 for the discharges before the Li coating [3]. The radial profiles of H measured using 8 channels PMT array based diagnostic were modelled using DEGAS2 neutral transport code. The details studies on the contribution from various atomic and molecular processes show that substantial amount of hydrogen molecules survive inside the plasma up to 4 cm from the limiter of ADITYA tokamak having typical edge Te of 10- 15 eV [4]. A PMT based system is recently developed to monitor the radial profile of visible continuum for the Zeff profile estimation of ADITYA-U tokamak plasma.
An indigenously developed impurity transport code based on semi–implicit numerical scheme was used for modelling of the radial profile of O4+ emissions at 650 nm from ADITYA plasma and the studies has revealed the higher oxygen diffusion coefficient at plasma edge than the neoclassical values as shown in figure 1b [5]. The O4+ emission profile were recorded using a high resolution visible spectroscopy system having 1 m focal length visible spectrometer and charge coupled device (CCD) camera, which capable of multi-track measurement simultaneously from eight lines of sight. The radial profiles of impurity toroidal rotation velocities from ADITYA-U plasmas were also obtained by measuring Doppler shifted passive charge exchange emission at 529 nm from C5+ ion. The lights were collected using a front-end optic placed inside the re-entrant tube from a tangential port of the tokamak. The data were converted using an Abel-like matrix inversion to get radial profile of impurity rotation and ion temperature. The maximum C5+ rotation velocity and ion temperature values were ~ 15 km/s and ~ 120 eV, respectively, for the analyzed discharges and the reversal of toroidal rotation was observed for the discharges having electron densities 3x1019 m-3 [6] as depicted in figure 1c. The spatial profile of temperature of neutral hydrogen and low ionized impurities ions of ADITYA-U plasmas were also determined from the Doppler broadened H and spectral lines from lowly ionized carbon and oxygen impurities after incorporating the influence of Zeeman Effect into the apparent broadening of spectral lines. It is found that the neutral temperature (TH) is having two components, warm and hot and their values vary in the range of 3 - 5 eV (shown in figure 1d) and 18 - 25 eV, respectively [7].
A toroidal grating based VUV survey spectroscopy system working in the 10 - 180 nm range regularly monitored the survey spectrum from low and highly ionized impurities in the plasmas enabling to do the details studies of impurity transport in the central region of the plasma. The measured VUV spectral lines from highly ionized irons, such as Fe14+ and Fe15+, were analyzed to study the ion impurity behaviour in high density ADITYA tokamak plasmas and its concentration was found to be lower than 0.1% [8]. For the first time, ADITYA-U has experimentally demonstrated the use of electromagnetic pellet injection for firing the impurity pellets into the tokamak plasmas for disruption mitigation studies. Lithium titanate (Li2TiO3) and Lithium carbonate (Li2CO3) particles of 50 – 80 micron particles are injected in to the plasma. The spectral lines of Li showed a self-absorption peak on top of the emission spectra during the injection, which has been diagnosed thoroughly to understand the disruption mechanisms.
References:
1. Banerjee S. et al. Phys. Plasmas 24, (2017) 102513.
2. Chowdhuri, M. B., et al. Plasma Sci. Technol. 15 (2013) 123.
3. Chowdhuri, M. B., et al. Plasma Phys. Control. Fusion 62 (2020) 035015.
4. Dey, R., et al. Nucl. Fusion 59 (2019) 076005.
5. A Bhattacharya, Phys. Plasma 27, ( 2020) 023303
6. Shukla, G., et al. Nucl. Fusion 59 (2019) 106049.
7. Yadava, N., et al. Atoms 7 (2019) 87.
8. Patel, S., et al. Nucl. Fusion 59 (2019) 086019.Speaker: RANJANA MANCHANDA (INSTITUTE FOR PLASMA RESEARCH) -
The EUROfusion Work Package PFC (Plasma-Facing Components) focuses on critical plasma-surface interaction studies and components qualification in view of upcoming ITER operation and in preparation for DEMO exhaust solutions. This poster gives an overview of the latest main results in WP PFC, as well as their implications for ITER and DEMO.
Helium-Tungsten Interaction
The Helium-Tungsten (He-W) interaction was studied extensively in the full W tokamak WEST (CEA), in dedicated experiments in the divertor manipulator of ASDEX Upgrade with He pre-damaged W (MPG), in dedicated high fluence experiments on ITER monoblocks in the MAGNUM-PSI device (DIFFER), and compared with other laboratory studies and first principle modelling. The goal is the predictability towards He-W interaction in the different phases of ITER (He plasmas and DT plasmas with He ash). Figure 1 shows an example of the formation of W fuzz in samples exposed to ASDEX Upgrade H-mode plasmas. The results, showing pronounced evolution of the surface morphology and the creation of He nanobubbles below the surface, will be used to benchmark models relevant for lifetime estimations of the first ITER divertor. This work related to He-W interaction is done jointly with the ITER Organisation.
Fig.1) Formation of W fuzz in AUG [Brezinsek et al.]Tungsten surface morphology and erosion
The erosion yield of the W fuzz has not only been studied under tokamak conditions, but also in laboratory arrangements with monoenergetic ion beam bombardment. These experiments provide the sputtering yield of 3D structured surfaces, as well as W fuzz, as a function of impact energy and impact angle. Complementary simulations of the surface evolution of e.g. W fuzz was performed with the TRI3DYN code. The reference binary-collision approximation code 3DTRIMSP has been benchmarked in 2018 against ion beam experiments and 3D structures and provides the input basis for larger scale modelling with ERO2.0 and WallDYN 3D. Overall, the combination of these laboratory and tokamak experiments will be used to benchmark also global migration codes with novel surface roughness modules (e.g. in ERO2.0).
Fig. 2) Preferential sputtering direction at rough surfaces [Eksaeva et al.]Fig. 2 shows an example of the so-called needle formation under perpendicular impact on low-Z surfaces (Be), modelled with the roughness surface module implemented in ERO2.0. This example is an outcome of the long-standing PISCES-B cooperation between WP PFC and UCSD with regular guest scientists from EU send to the US to perform studies related to Be.
Finally, 3D modelling utilising the ERO2.0 code without the surface roughness module has been applied to benchmark JET (Be, W) and WEST (W) experiments as well as to predict the Be migration in ITER. A typical example from the WEST application is shown in fig. 3 providing the experimental plasma information, the plasma background as well as the corresponding ERO2.0 run. In the future, dedicated runs with the surface roughness of technical surfaces and appropriate material mixing will be done for the different devices in support of predictions for ITER and DEMO.
Fig. 3) ERO2.0 modelling of W erosion in WEST [Gallo et al.]Qualification of ITER-like monoblocks
Tests of ITER-like monoblocks at elevated temperature were performed in high fluence experiments in MAGNUM-PSI, reaching record values of D+ fluence of 1031m-2, comparable to a year of ITER divertor operation without any visible damage and extremely low fuel retention. Re-crystallisation of W occurred in the high fluence experiments in D at the theoretically predicted material temperatures, which gives confidence in the lifetime predictions of the ITER divertor. Combined exposure of He and D ions in PSI-2 showed no impact on the recrystallization temperature, but on changes in the morphology. Additional emphasis was put on the study of monoblock castellation and shaping. ITER monoblocks with different shapings were installed in WEST for this purpose. The local damage of optical hot spots, which are predicted to occur in ITER at the projection of toroidal gaps, has been observed. Dedicated PIC modelling was applied and identified the particle and power flux to these locations. The same modelling was used for ITER predictions about castellation and shaping.
Parallel to these studies, ELM simulations by combined laser and plasma exposition and e-beam impact were carried out up to the record number of 106 ELM-like pulses, to study the response of W to repetitive transient heat loads. All three W PFC studies contribute to the determination of the operational window for the W divertor in ITER regarding heat loads, transients and surface temperature conditions.
Complementary, studies with plasma and high heat flux exposure on advanced materials such as PIM-W, WfW etc. - developed under WP MAT- started in WP PFC in 2019. Multiple disruption-like loads on W were carried out in the plasma gun facility QSPA in KIPT. The surface damage evolution was compared to previous studies on the ITER-grade reference W regarding synergistic effects with combined plasma and heat load.
Fuel retention
Finally, dedicated experiments were carried out to quantify the fuel retention in W for combined D and N2 exposure in laboratory and linear plasma conditions. A dramatic enhancement of the retention in the near-surface area has been identified, with local fuel content in the percentage range depending on the exposition conditions. WN formation has been observed near the surface, acting in some cases as permeation barrier. Thus, the actual enhancement in fuel retention is solely near the surface, which can interact during plasma operation with the recycling and fuelling of tokamak plasmas. The second system investigated regarding the fuel retention by implantation is the mixed He-D system in W in combination with neutron damage proxies. Self-damaging by W ions was applied to simulate defects in W and mimic the impact of fusion neutrons. Enhanced retention by the synergistic effects has been documented. Detailed modelling was applied to simulate the observed enhancement of retention in the near surface. The fuel release efficiency was documented by stepwise heating of the W material. Fuel recovery at high temperatures is challenging if He impact and neutron damage are present simultaneously.
In order to quantify in-situ or in-operando the retention of T and D in the first wall material of fusion devices, a dedicated task utilising LIBS and LIA-QMS was executed successfully in 2019 (see contribution of H. van der Meiden et al., “LIBS - monitoring of tritium and impurities in the first wall of fusion devices”). The experiments demonstrated the capability to resolve D and T as well as quantify the fuel in different types of ITER-like codeposits containing in particular Be and W as well as impurities as found in JET samples.
Bibliography
S. Brezinsek et al., PSI 2020, Jeju, KoreaA Eksaeva et al. PSI 2020, Jeju, Korea
A Gallo et al., 2020 Phys. Scr.2020 014013
Speaker: Michael Reinhart (EUROfusion) -
Achieving net energy production in magnetic confinement fusion devices is a key milestone in the quest for fusion energy. With the mission of demonstrating net fusion energy, the SPARC tokamak is being designed jointly by the MIT Plasma Science and Fusion Center and Commonwealth Fusion Systems. Its study of reactor-relevant, alpha-heating-dominated scenarios and high power density regimes will help retire risk for ITER operations and for fusion power plants. A team of over 100 engineers and scientists is on track to deliver a toroidal field model coil using high-temperature superconductor (HTS) technology by 2021, with the engineering design of the tokamak progressing in parallel. Negotiations with potential host sites in the Northeast US are underway, with start of construction planned in 2021 and operation expected in 2025.
SPARC will be a pulsed machine operating with Deuterium-Tritium (DT) fuel and with ICRF auxiliary heating. The high strength of the magnetic field ($B_T>12.0T$ on axis), will allow operation at high plasma current and high absolute density, leading to net fusion output in a device with a size comparable to current tokamaks ($R_0<2.0m$). In particular, the SPARC mission objective has been established as demonstration and study of Q>2 plasma conditions, where Q is the ratio between the total fusion power and the external power absorbed in the plasma. Figure 1 depicts the poloidal cross-section of the Version 1C (SPARC V1C) design iteration, and Figure 2 indicates main plasma parameters for the baseline DT H-mode plasma discharge.
Following a traditional design workflow (1), SPARC parameters are first selected using empirical scaling laws and plasma operation contour (POPCON) analysis. Figure 3 represents the operational space for SPARC V1C for its baseline scenario, demonstrating that $Q\approx11$ can be reached with conservative assumptions ($H_{98}=1.0$ confinement, and $\nu_{Ti}=2.5$, $\nu_{ne}=1.3$ profile peaking factors, consistent with empirical predictions). Total fusion power remains below administrative limits for the machine ($P_{fus}<140MW$), and safety factor ($q^*=3.05$), normalized density ($f_G=0.37$) and normalized pressure ($\beta_N=1.05$) are at reasonably safe levels of operation.
The development and validation of theory-based reduced models allow integrated simulations to also inform the design of SPARC. To this end, simulations with the TRANSP code (2) coupled with the TGLF model (3) for turbulence and EPED (4) for pedestal stability are performed. Figure 4 depicts simulated temperature and density profiles. $H_{98}\approx1.0$ is predicted, and fusion gain results in $Q\approx8.2$. The good agreement between the two independent workflows (empirical and theory-based simulations) provides high confidence that SPARC will accomplish its $Q>2$ mission. During nominal operation, D-T(3He) ICRF minority heating at 120 MHz will be utilized for on axis heating of both 3He and T. AORSA and CQL3D (5) simulations are in good agreement with TRANSP, which uses TORIC (6) to model ICRF. Single-pass absorption is excellent, and minimum losses of ICRF power (~1%) to alphas are predicted.
Loss of fast ions (alphas and RF-tail ions) due to toroidal field (TF) ripple can be a major issue for the design of DT tokamaks, as it can lead to excessive localized wall heating. The effect of first-orbit, classical and TF ripple in SPARC has been studied with ASCOT (7). Simulations indicate that the total losses are small (<2%), due to the low edge TF ripple (0.15%) in the SPARC design. There is no concentration of losses toroidally and only modest concentration poloidally in the current TF design.
Managing divertor heat flux will be challenging in SPARC, but unmitigated levels are comparable to ITER. To ensure divertor survivability with conservative assumptions (i.e., not relying on partial or complete divertor detachment), the poloidal field coil set and central solenoid are being designed to ensure that a fast strike point sweep (~1 Hz) can be achieved. Thermal simulations of the divertor target indicate that sweeping during the flat-top is sufficient to ensure divertor survivability with only a moderate divertor radiation fraction. SPARC will be equipped with impurity gas injection to attain detached-divertor scenarios. The feasibility of an “advanced” divertor is also being assessed.
In summary, SPARC will be an important experiment to study burning plasma physics and will be a proof-of-principle for high-field, compact fusion power plants. The SPARC design is converging towards a self-consistent model of the machine with robust engineering and physics. Conservative estimates of fusion gain show significant margin for the Q>2 mission, leaving room for extensive exploration of burning plasma physics regimes.
This work was supported by Commonwealth Fusion Systems.
(1) ITER Physics Expert Group on Confinement and Transport et al. Nucl. Fusion 39, 2175 (1999).
(2) J. Breslau et al. USDOE SC-FES (2018).
(3) G.M. Staebler et al. Phys. Plasmas 14, 055909 (2007).
(4) P.B. Snyder et al. Phys. Plasmas 16, 056118 (2009).
(5) E. F. Jaeger et al. Phys. Plasmas 13, 056101 (2006).
(6) M. Brambilla et al., Plasma Phys. Con. Fus. 41, 1 (1999).
(7) A. Snicker et al. Nucl. Fusion 52, 094011 (2012).Speaker: P. Rodriguez Fernandez (MIT Plasma Science and Fusion Center) -
Inertial confinement fusion (ICF) aims to assemble and confine a dense, high pressure fusion fuel over a relatively short timescale (≪1 μs) compared to magnetic confinement fusion (> 1 s). This is typically accomplished by imploding a spherical capsule at high implosion velocities (>350 km/s) to obtain the fuel temperatures (>4 keV) and areal densities (ρR >0.3 g/cm2) required for ignition.1 Magneto-inertial fusion (MIF) utilizes magnetic fields that relax these requirements by limiting thermal conduction losses and introducing magnetic confinement of charged fusion products. On the Z Machine at Sandia National Laboratories, we are pursuing a specific pulsed-power2 driven MIF concept called Magnetized Liner Inertial Fusion (MagLIF).3 MagLIF is the first MIF concept to demonstrate fusion-relevant temperatures, significant neutron production, and magnetic trapping of charged fusion products4,5, and has the potential to generate multi-MJ yields and significant fuel self-heating on a next-generation pulsed power machine.6
In MagLIF, a centimeter-scale cylindrical tube, or “liner,” is filled with a fusion fuel (typically deuterium gas), pre-magnetized using an axial magnetic field of 10-20 T using Helmholtz coils, pre-heated to an average temperature of 100-200 eV via a kilojoule-class laser, and finally radially imploded over ~100 ns via the Lorentz force to velocities of ~70 km/s using 15-20 megaamperes of current from the Z Machine. This process is schematically demonstrated in Figure 1. The laser preheat increases the initial adiabat of the fuel, which is then compressed in a quasi-adiabatic implosion to reach fusion-relevant conditions. The axial magnetic field, which flux-compresses to >1000 T near peak convergence, limits thermal conduction losses from the hot fusion fuel to the comparatively cold liner walls during the implosion and simultaneously increases trapping of charged fusion particles in the narrow radial direction during stagnation.
The first MagLIF experiments demonstrated that the axial magnetic field, the laser preheat, and the implosion are all required to generate thermonuclear fusion yields. Without any of these inputs, no significant yield could be generated. Utilizing pure deuterium fuel, these experiments produced multi-keV fuel temperatures, neutron energy spectra consistent with a thermonuclear plasma, and produced yields up to 2x1012 neutrons—a DT equivalent energy of 0.3 kJ.4 Analysis of the secondary deuterium-tritium (DT) neutrons demonstrated the fuel column was highly magnetized, with the fuel radius exceeding the average Larmor radius of DD-produced fast tritons, a promising and necessary requirement for trapping of alpha particles.5,7
Subsequently, our efforts have focused on improving the stability of liner implosions, increasing the fuel preheat energies, and simultaneously increasing the applied magnetic field and current delivered. Controlling the magneto Rayleigh-Taylor (MRT) instability is required to achieve uniform compression and minimize areal density variations in the liner. Simulations suggested that applying a dielectric coating to our beryllium liners would mitigate the early-time electrothermal instability,8 which is believed to seed the more deleterious MRT instability. Experiments verified this stabilizing effect9 and have produced highly uniform stagnation columns compared to uncoated targets, as shown in Figure 2.
To increase the laser preheat, we developed a new laser platform that incorporates phase smoothing10 of the laser beam in addition to a ~20 J pre-pulse that disassembles the laser-entrance window (a thin polyimide foil) on the top of the target that is required to initially contain the fusion fuel. This platform demonstrated an energy coupling efficiency of ~50% of the main ~2 kJ laser pulse to the fuel and simultaneously reduced mix from the polyimide foil into the fusion fuel.11 The axial magnetic field and current delivered to the target were simultaneously increased from 10 to 15 T and 16.1 MA to 19.4 MA, respectively, by reducing the inductance of the power-feed leading up to the target, increasing the anode-cathode gaps to reduce parasitic current losses, and employing more powerful Helmholtz coils. Simultaneously implementing the increased preheat, axial magnetic field and current delivered resulted in a record MagLIF performance of 1.1x1013 neutron yield (equivalent to 2 kJ DT energy produced), nearly an order of magnitude greater than previous experiments on this platform.
Numerical simulations show additional improvements are attainable by further increasing the applied magnetic field to 30 T, the laser-preheat energy coupled to 6 kJ, and target current to 22 MA, with >100 kJ DT fusion yields produced in 2D simulations using parameters that should be achievable on the Z Machine. Our efforts in the next five years are directed at increasing these input parameters simultaneously. Scaling this target to a next-generation facility is encouraging—fusion yields in excess of 10 MJ are possible with currents of 60 MA, laser preheat energies of 40 kJ, and applied magnetic fields of 19 T. Even larger yields approaching a gigajoule may be possible by propagating the fusion burn into a layer of frozen DT ice on the inside surface of the liner.6,12
Despite the impressive predicted yields, developing MagLIF as an energy producing source will be challenging.13 While the electrical energy delivered to the target is efficient (5-10%), it also generates a destructive post-shot explosion (even in the absence of fusion yields) that destroys the metallic power feed that delivers current to the target, limiting present-day operations to a single shot per day. Gigajoule producing targets would produce a tremendous number of neutrons that could adversely affect pulsed power components while activating materials to potentially hazardous levels. Conceptual solutions to these problems have been investigated, including recyclable transmission lines14 and neutron-absorbing blankets;15 however, our present efforts in MagLIF are directed towards demonstrating the fundamental physics and possibility of attaining significant fusion gains on larger-scale facilities. Surmounting this scientific difficulty remains the first obstacle towards harnessing a magneto-inertial fusion-energy producing system.
References
1. J. Lindl, Phys. Plasmas 2, 3933 (1995).
2. R.D. McBride et al., IEEE Trans. Plasma Sci. 46, (2018).
3. S.A. Slutz et al., Phys. Plasmas 17, 056303 (2010).
4. M.R. Gomez et al., Phys. Rev. Lett. 113, 155003 (2014).
5. P.F Schmit et al., Phys. Rev. Lett. 113, 155004 (2014).
6. S.A. Slutz and R.A. Vesey, Phys. Rev. Lett. 108, 025003 (2012).
7. P.F. Knapp et al., Phys. Plasmas 22, 056312 (2015).
8. K.J. Peterson et al., Phys. Rev. Lett. 112, 135002 (2014).
9. T.J. Awe et al., Phys. Rev. Lett. 116, 065001 (2016).
10. M. Geissel et al., Phys. Plasmas 25, 022706 (2018).
11. A.J. Harvey-Thompson et al., Phys. Plasmas 25, 112705 (2018)
12. A.B. Sefkow et al., Phys. Plasmas 21, 072711 (2014).
13. D.B. Sinars et al., J. Fusion Energy 35, 78 (2016).
14. S.A. Slutz et al., Phys. Plasmas 10, 429 (2003).
15. B.G. Logan, Fusion Engineering and Design, 22, 1953 (1993).This paper describes objective technical results and analysis. Any subjective views or opinions that might be expressed in the paper do not necessarily represent the views of the U.S. Department of Energy or the United States Government.
Sandia National Laboratories is a multimission laboratory managed and operated by National Technology and Engineering Solutions of Sandia, LLC., a wholly owned subsidiary of Honeywell International, Inc., for the U.S. Department of Energy’s National Nuclear Security Administration under contract DE-NA-0003525.
Speaker: D. A. Yager-Elorriaga (Sandia National Laboratories) -
Abstract
The COMPASS tokamak, operated in the Institute of Plasma Physics of the Czech Academy of Sciences in 2009 – 2020, is one of few devices with an ITER-like plasma shape. Its flexibility, extensive set of diagnostics, and NBI heating allow to address key issues in the fusion research in support of ITER and DEMO, such as edge and SOL physics, the L-H transition, runaway electrons and disruption studies, plasma-wall interaction.
Recent results related to these fields, obtained in the last two years, till the final COMPASS tokamak shutdown in September 2020, are the subject this contribution.L-H transition
Extensive sets of experiments have been performed recently in order to study the L-H transition.
The dependence of the L-H power threshold on the X-point height above the divertor PLH(|X-div|) was analyzed in the framework of causal graphical modelling. This motivated the separation (conditioning) of transitions into those with q95~3 and those with higher or lower q95 and also the normalization (counterfactual reasoning) of PLH to a common reference density in order to block confounding effects. The results (see figure) show a clear linear trend where PLH increases by 30-40 kW (~18%) per 1 cm of the X-point height above the divertor. While the trend with |X-div| is similar for all the discharges, those with q95~3 have a base value of PLH larger by 50 kW. This 30% increase in PLH around q95~3 seems to be related to the presence of intrinsic error fields.
![Density-normalized L-H power threshold P<sub>LH</sub> as a function of the X-point height above the divertor |X-div| (left). Critical radial electric field E<sub>r</sub> in the SOL at which the L-H transition occurs (right). In both figures, several outliers are observed for q<sub>95</sub> close to 3. [1, 2]](https://www.hron.name/test/figure.jpg)
Simultaneous measurements of the radial electric field Er in the SOL and inside the separatrix shows that Er increases (in absolute value), both upstream and downstream and inside the separatrix, with decreasing X-point height. This is qualitatively consistent with transport modeling in [3] and also consistent with the idea that the consequent change in the ExB shearing rate could be responsible for the change of PLH.
Furthermore, the effect of controlled HFS error field (EF) on the L-H transition was studied in detail, utilizing the unique COMPASS HFS 3D coils. Using these coils, a displacement of the central solenoid was simulated, while different sets of coils were used to assess the error field correction (EFC) from the LFS and top/bottom of the vessel, which is of high importance for ITER, having in mind the detrimental effects of HFS EF observed on NSTX-U recently [4]. It was shown that the residual EF, after the correction of HFS EF by LFS EFC, can lead to disruptions if present during the L-H transition. While in case of NBI-assisted L-H transitions the disruption rate was around 50% as reported in earlier COMPASS results, in case of purely Ohmic L-H transitions (induced by X-point ramp) the disruptions were inevitable. The critical parameter appears to be the plasma rotation which is very low during Ohmic L-H transitions in COMPASS (and is also expected to be generally low in ITER), but injection of even a small external momentum (PNBI < 100 kW) is sufficient to prevent the disruption.
Run-away electrons (RE) physics
Both experiments and modelling effort contributed significantly to the understanding of crucial topics, such as more efficient mitigation, suppression of RE beam generation, its feedback control, beam detection, transport of RE and their interaction with MHD.
Radial stability of the relativistic RE beam and the role of RE energy was studied and based on this, a new algorithm for the RE beam feedback control was implemented [5]. Vertical electron cyclotron emission (V-ECE) diagnostics has been commissioned [6] and used to monitor RE seed population in the plasma, which is helpful in the mitigation scenarios. The evolution of the RE beam impact and its overall energy was measured using a new dedicated calorimetry head. Studies of low and high Z impurity material impact on the RE beam dynamics showed very promising results thanks to a new vacuum pellet injector system. Besides systematic surveys of the effects of the massive gas injection and impurity seeding, the effect of externally applied resonant magnetic perturbation on the RE dynamics was addressed [7].
A new full orbit particle tracking code taking into account radiation damping and 3D perturbed magnetic field was developed. RE transport model in presence of natural magnetic perturbations based on fractional diffusion theory was developed and qualitative comparison with experiments was made, evaluation of the RE diffusion coefficient with a guiding-center particle code is ongoing. The conditions under which the radiation reaction acting on the charged particle prevails over the collisions were investigated analytically.
Disruptions studies
Disruption studies are focused on electromagnetic loads on the machine. For the first time simultaneous measurements of plasma current asymmetries as well as toroidal and poloidal vessel currents including their poloidal distribution have been performed [8]. New diamagnetic technique for measurements of poloidal current in the wall has been successfully validated in experiments on COMPASS [9]. The disruption forces modelled by CarMa0NL code have been validated against the COMPASS data, and then applied for the design of COMPASS-U vacuum vessel [10].
Two special divertor tiles were installed in COMPASS to directly measure currents flowing between these tiles and the plasma during vertical displacement events (VDEs). They allow better understanding of current pattern distribution within vessel structure and divertor and aim on testing the model of asymmetric toroidal eddy currents (ATEC) [11].
The vacuum vessel motion during disruptions has been studied simultaneously by displacement sensors and accelerometers.
Liquid Metal Divertor (LMD) experiments
Liquid metals are considered as an option to overcome issues related to conventional plasma facing components (PFC), such as melting of leading edges, cracking, morphology and heat capacity deterioration subsequent to neutron irradiation or re-crystallization. One candidate technology is the capillary porous system (CPS) where the liquid metal is impregnated in a metallic mesh and confined against MHD effects by capillary forces. However, this potential solution comes with new issues such as resilience to transients, tritium retention, evaporation, etc. Most of these issues were investigated experimentally for the first time under ELMy H-mode conditions in a tokamak divertor. [12, 13].
A specially designed CPS module filled with liquid Li and with a liquid Sn alloy (LiSn) was installed in the COMPASS divertor for two separate dedicated power exhaust experiments. Good power handling capability of both liquid metals was observed for an averaged deposited, perpendicular heat flux up to 12 MW/m2. The CPS module was exposed to ELMs with relative energy ~3% and a local peak energy fluence ~15 kJ/m2 locally. No droplets were ejected from the CPS surface and no damage of the CPS mesh was observed, as well as no contamination of the core and SOL plasmas by Sn.
References
1 J Seidl, H-mode workshop, 2019
[2] O Grover, EFTSOMP workshop, 2019
[3] AV Chankin, NME, 2017
[4] CE Myers, 2016
[5] O Ficker, NF, 2019
[6] M Farnik, RSI, 2019
[7] J Mlynar, PPCF, 2019
[8] E Matveeva, 46th EPS Plasma Physics, 2019
[9] VV Yanovskiy, 46th EPS Plasma Physics, 2019
[10] VV Yanovskiy, IAEA FEC 2020
[11] R Roccella, NF, 2016
[12] R Dejarnac, PSI, 2020
[13] J Horacek, PSI, 2020Speaker: Dr Martin Hron (Institute of Plasma Physics of The Czech Academy of Sciences) -
The TCV tokamak continues to leverage its unique shaping capabilities, flexible heating systems and modern control system to address critical issues in preparation for ITER and a fusion power plant. For the 2019-20 campaign its configurational flexibility has been enhanced with the installation of divertor gas baffles and its diagnostic capabilities with an extensive set of upgrades. Experiments are performed in part by topical teams, under the auspices of the EUROfusion medium-size tokamak programme and local teams at the Swiss Plasma Center together with international collaborators, resulting in a rich and focused scientific programme.
Auxiliary heating is provided by NBI and ECRH. An improved acceleration grid for the NBI system reduced power losses in the duct allowing for a 2.5x increase in injected energy with an injected power up to 1.3MW. The legacy ECRH system of 1.4MW from two 83GHz gyrotrons (X2) and 0.9MW from two 118GHz gyrotrons (X3) was enhanced by a 1MW dual frequency (84/126GHZ) gyrotron for X2 or X3 heating. The new gyrotron performs as designed validating the numerical models used for its development.
The most conspicuous upgrade was the installation of removable gas baffles that separate the vessel into main and divertor chambers, Fig. 1. The baffles seek to increase the divertor neutral pressure and thereby facilitate the extrapolation to future devices, such as ITER that will rely upon operation with high divertor neutral pressure [1]. They allow for various divertor and limited configurations. Experimentally, Ohmically heated diverted discharges confirm SOLPS-ITER predictions with up to a 5x increase in divertor neutral pressure. Optional fuelling into the divertor or the main chamber can disentangle the effects of fuelling rate, divertor or main chamber neutral pressure, $p_\mathrm{n,div/main}$, and plasma density, $n_\mathrm{e}$.
As predicted, increasing $p_\mathrm{n,div}$ in the baffled divertor facilitates access to detachment, which commences at ~30% lower $n_\mathrm{e}$. Scanning the plasma plugging by displacing the X-point with respect to the baffles indicates that the installed divertor closure may be close to optimal. Experiments also provide further evidence that the onset of detachment is determined by $p_\mathrm{n,div}$ rather than $n_\mathrm{e}$. These experiments ultimately seek to validate edge models. Reversing the toroidal field direction reveals changes in the target currents and the formation of a potential well below the X-point in reverse field as predicted by SOLPS-ITER including drifts.
All previously obtained alternative configurations were achieved in the baffled divertor with detailed investigations first focusing on the Super-X divertor [1]. Specific configurations designed to disentangle the effects of a large target radius, $R_\mathrm{t}$, and the angle between the divertor leg and the target surface confirm an expected strong dependence of the detachment onset on the angle, whereas a predicted dependence on $R_\mathrm{t}$ remains elusive.
Two new gas-puff imaging (GPI) systems, diagnosing the X-point region and the outboard midplane, have greatly increased the ability to investigate scrape-off layer (SOL) transport. Characterisation of the SOL turbulence was also extended into H-mode, linking $p_\mathrm{n,div}$ to the formation of the density shoulder [2].
H-mode studies were facilitated by an apparent reduction in the power threshold, $P_\mathrm{LH}$, for the baffled divertor. Here, the inter-ELM target temperatures were lower and nitrogen seeding led to detachment. At the H-mode density limit, a full MARFE develops as a dense strongly radiating region at the X-point that subsequently moves up the HFS edge.
Particular attention was dedicated to the pedestal in type-I ELMy H-modes, where the baffles lead to significantly higher $T_\mathrm{e,ped}$ and, hence, higher $p_\mathrm{e,ped}$, Fig. 2. The pedestal degrades with fuelling that increases $n_\mathrm{e,sep}$, but decreases $T_\mathrm{e,ped}$, consistent with previous findings. The role of $n_\mathrm{e,sep}$ in the pedestal is further highlighted in discharges with a range of $R_\mathrm{t}$ that require different fuelling rates to obtain the same $n_\mathrm{e,sep}$, but then display the same pedestal characteristics, highlighting alternative divertors’ weak effect on pedestal and core properties.
In a continued effort to extrapolate ELMy H-mode performance to the ITER baseline (IBL) scenario, NBI and X3 heated H-modes succeeded in matching the ITER targets of $\kappa$=1.7, $\delta$=0.4, $\beta_\mathrm{N}$=1.8 and $q_{95}$=3.0 whilst retaining good confinement ($H_\mathrm{98y2}$~1) [3]. The Greenwald fraction reached 0.6, but at those densities X2 ECH absorption becomes unreliable and ELM triggered NTMs were only avoided by lowering $I_\mathrm{P}$ with a stationary demonstration at $q_\mathrm{95}$=3.6.
In the quest to ELM-free regimes, negative triangularity (NT), Fig. 1 (c), is confirmed as an attractive scenario with an L-mode confinement matching that of H-modes for positive triangularity. NBI heating extended the operating space of NT plasmas to the IBL value of $\beta_\mathrm{N}$. However, similarly to the low-$q_{95}$ IBL in TCV, these discharges are prone to NTMs, but inaccessible to X2 ECH control. Reciprocating probe plunges past the LCFS of Ohmic NT plasmas confirm a reduced turbulent E×B flux extending from the core into the plasma edge. The measurements are corroborated by measurements from the new mid-plane GPI system.
TCV continues to address critical aspects of the discharge evolution that may limit plasma performance or even pose a danger to components in ITER and future power plants. This includes NTMs, with a successful validation of a new analytical model of the classical island stability, which will facilitate NTM pre-emption in future devices. Enabled by NBI, fast-ion studies are gaining prominence with the development of robust scenarios that display rich, fast-ion driven, MHD spectra and the commissioning of a fast-ion loss detector. Further experiments aim at understanding and controlling runaway electrons (REs) [4], whether created at low density or by mitigated disruptions, as they may cause severe damage in larger devices. Filtered imaging for multiple wavelength ranges provided the first measurements of synchrotron radiation on TCV, Fig. 3, revealing information about location, size and energy distribution of REs and even allowing to detect pre-disruption seed distributions. RE scenarios were extended from Ne and Ar to He, Kr and Xe injection, and to NT and diverted configurations, increasing the space for model validation. In addition, strategies to purge the impurities after the RE beam formation with further $\mathrm{D}_2$ injection are being explored.
TCV also continues to employ its flexible digital control system to enhance available control solutions [5]. With a view to future long-pulse tokamak discharges, a generic plasma control framework has been developed, implemented and applied to avoid density limit disruptions by controlling the NBI power based upon an estimated proximity to the disruptive boundary. An ability to re-assign EC sources to $\beta_\mathrm{N}$ or NTM control was demonstrated. Plasma exhaust control, for future reactors, was explored using an estimate of the C-III radiation profile along the divertor leg, indicative of the local $T_\mathrm{e}$, with a feedback control of the distance of the radiation from the target to the X-point demonstrated using gas injection as actuator for both L- and H-mode scenarios.
Various short baffled and un-baffled campaigns are planned for 2020 highlighting yet another dimension in TCV’s signature flexibility.[1] C. Theiler, et al., this conference, [2] N. Vianello, et al., this conference, [3] O. Sauter, et al., this conference, [4] G. Papp, et al., this conference, [5] F. Felici, et al., this conference.
Speaker: Holger Reimerdes (Ecole Polytechnique Fédérale de Lausanne (EPFL), Swiss Plasma Center (SPC)) -
KSTAR$^{1,2}$ program has been focused on resolving the key physics and engineering issues for ITER and future fusion reactors utilizing unique capabilities of KSTAR. First of all, a new advanced scenario was developed targeting steady-state operation based on the early diverting and heating during the ramp-up phase of plasma current and significant progress has been made in shape control to address the MA level of plasma current and stationary ITER-similar shape (ISS). It is demonstrated effective use of the H&CD with instrumented plasma control and shaping parameters became a key to access to the advanced operation scenarios such as high $β_p$, high $l_i$, high $q_{min}$, hybrid, internal transport barrier (ITB) and low $q_{95}$ operation. The examples of advanced scenarios are shown in Figure 1. The stationary ITB (fig 1a) is successfully reproduced with comparable confinement as H-mode level ($H_{89}$ ~ 2) both in limited and USN configuration, a low qmin scenario (fig 1b) is developed based on early diverting and delayed core heating approach and finally stable long-pulse H-mode operation (fig 1c) was extended upto 88 sec.
Recent KSTAR 3D experiments have focused on several ITER-relevant issues, such as divertor heat flux broadening in 3-row vs 2-row resonant magnetic perturbations (RMPs) on ELM-crash suppression, RMP-driven ELM-crash-suppression on ITER-like low $q_{95}$ (~3.2-3.4) and the characterization of ELM-crash suppression window in terms of normalized electron collisionality ($\nu^*_e$) and plasma toroidal rotation ($V_{tor}$) at pedestal top. Strong up-down asymmetry in 3-row configuration was identified and effect of the kink/anti-kink configuration was also clarified for ELM suppression in LSN plasmas. We have demonstrated the ISS-compatible RMP control in KSTAR using n=2, +900 phasing RMP, although the ISS has been more vulnerable to mode-locking than typical KSTAR configuration. A detailed study of the KSTAR database (where RMP configuration of all the discharges belongs to n=1, +900 phasing) showed that the ELM-crash suppression phase in KSTAR is in the range of 0.2 < $\nu^*_e$ < 1.2 and $V_{tor}$> 40 km/s. During the ELM suppression phase, coexistence of filamentary mode and smaller scale turbulent eddies at pedestal with broad-range of wave number ($k_θ$<1.1 $cm^{-1}$ and frequency (f<100 kHz) is identified by ECE imaging (ECEI) and strong energy exchange of the filamentary and turbulent modes was measured. The bicoherence analysis of the edge harmonic oscillations (EHOs) at natural ELM-less mode shows that there is a strong nonlinear interaction between EHOs, and the nonlinear interaction of EHOs has a significant effect on the ELM structure and dynamics.
Cross-validation between the advanced diagnostics and the modeling provides new insight on the basic transport process at KSTAR. For example, in the recent MHD-quiescent KSTAR plasmas non-diffusive avalanche-like electron heat transport events are observed by the ECEI and these observations have been successfully reproduced by gyrokinetic simulations indicating the broad range of spatial scales up to the minor radius. In addition, various studies utilizing the KSTAR fluctuation diagnostics demonstrated the importance of the turbulence characteristics in plasma rotation and confinement. The extensive study of the intrinsic rotation in Ohmic plasmas found a clear link between the counter-current toroidal rotation direction and the quasi coherent mode (QCM) which is measured by the Microwave Imaging (MIR). The improved confinement in the low rotation experiment was correlated with the suppression of the broadband (~200 kHz) ECEI fluctuations, and Collective Thomson Scattering provides a detailed measurement on the high-k density turbulence which is suppressed during the typical LH transition. Finally, strong interaction between fast-ion and EP driven MHD mode was identified with Fast ion $D_α$ (FIDA) diagnostics.
KSTAR provided unique demonstration on the performance of symmetric multiple Shattered Pellet Injections (SPIs) which is the main strategy of ITER for disruption mitigation. It was shown successfully the current quench rate changes proportionally as the time difference varies from several percent to several tens of percent of the thermal quench (TQ) duration (1~2 ms) and it was demonstrated that peak density was increased twice with dual SPIs compared with a single SPI and energy can be radiated when multiple SPIs are injected simultaneously, as planned in ITER.
Lastly, the research plan in near term will be addressed with the machine upgrades. KSTAR will focus on the development of the DEMO/ITER relevant operational scenario, i.e., high-beta steady-state operation with benign MHD activities which will require robust plasma control in strong shaping, control of MHD modes and thorough analysis of the fundamental physics processes. In these regards, KSTAR upgrades will includes extensive NBI (off-axis, 6MW) & RF (Helicon CD, 4MW) heating & current drive capabilities and the installation of new tungsten divertors with active cooling.References:
$^1$G.S. Lee et al, Nucl. Fusion 40 575 (2000) 575
$^2$H. K. Park et al, Nucl. Fusion 59 (2019) 112020 (13pp)Speaker: Si-Woo Yoon (National Fusion Research Institute) -
The 2019-2020 scientific and technological programme exploits JET’s currently unique capabilities: Tritium handling and ITER-like wall (ILW: Be wall and W divertor). It is the culmination of years of concerted scientific and engineering work, with the ILW installation in 2010, improved diagnostic capabilities, now fully available, a major Neutral Beam Injection (NBI) upgrade providing record power in 2019 (P$_{NBI}$ up to 32MW), and the technical & procedural preparation for safe operation with tritium. Research along three complementary axes yielded a wealth of new results since last IAEA. Firstly, the JET plasma programme delivered scenarios suitable for high fusion power (P$_{FUS}$) and alpha particle ($\alpha$) physics in the coming D-T campaign (DTE2), with record sustained neutron rates, as well as plasmas for clarifying the impact of isotope mass on plasma core, edge and plasma-wall interactions, and for ITER pre-fusion power operation (e.g. L-H transition in He plasmas). The efficacy of the newly installed Shattered Pellet Injector (SPI) [ref1] for mitigating disruption forces and runaway electrons was demonstrated, informing ITER disruption management. Secondly, research on the consequences of long-term exposure to plasma in the ILW was completed, with emphasis on wall damage and fuel retention, and including analyses of wall materials and dust particles. This will help validate assumptions and codes for the design & operation of ITER and DEMO. Thirdly, the nuclear technology programme aiming to deliver the maximum technological return from operations with D, T and D-T [ref3] benefited from the high D-D neutron yield in 2019 (2.26x10$^{19}$n), securing new results for validating radiation transport and activation simulation codes, and nuclear data for ITER. Measuring systems are ready for collecting data in T and in D-T campaigns producing 14MeV neutrons.
Integrated scenarios preparation for high P$_{FUS}$ sustained for 5s (i.e. relevant to energy confinement times in JET) progressed significantly for the two routes investigated: ‘Baseline’ (q$_{95}$~3, I$_P$≥3MA, $\beta$$_N$<2) and ‘Hybrid’ (tailored q-profile, q$_{95}$~4.5, I$_P$ ≤2.7MA, $\beta$$_N$≥2.4). Peak neutron rate of 4.2x10$^{16}$n/s (2.7x10$^{16}$n/s averaged over 5s) are obtained simultaneously with tolerable divertor temperatures and controlled high/medium Z impurity for the full pulse duration in Baseline plasmas at 3.3T/3.5MA, with P$_{TOT}$=34MW (NBI and Ion Cyclotron Resonance Heating (ICRH)). Pellets help controlling the ELM frequency (f$_{ELM}$) needed for impurity flushing, with low total D$_2$ throughput for high confinement. Hybrid plasmas developed to 3.4T/2.3MA reached 4.8x10$^{16}$n/s but MHD avoidance and f$_{ELM}$ control must be optimised for improved, steady performance. The equivalent P$_{D-T}$ for these pulses is consistent with past predictions at same B$_T$, I$_P$, P$_{TOT}$ [ref4], giving confidence in the theory-based modelling. Further gains are likely with 40MW now reachable and higher I$_P$, with divertor heat loads controlled by strike-point sweeping, thus prospects for reaching the target (5x10$^{16}$n/s) are good. In these conditions P$_{D-T}$=11-16MW is predicted by theory-based physics models, with range due to uncertainties in the pedestal predictions and to whether isotope effects are included or not. Fast particle diagnostics, significantly improved since DTE1, can now detect small amount of $\alpha$’s, as shown in dedicated experiments making use of 3-ion ICRH scheme (D-(D$_{NBI}$)-3He) to create MeV range particles, with $\alpha$ (≈10$^{16}$s$^{-1}$) from D+$^3$He reactions. Simultaneous detection of He and hydrogen isotopes with an enhanced high resolution sub-divertor residual gas analyser, as planned for ITER, has been demonstrated.
Experiments and modelling preparing the T campaign. Observations that the impact of isotope on H-mode plasmas comes mainly from the pedestal [ref5] motivated recent gyrokinetic (GK) theoretical investigations of JET pedestals showing that the toroidal branch of the ETG instability can be driven at ion-scale poloidal wavelengths and may be responsible for significant inter-ELM pedestal heat transport. However, in some regimes, isotope effects on core plasma may also be important [ref4]. New experiments in D$_2$ and H$_2$ L-mode plasmas and related core GK modelling show that, in plasmas with a strong stabilizing effect of fast particles, differences in fast particle content with isotope mass may lead to strong deviations from the gyro-Bohm scaling of core transport. Recently developed ICRH-only H-mode plasmas (low input torque, dominant e-heating) show the same normalised confinement factor (H$_{98(y,2)}$) and T$_e$ profiles as their NBI-only counterpart at same PTOT, but n$_e$ profile for the NBI case is 50% higher due to NBI fuelling and possibly different particle transport. Work to clarify the impact of edge/divertor was performed. Experiments at 2MA/2.3T, low triangularity ($\delta$) demonstrated that changes in H$_{98(y,2)}$ and pedestal T$_e$ due to divertor configuration can be condensed into a single trend when mapped to the target T$_e$ as the main parameter governing recycling conditions rather than D$_2$ fuelling rate. A high performance neon seeded scenario (2.7T/2.5MA, high-δ shape) with edge conditions closer to ITER was developed. Neon seeding leads to significant increase (by ≈50%) in pedestal pressure and T$_e$ (from 0.4keV to 0.8keV) and in H$_{98(y,2)}$ (from 0.6 to 0.9) with mitigated divertor power loads. The well diagnosed discharges are used for validating physics-based SOL-edge modelling, increasing confidence in ITER divertor design basis and supporting deployment of neon over chemically reactive N$_2$ as seed gas in ITER.
JET disruption management programme is in two parts: 1) disruption avoidance based on improved termination techniques and on real-time detection of unhealthy plasmas with jump to controlled termination, causing significant reduction in disruption rate in baseline (60% to 20%) and hybrid plasmas (9%), and 2) disruption mitigation with SPI performed as part of a collaboration between ITER, US and Europe. After successful installation and commissioning, extensive experiments took place with the JET SPI demonstrating very good reliability. By varying the neon content in the SPI pellets, the disruption current quench time can be controlled efficiently in JET, scaling to the range required by ITER. High Z impurity SPI also demonstrated run-away electron suppression. Additionally, it was discovered that D$_2$ SPI applied to a high current run-away beam leads to benign impacts on the wall, suggesting a new potential solution for run-away electron control in ITER.
Plasma-facing components (PFC) long term exposure in ILW. Retrieval of PFC, wall probes (including test mirrors) and dust for ex-situ studies after three ILW campaigns provide deep insight into material erosion and deposition. Low mobilization of dust during in-vessel operations is shown. Emphasis was placed on material damage such as melting of the Be upper dump plates (UPD) and the identification of factors triggering this process. Comprehensive studies including imaging survey, morphology changes, mass loss and fuel inventory analysis on the most affected UDP tiles was performed. The undisputed reason for melting was unmitigated disruption events which tend to move the melt layers in the poloidal direction resulting in formation of upwards going waterfall-like structures of molten metal. The halo current is believed to provide the j x B force driving the melt layer motion. Global material migration results constitute a unique dataset for modelling and thus improved predictions for ITER.
[ref1] L. R. Baylor, et al., Nucl. Fusion 68 (2019) 211, [ref2] I. Jepu et al., Nucl. Fusion 59 (2019) 086009, [ref3] P. Batistoni et al, Fusion Engineering and Design Vol 109-11, 2016, [ref4] J. Garcia et al. Nucl. Fusion 59 (2019) 086047 [ref5] C. F. Maggi et al., Plasma Phys Control Fusion 60 (2018) 014045Speaker: Dr Joelle Mailloux (UKAEA) -
A. Bhattacharjee(a), B. Allen(g), C.-S. Chang(a), H. Chen(f), Y. Chen(f), J. Cheng(f), E. D’Azevedo(b), P. Davis(e), J. Dominski(a), M. Dorf(c), M. Dorr(c), S. Ethier(a), A. Friedman(c), K. Germaschewski(h), R. Hager(a), A. Hakim(a), G. Hammett(a), J. Hittinger(c), S. Janhunen(d), F. Jenko(d), S. Klasky(b), S. Ku(a), R. Kube(a), L. LoDestro(c), N. Mandell(a), G. Merlo(d), A. Mollen(a), M. Parashar(c), J. Parker(c), S. Parker(f), F. Poli(a), L. Ricketson(c), A. Scheinberg(a), M. Shepherd(i), A. Siegel(g), S. Sreepathi(b), B. Sturdevant(a), E. Suchyta(b), P. Trivedi(a), G. Wilkie(a), and M. Wolf(b)
(a) Princeton Plasma Physics Laboratory, Princeton, NJ, (b) Oak Ridge National Laboratory, Oak Ridge, TN, (c) Lawrence Livermore National Laboratory, Livermore, CA, (d) Institute for Fusion Studies, University of Texas at Austin, Austin, TX and Max Planck Institute for Plasma Physics, Garching, (e) Rutgers University, New Brunswick, NJ, (f) University of Colorado, Boulder, CO (g) Argonne National Laboratory, Chicago, IL
(h)University of New Hampshire, Durham, NH (i) Rensselaer Polytechnic Institute, Troy, NYWhole Device Modeling (WDM) is generally described as assembling physics models that provide an integrated simulation of the plasma. All components that describe a magnetic confinement device, from macroscopic equilibrium to micro-turbulence and control systems, are included in WDM, which describes the evolution of a plasma discharge from start-up to termination. Economical and safe operation of burning plasma devices requires predictive WDM with a confidence level established by validation and uncertainty quantification. Simulations covering the whole device, while certainly not a substitute for experiments, are much more cost-effective than building multiple billion-dollar facilities to test new ideas or concepts, similar to how aircraft manufacturers used simulations to reduce the number of physical wings they needed to build in designing superior aircraft (1).
The High-Fidelity Whole Device Model of Magnetically Confined Fusion Plasma is an application (hereafter referred to as WDMApp) (2,3) in the DOE Exascale Computing Project (ECP). The ECP is a DOE 413.3b project---the largest in the DOE Office of Science--- and is governed by the same rigorous rules of operation as major experimental facilities. The ultimate problem target of the project is the high-fidelity simulation of whole device burning plasmas applicable to an advanced tokamak regime (specifically, an ITER steady-state plasmas with ten-fold energy gain), integrating the effects of energetic particles, plasma-material interactions, heating, and current drive. The most important step in this project, and one that involves the highest risk, is the coupling of two existing, well-established, extreme-scale gyrokinetic codes – the GENE continuum code for the core plasma, and the XGC particle-in-cell (PIC) code for the boundary plasma. We have accomplished this challenging milestone for the first time in the magnetic fusion community. Fig. 1 demonstrates a coupled GENE-XGC in a WDMApp simulation for nonlinear ITG turbulence.
These developments would not be possible without the remarkable advancements in edge turbulence simulation codes for which XGC is an exemplar, along with COGENT and GKEYLL.
COGENT is a continuum gyrokinetic code for edge plasmas (4, 5). The code is distinguished by its use of fourth-order conservative discretization and mapped multiblock grid technology to handle the geometric complexity of a tokamak edge. It solves full-f gyrokinetic equations for an arbitrary number of plasma species, which can also be coupled to a set of lower-dimensionality fluid equations in cases where a reduced fluid model is adopted to describe electrons or neutrals. The code offers a number of collision models, ranging from the simple Krook operator to the fully nonlinear Fokker-Plank operator, and includes an ad-hoc anomalous transport model that can be utilized for the case of 4D axisymmetric transport calculations. Recent applications of the COGENT code to the analysis of cross-separatrix edge plasma properties include (a) 4D calculations, which demonstrate the values of radial electric field and toroidal rotation comparable to those observed on the DIII-D facility, and (b) 5D calculations of ITG turbulence, which elucidate the role of magnetic shear stabilization in the X-point region.
The GKEYLL project (6) is developing a continuum gyrokinetic capability that can evolve the electromagnetic gyrokinetic equations in the tokamak edge. The code uses a Hamiltonian form of the full-f equations, and for electromagnetic terms, uses a symplectic formulation. A novel version of a high-order discontinuous Galerkin scheme is used, ensuring that total energy (particles plus fields) is conserved by the spatial discretization. GKEYLL has performed the first fully nonlinear full-f continuum simulations of electromagnetic gyrokinetics in the scrape-off layer (SOL), including sheath boundary conditions. In Fig. 2 we show a snapshot of the turbulence profiles when statistical steady-state has been obtained. Shown are density and temperature contours near the midplane. Intermittent blob-like structures are seen ejected from the source region as they propagate outwards. Comparisons with electrostatic simulations show that the turbulence is larger amplitude and much more intermittent in the electromagnetic case. Also shown are magnetic field lines between the top and bottom divertor plate being stretched by blobs. Full details of the scheme and detailed description of the results are described in (7).
References
(1) “Case Study: Boeing Catches a lift with High Performance Computing,” Report by Council on Competitiveness, 2009
(2) G. Merlo, J. Dominski, A. Bhattacharjee, C.-S. Chang, F. Jenko, S. Ku, E. Lanti, and S. Parker, Phys. Plasmas 25, 062308 (2018)
(3) J. Dominski, S.-H. Ku, C.-S. Chang, J. Choi, E. Suchyta, S. Parker, S. Klasky, and A. Bhattacharjee, Phys. Plasmas 25, 072308 (2018)
(4) M. Dorf and M. Dorr, Contr. Plasma Phys. 58, 434 (2018)
(5) M. Dorf and M. Dorr, Contr. Plasma Phys. (2020); available online DOI: 10.1002/ctpp.201900113
(6) A. Hakim, N. Mandell, T. Barnard, M. Francisquez, G. Hammett, and E. Shi, Continuum Electromagnetic Gyrokinetic Simulations of Turbulence in the Tokamak Scrape-Off Layer and Laboratory Devices, submitted to Phys. Plasmas. (2020)
(7) N. Mandell, A. Hakim, G. Hammett, and M. Francisquez, J. Plasma Physics 86, 149 (2020)Speaker: Mr Amitava Bhattacharjee -
A new era of predictive integrated modeling has begun. The successful validation of theory-based models of transport, MHD stability, heating and current drive, with tokamak measurements over the last 20 years, has laid the foundation for a new era where these models can be routinely used in a "predict first" approach to design and predict the outcomes of experiments on tokamaks today. The capability to predict the plasma confinement and core profiles with a quantified uncertainty, based on a multi-machine, international, database of experience, will provide confidence that a proposed discharge will remain within the operational limits of the tokamak. Developing this predictive capability for the first generation of burning plasma devices, beginning with ITER, and progressing to tokamak demonstration reactors, is a critical mission of fusion energy research. Major advances have been made using this predict first methodology. Extensive predictive modeling has informed the planning for the JET D-T campaign. This includes integrated modeling of JET hybrid regimes with newly upgraded heating sources, for various concentrations of deuterium D and tritium T. The self-consistent profiles of tungsten, ion and electron temperature, toroidal rotation and densities, have been predicted using theory-based turbulence and neoclassical transport models. The EPED model predicts it is possible to access the, higher pressure, super-H pedestal regime for JET achievable shapes. This prediction has been confirmed with DIII-D experiments. Super-H experiments on JET are planned. A new high accuracy neural network fit to the QuaLiKiz transport model has been completed, opening the way to time dependent predictions, at near real time speed, of complete tokamak discharges. Neural network fits to the TGLF and Multi-Mode models are progressing. The EAST tokamak is using predictive modeling to optimize the high bootstrap fraction regime for fully non-inductive operation and to plan future upgrades of power and current drive systems. A new integrated modeling workflow called TRIASSIC is being developed and tested on the KSTAR tokamak. Predictive modeling of CFETR is informing the design activity. ITER is using predictive modeling to simulate phases of the experimental operations plan. An overview of several of these recent advances will be presented, providing the integrated modeling foundations of experimental successes, as well as progress towards the goal of integrated predictive modeling for experimental design. Two examples, selected from the many advances in the prediction of tokamak experiments, are summarized in this synopsis.
1st example: The fast response of cold pulses due to impurity injection in tokamaks, with an inversion of the inward electron temperature pulse from decrease to increase, has long been argued to be inconsistent with a local transport paradigm [1]. The first demonstration that the cold pulse temperature response could be captured by a local turbulence transport model (TGLF [2]) was performed for the C-MOD tokamak [3]. Only electron and ion temperatures were predicted in these cases, with the density profile being evolved in a prescribed way. It was found that the inversion of the electron temperature pulse from decrease to increase was caused by the stabilization of the trapped electron mode (TEM) by the flattening of the electron density profile. In discharges where the TEM mode was not dominant there was no inversion in agreement with experiment. The transport model was then used in predict first method to simulate the cold pulse response [4] in the DIII-D tokamak. The very fast, high spatial resolution, density profile data on DIII-D confirmed the speed of the prescribed density response and the electron temperature response predictions were confirmed. The final step was to prove that the TGLF model could predict the fast density response to the impurity injection. This required adding the injected impurity density to the transport modeling. This integrated modeling was performed for experiments on the ASDEX Upgrade tokamak [5]. The predicted electron temperature response is compared with data in
. It was found that the destabilization of the ion temperature gradient mode (ITG) by the transiently hollow impurity profile increased the speed of propagation of the electron density pulse into the core. Thus, the speed of the combined electron, ion, and impurity, temperature and density pulses were accurately modeled and new physics insights were discovered. This is a convincing proof that local turbulence transport can account for the paradoxical cold pulse phenomenon.
2nd example: The new upgrades to off-axis NBI current drive capability on DIII-D were preceded by state of the art integrated modeling [6] illustrated by the advanced tokamak predictions in
. The profiles in Fig. 2 are a steady state self-consistent solution of the pedestal structure (height and width), core transport, MHD equilibrium and heating and current drive using validated theory-based models. An iterative high performance workflow IPS-FASTRAN was developed to find the integrated optimum solution [6]. Well validated theory-based models for MHD equilibrium (EFIT) and stability (DCON), turbulent transport (TGLF), pedestal structure (EPED1), neutral beam heating and current drive (NUBEAM) and electron cyclotron heating and current drive (TORAY-GA) were integrated. The IPS-FASTRAN modeling predictions have been confirmed with experiments showing good agreement that will be reported at the FEC 2020 conference. Verification of the accuracy of these predict first method simulations are a valuable test of the new capabilities. The same integrated modeling workflow is being used in the design of the CFETR, CAT Fusion Pilot Plant and SPARC tokamaks and to predict ITER plasmas.
This work was supported by the US Department of Energy under DE-FG02-95ER54309, DE-FC02-04ER54698, DE-SC0019736
[1] K. W. Gentle, W. L. Rowan, R. V. Bravenec, G. Cima, T. P. Crowley, et al., Phys. Rev. Lett. 74 (1995) 3620
[2] G. M. Staebler, J. E. Kinsey, and R. E. Waltz, Phys. Plasmas 14, (2007) 055909.
[3] P. Rodriguez-Fernandez, A. E. White, N. T. Howard, B. A. Grierson, G. M. Staebler, et al., Phys. Rev. Lett. 120 (2018) 075001.
[4] P. Rodriguez-Fernandez, A. E. White, N. T. Howard, B. A. Grierson, L. Zheng, et al., Phys. Plasmas 26, (2019) 062503.
[5] C. Angioni, E. Fable, F. Ryter, P. Rodriguez-Fernandez, T. Putterich, and the ASDEX Upgrade team, Nuclear Fusion 59 (2019) 106007.
[6] J. M. Park, J. R. Ferron, C. T. Holcomb, R. J. Buttery, W. M. Solomon, et al., Phys. Plasmas 25 (2018) 012506.Disclaimer-This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights. Reference herein to any specific commercial product, process, or service by trade name, trademark, manufacturer, or otherwise, does not necessarily constitute or imply its endorsement, recommendation, or favoring by the United States Government or any agency thereof. The views and opinions of authors expressed herein do not necessarily state or reflect those of the United States Government or any agency thereof.
Speaker: Gary Staebler (General Atomics) -
Since the last IAEA-FEC, the EAST research programme has been, in support of ITER and CFETR, focused on development of the long-pulse steady-state (fully non-inductive) high beta H-mode scenario with active control of stationary and transient divertor heat and particle fluxes $^{[1]}$. The operational domain of the steady-state H-mode plasma scenario on EAST has been significantly extended with the ITER-like configuration, plasma control and heating schemes. Several important milestones on the developments of the high beta H-mode scenario and its related key physics and technologies have been achieved.
-
A minute time scale long-pulse steady-state high beta H-mode discharge (shown in figure 1) with the major normalized plasma parameters similar to the design of the CFETR Phase-III 1GW fusion power operation scenario ($\beta_P$ ~ 2.0, $\beta_N$ ~ 1.6, $f_{bs}$ ~ 50%, $H_{98(y2)}$ > 1.3 at $q_{95}$ = 6.5~7.5) has been successfully established and sustained by pure RF heating on EAST with the ITER-like tungsten divertor $^{[2]}$, as shown in figure 1. The important feature of this high beta H-mode plasma scenario is that, due to the stabilization effect of the Shafranov shift on the plasma turbulence, a higher $\beta_P$ results in a better plasma confinement as shown in figure 2. Further simulation suggested that a high density gradient promotes the ITB formation in high $\beta_P$ plasmas, which might further benefit the development of this high beta plasma scenario towards a high density regime.
-
Active control of divertor radiation has been successfully integrated into the high beta H-mode ($\beta_P$ ~ 2.5, $\beta_N$ ~ 2.0, $f_{bs}$ ~ 50%) plasma scenario without a degradation of the plasma confinement ($H_{98(y2)}$ > 1.2) at high density ($n_e$/$n_{GW}$ ~ 0.7) and moderate edge safety factor ($q_{95}$ ~ 6.7) $^{[2]}$. The peak heat flux on the tungsten divertor was reduced by ~30% with active impurity seeding of a mixture of 50% neon and 50% deuterium. The high-Z impurity concentration in the plasma core has been well controlled in a low level by applying the on-axis ECRH and reducing the fast ion losses through beam energy optimization.
-
The grassy-ELM regime has been extended to the normalized parameter space designed for the CFETR 1GW fusion power operation scenario. This regime exhibits good compatibility with high $f_{bs}$ and fully non-inductive operation, being characterized by a low pedestal density gradient and a wide pedestal, which prevents large-ELM crashes due to an expansion of the peeling-ballooning boundary after the initial pedestal collapse, indicated by BOUT++ nonlinear simulations. High separatrix density makes this regime especially suitable for operation with divertor detachment. Several feedback control schemes have been developed to achieve sustained detachment with good core confinement $^{[3]}$. This includes control of total radiation power, target electron temperature, and particle flux measured by divertor Langmuir probes or a combination of the control of target electron temperature and AXUV radiation near the X point. Integration of these detachment feedback control schemes with the grassy-ELM regimes and the high $\beta_P$ scenario has been demonstrated with neon seeding, which provides an integrated high beta scenario applicable to long-pulse operation $^{[4]}$.
-
ELM suppression has been achieved using various different methods on EAST. Full suppression of ELMs has been demonstrated, for the first time, for ITER-like low torque injection plasmas by using n=4 resonant magnetic perturbations (RMPs) $^{[5]}$. A moderate reduction (~5%) of the energy confinement has been observed together with significant reduction of both the plasma density(20%)and the Tungsten concentration (a factor of 2) during ELM suppression. The ELM suppression window agrees well with the prediction by MARS-F modelling. Robust ELM suppression by Boron powder injection without confinement degradation or even with confinement improvement has been achieved in a wide parameter range. EHO-like edge coherent modes were excited during the ELM suppression phase by Boron powder injection. In addition, simulation results from BOUT++ confirm that both the helical current filaments (HCF) driven by lower-hybrid waves (LHWs) and RF sheath effects on the ICRF antenna contribute to ELM suppression.
-
A flowing liquid lithium (FLiLi) limiter plate has been successfully assembled and tested in EAST for the first time. This plate was designed based on the concept of liquid metal infused trenches (LiMIT), which is using thermoelectric MHD to drive liquid Li flow along the surface channels. The preliminary results show that with the increase of Li flow rate, the fuel particle recycling is gradually reduced, and the plasma performance is slightly improved. There was no obvious Li burst and limiter damage even at a high injection power of 5.5 MW. In addition, ELM mitigation was observed with FLiLi operation.
-
For the first time, EAST has been operated with helium to support the ITER needs. The H-mode power threshold in a helium plasma is found to be 1.2-2.2 times higher than the scaling law in deuterium plasma with pure RF-heating $^{[6]}$. The $C_{He}$ plays a crucial role in determining the energy confinement and pedestal characteristics in helium H-mode plasmas. Divertor detachment is more difficult to achieve in He than in D. The control of divertor heat loads and W sources is achieved by RMP along with neon impurity seeding. The inter-ELM W erosion rate in He is about 3 times that in D with similar divertor conditions, while the intra-ELM W sputtering source shows a strong positive correlation with the ELM frequency $^{[7]}$.
-
A new lower tungsten divertor with a higher closure has been designed and to be installed on EAST in 2020. Several key subsystems, including the heating & current drive, cryogenic, plasma control and diagnostics will be upgraded accordingly for achieving the next milestones i) >400s long-pulse H-mode operation with ~50% bootstrap current fraction, and ii) demonstration of power exhaust at ~10 MW power injection for >100s.
Reference:
1) B. N. Wan et al 2019 Nucl. Fusion 59 112003
2) X. Z. Gong et al 2020 this conference
3) L. Wang et al 2020 this conference
4) G. S. Xu et al 2020 this conference
5) Y. W. Sun et al 2020 this conference
6) B. Zhang et al 2020 this conference
7) R. Ding et al 2020 this conference
Speaker: Baonian Wan (Institute of Plasma Physics, Chinese Academy of Sciences) -
-
Construction of JT-60SA is progressing on schedule towards completion of assembly in March 2020 and the first plasma in September 2020. As of January 2020, manufacture and assembly of all the main tokamak components have been successfully completed satisfying technical requirements including functional performances and dimensional accuracies. Development of plasma actuators and diagnostics is also going well such as achievement of long sustainment of high energy intense negative ion beam. Commissioning of the power supply and the cryoplant has also satisfied requirements. Development of all the control systems and evaluation procedures of tokamak operation has been completed towards the Integrated Commissioning starting in April 2020, and plasma operation scenarios in the first plasma phase have been established. Unique importance of JT-60SA for H-mode and high-beta steady-state plasma research has been confirmed using advanced integrated modellings. These experiences of assembly, integrated commissioning and plasma operation of JT-60SA contribute to ITER risk mitigation and efficient implementation of ITER operation.
Introduction
The JT-60SA (R/a =3m/1.2m, Ip-max =5.5MA, heating power = 41MW x 100s) project [ref.1] was initiated in 2007 under the framework of the Broader Approach agreement by EU and Japan for early realization of fusion energy by conducting supportive and complementary works for ITER towards DEMO. Construction of JT-60SA is progressing successfully towards completion of assembly in Mar. 2020 and the first plasma in Sep. 2020 by the very close collaboration between QST in Japan, F4E in Europe, EU Voluntary Contributors and EUROfusion. The JT-60SA Research Plan [ref.2] covering its machine lifetime of ~ 20 years coordinated with ITER and DEMO schedules has been established with variety of plasma prediction using integrated modeling codes [ref.3]. Recently in Nov. 2019, a new collaboration arrangement between ITER and JT-60SA was signed which covers assembly, integrated commissioning and operation/experiments for finalization of ITER component design, risk mitigation and efficient implementation of ITER operation.Tokamak Construction
After the last IAEA FEC [ref.1], manufacture of all remaining tokamak components has been completed successfully including, superconducting Centre Solenoid (CS), thermal shields, Cryostat Top Lid, Cryolines, etc. As of Dec. 2019, the closure of the vacuum vessel has been accomplished, and the tokamak has been covered by the Cryostat Vessel Body (Fig.1). All the tokamak components have been assembled with excellent dimensional accuracy of ±1mm thanks to careful and smooth positioning using specially designed jigs, high accuracy measurement by Laser trackers, and fine adjustment utilizing sims. The magnetic field error is now expected below 10-4 Bt as designed. Commissioning operation of all large power supply systems, the Quench Protection Cirquit, the Switching Network Units and Super Conducting Magnet Power Supplies, has also been progressed with few residual commissioning activities still ongoing. The commissioning operation of the Cryoplant (equivalent refrigeration capacity of 9 kW at 4.4K) has also been successfully completed by satisfying the required performances.Plasma Hating Systems
For the heating systems, Positive-ion source NBs (85keV, 100sec, 20MW by 12 unit), Negative-ion source NBs (500keV, 100sec, 10MW by 2 units), ECH with multiple frequency Gyrotron (110GHz & 138GHz for 100s and 82GHz for 1 sec) and movable launchers, R&D have been steadily progressing and the targets of their development have been achieved. In particular, high energy intense hydrogen negative ion beams with 500 keV, 154 A/m2 for 118 s, which exceeds the requirement for JT-60SA, has been demonstrated by using a semi- cylindrical negative ion source with a three-stage accelerator. This result was realized by integration of i) stable voltage insulation by suppression of arching, ii) precise beam control and iii) stable negative ion production by maintaining the temperature balance in the negative ion source.
Integrated Commissioning and Control Systems
From April 2020 to Feb. 2021, the integrated commissioning is planned with the first plasma in Sep. 2020 and subsequent 5 months of machine commissioning with plasmas (‘the first plasma phase’). In this phase, the goal of plasma operation is to demonstrate equilibrium controllability of MA-class (<2.5MA) diverted plasmas with the full performance superconducting coil systems, 1.5MW ECH and upper divertor. For such tokamak operations, the Supervisory Control System and Data Acquisition System (SCSDAS) has been developed having the roles of (a) plant monitoring and machine state management, (b) discharge sequence management, (c)real-time plasma control, (d) device protection and human safety, (e) data storage, archive, and database management etc. For plasma controls, we have simulated operation scenarios of the first plasma phase with a newly developed advanced codes with control logics, such as pre-magnetic optimization scheme, plasma equilibrium control with iso-flux control method, control gain optimization method, and strategies for accessing stable operational regimes. Figure 2 shows the discharge scenario at Ip=2.5MA. EC wall cleaning operations and EC-assisted breakdown are also explored with optimized EC injection and toroidal / horizontal field. These results in the JT-60SA first plasma phase will contribute to highly valued subjects in ITER first plasma/subsequent operations.
Scenario Development for ITER and DEMO and Risk Mitigation of ITER
After machine enhancements in 2021-2022, physics experiments will start in 2023 using in-vessel coils, particle fueling and pumping with lower divertor, enhanced diagnostics and high heating power of 26MW at Ip up to 5.5 MA. Toward this phase, variety of predictions of H-mode and high-beta steady-state plasmas covering divertor-SOL-Pedestal-Core have been progressing using advanced integrated modellings including newly-developed globally optimized steady-state transport solver GOTRESS coupled with turbulence models and pedestal models, Gyrokinetic theory based neural-network transport modeling DeKANIS, etc. These studies have confirmed the unique and important characteristics of JT-60SA (highly-shaped, high-beta, 500keV high energy ions, electron heating, controllable rotation etc.) for study of fusion plasma physics such as impacts of fast ions and plasma shape on microturbulence. As for operation scenario development of high-beta steady-state with controlled divertor heat load, an important result has been achieved using the integrated divertor code SONIC upgraded to treat multiple impurity species simultaneously. The result has shown that ‘mixture-seeding of Ar with small amount of Ne’ can keep the peak heat load below allowable 10MW/m2 together with smaller Ar concentration in the SOL and core plasmas than an Ar-only case. These studies have also confirmed significant roles of JT-60SA for ITER risk mitigation (disruption and ELM mitigation) including magnetic perturbation effect on both transient and stationary heat load, vertical displacement event, plasma response to massive gas injection, pedestal and ELM stability and control with Pellet and RMP.[ref.1] P. Barabaschi, Y. Kamada, H. Shirai and JT-60SA Intrgrated Project Team,
Nucl. Fusion 59 (2019) 112005.
[ref.2] JT-60SA Research Plan - Version 4.0, Sept. 2018,
http://www.jt60sa.org/pdfs/JT-60SA_Res_Plan.pdf
[ref.3] G. Giruzzi, M. Yoshida et al., Plasma Phys. Control. Fusion, 62 (2020) 014009.Speaker: Yutaka Kamada (National Institutes for Quantum and Radiological Science and Technology) -
DIII-D physics research addresses critical challenges for operation of ITER and the next generation of fusion energy devices through a focus on innovations to provide solutions for high performance long pulse operation, development of scenarios integrating high performance core and boundary plasmas, and fundamental plasma science and model validation. Substantial increases in off-axis current drive efficiency from an innovative top launch system for EC power (Fig. 1), and in pressure broadening for Alfven eigenmode control from a co-/counter-Ip steerable off-axis neutral beam, both improve the prospects for optimization of future long pulse/steady state high performance tokamak operation. A high beta-p optimized-core scenario with an internal transport barrier that projects nearly to Q=10 in ITER at 9 MA was coupled to a detached divertor, and a Super H-mode optimized-pedestal scenario with co-Ip beam injection (Fig. 2), was coupled to a radiative divertor. Fundamental studies into the evolution of the pedestal pressure profile, and electron vs. ion heat flux, measuring both density and magnetic field fluctuations, validate predictive models of pedestal recovery after ELMs (Fig. 3).
Link to High Resolution Figures 1, 2, 3
The achievement of more than double the off-axis ECCD efficiency using top launch geometry compared with conventional low field side (LFS) launch, as predicted by quasi-linear Fokker-Planck simulations, is due to the longer absorption path for the EC waves which also interact with higher v|| electrons that suffer fewer trapping effects than outside launch. In addition, the new unique co-/counter-Ip steerable off-axis neutral beam broadens the energetic particle (EP) pressure profile and reduces Alfven eigenmode (AE) drive in scenarios with both high toroidal rotation and those with net zero average input torque. New EP measurements show a beam current threshold for Compressional AEs, insensitivity of Beta-induced Acoustic AEs to fast beam ions, and resolution of phase-space flows caused by AEs, from first-of-a-kind Ion Cyclotron Emission (ICE) and Imaging Neutral Particle Analyzer (INPA) data.
Studies of high current runaway electron (RE) beams reveals excitation of current-driven (low safety factor) kink instabilities that promptly terminate the RE beam on an Alfvenic time-scale, offering an unexpected alternate pathway to RE beam mitigation without collisional dissipation. Newly developed real-time stability boundary proximity control and neural-net-based Vertical Displacement Event (VDE) growth-rate calculations are shown to prevent VDEs. The effectiveness of emergency shutdown and disruption prevention tools projects to at least 50% of ITER disruptions being delayed until normalized-Ip is at safe levels, and demonstration of a novel technique for healing flux surface with 3D fields shows promise for providing current quench (CQ) control. Single and multiple Shattered Pellet Injection particle assimilation rates and current quench (CQ) densities are shown to be predictable from 0-D simulations and empirical scaling laws.
Several core-edge integration scenarios demonstrate coupling of a high performance core and radiative divertor operation for target heat flux control. High density and stored energy plasmas with Super H-mode edge pedestals were made both in a lower single null shape accessible by JET and in a higher triangularity near double null shape coupled to a radiative divertor for target heat flux control using nitrogen injection in a core-edge integrated scenario. High-performance plasma with high poloidal beta, large Shafranov shift, and Te and Ti internal transport barriers coupled to a detached divertor with active feedback-controlled Nitrogen puffing also demonstrated integration of core-edge solutions. A high performance hybrid core demonstrated compatibility with radiative divertor operation using Neon or Argon gas injection. Core impurity peaking in the hybrid was substantially reduced using near-axis electron cyclotron heating.
The ability to predict the impurity seeding needed for divertor dissipation has advanced through new capability for measuring charge-state resolved densities of impurity species in the divertor. Also electric drifts in detached divertors with convection dominated heat transport lead to expanded radiative volume. Using these advances, SOLPS-ITER simulations show the synergy between SOL drifts and the SAS divertor geometry for achieving lower density detachment. Modeling of intra-ELM tungsten gross erosion with an analytic Free-Streaming plus Recycling Model is now validated in ITER-relevant mitigated-ELM regimes using pellet pacing and RMPs. SOL tungsten transport in plasmas with both BT directions is consistent with strong entrainment in SOL flows and ExB drift effects.
Advances in pedestal physics through new measurements of density and internal magnetic fluctuations suggest a possible role for micro-tearing and trapped electron modes in DIII-D pedestal transport. Main ion CER measurements indicate ion heat flux is anomalous at low collisionality and transitions to near neoclassical levels at high collisionality. Plasma rotation scans, and both new non-linear analytic theory and 2-fluid code simulations, confirm that ELM suppression by RMPs requires near zero ExB velocity at the top of the pedestal, and achieving suppression appears to be closely linked to a high field side plasma response. The wide pedestal QH-mode regime was obtained with zero input beam torque, electron heating, and LSN shape, consistent with requirements for ITER.
Recent fundamental research on L-H mode power threshold physics shows that turbulence driven shear flow through Reynolds stress and the coexistence of modes associated with various instabilities can lower the L-H power threshold across multiple parameters: eg. q95 and ion grad-B drift direction. Application of RMPs raises turbulence decorrelation rates and reduces Reynolds stress driven flow and flow shear, hence increasing the L-H power threshold. Finally, plasmas with negative triangularity show weak power degradation of H-mode level core confinement while maintaining an L-mode-like edge without ELMs.
In 2020 and beyond DIII-D will install additional tools for optimizing tokamak operation through current and heating profile control using a low field side 1 MW helicon high harmonic fast wave CD system, a unique high field side Lower Hybrid CD system, increased ECH power, and coupling to boundary advances using a new high power closed divertor and a wall insertion test station. Experiments will continue the optimized coupling of high performance core and high power density divertor solutions.
This work was supported in part by the US DOE under contracts DE-FC02-04ER54698 and DE-AC52-07NA27344Speaker: Max Fenstermacher (LLNL @ DIII-D) -
We present here recent highlights from Wendelstein 7-X (W7-X), the most advanced and largest stellarator in the world, in particular stable detachment with good particle exhaust, low impurity content, and energy confinement times exceeding 100 ms, maintained for tens of seconds, as well as proof that the reduction of neoclassical transport through magnetic field optimization is successful. W7-X, which has a magnetic field strength of 2.5 T and a plasma volume of 30 m$^3$, started operation in 2015 [1-5]. Following the installation of a full set of in-vessel components, in particular 10 passively cooled fine-grain graphite test divertor units, it was operated again in 2017 and 2018. Plasma pulses up to 100 s were successfully sustained [6], despite the lack of active cooling. Stable and complete detachment was achieved routinely. The pumping efficiency was initially relatively low [7-9] but it significantly improved later. Detachment with high pumping efficiency was achieved for up to 28 seconds at a heating power of 5 MW with a very low impurity content [10] (Figure 1), indicating control of divertor-heat-flux, plasma density, and impurity content, and giving confidence for reaching the foreseen high-performance, quasi-steady-state (30 minutes) discharges in the future [11]. The performance of the W7-X divertor, and the behavior and parameters of the edge- and scrape-off-layer plasma are now understood in quite some detail [eg. 12], thanks to measurements from a suite of diagnostics [eg. 13-17].
The earlier-reported stellarator triple-product record discharge [18] has now been shown to provide proof that the optimization for reduced neoclassical transport in W7-X was successful, Figure 2 [19]: The high temperature (appr. 3.5 keV for both ions and electrons in the center) and high hydrogen ion density (appr. 7x1019 m-3 in the core) were achieved with 5 MW of heating, and an energy confinement time of 0.22 s corresponding to about 1.4 times the energy confinement time expected from the ISS04 stellarator scaling [20]. For other, less optimized stellarators scaled to the W7-X size and magnetic field strength, similar plasma temperature and density profiles would have required significantly higher heating power to balance neoclassical transport, in particular in the mid-radius (strong gradient) region.
A number of discharges with similar performance to the triple-product-record discharge have since been achieved. These are generally characterized by core density peaking, and a reduction of turbulent density fluctuations. Without such turbulence reduction, the central ion temperature appears to be clamped to appr. 2 keV [21]. These findings are consistent with W7-X transport usually being dominated by ITG turbulence, but stabilized by strong density gradients in a so-called stability valley [22], as exemplified in Figure 3 for the W7-X standard configuration.
During turbulence-dominated phases, impurity confinement times are low (of order the energy confinement time) and no impurity accumulation is seen, but they can be very large if the turbulence is suppressed, and this then leads to impurity accumulation [23]. Recent findings from the Large Helical Device (LHD) show that ITG-dominated discharges readily mix hydrogen isotopes, whereas electron-scale (trapped-electron mode) turbulence does not [24]. It is tentatively concluded that a non-negligible amount of ITG turbulence is beneficial for impurity control as well as for fuel (isotope) exchange and helium exhaust in a stellarator fusion reactor, whereas too much ITG turbulence could potentially clamp the ion temperature below the burn point. These and other recent results [see eg. 25-31] will be put into the context of future goals for the W7-X, the world stellarator program, and the magnetic confinement fusion program in general.
References
1 T. Klinger et al, Plasma Phys. Controlled Fusion 59(1) 014018 (2017)
2 H.-S. Bosch et al, Nuclear Fusion 57, 116015 (2017)
3 R. C. Wolf et al, Nuclear Fusion 57 102020 (2017)
[4] T. Sunn Pedersen et al, Physics of Plasmas 24 055503 (2017)
[5] A. Dinklage et al, Nature Physics (2018) https://doi.org/10.1038/s41567-018-0141-9
[6] T. Klinger et al, Nuclear Fusion 59 112004 (2019)
[7] T. Sunn Pedersen et al, Nuclear Fusion 59 096014 (2019)
[8] D. Zhang et al, Phys. Rev. Lett. 123, 025002 (2019)
[9] D. Zhang et al, this conference (2020)
[10] M. Jakubowski et al, submitted to Phys. Rev. Lett. (2020)
[11] M. Jakubowski et al, this conference (2020)
[12] F. Reimold et al, this conference (2020)
[13] V. Perseo et al, Nuclear Fusion 59 124003 (2019)
[14] V. Perseo et al, this conference (2020)
[15] T. Barbui et al, JINST 14 C07014 (2019)
[16] C. Killer et al, Plasma Phys. Control. Fusion 61 125014 (2019)
[17] C. Killer et al, this conference (2020)
[18] T. Sunn Pedersen et al, Plasma Phys. Control. Fusion 61 014035 (2019)
[19] C. Beidler et al, in preparation (2020)
[20] H. Yamada et al Nuclear Fusion 45 1684 (2005)
[21] M. Beurskens et al, this conference (2020)
[22] J. A. Alcusón et al, Plasma Phys. Control. Fusion 62 035005 (2020)
[23] A. Langenberg et al, this conference (2020)
[24] K. Ida et al, Phys. Rev. Lett. 124, 025002 (2020)
[25] J. Geiger et al, this conference (2020)
[26] Y. Feng et al, this conference (2020)
[27] G. Fuchert et al, this conference (2020)
[28] K. Aleynikova et al, this conference (2020)
[29] H. Laqua et al, this conference (2020)
[30] S. Lazerson et al, this conference (2020)
[31] A. Dinklage et al, this conference (2020)Speaker: Prof. Thomas Sunn Pedersen (Max Planck Institute for Plasma Physics) -
Experiments on ST40 towards burning plasma conditions
M. P. Gryaznevich for TE.Ltd team
Tokamak Energy Ltd, 173 Brook Drive, Milton Park, Abingdon, OX14 4SD, UK
e-mail: mikhail.gryaznevich@tokamakenergy.co.ukSpherical Tokamak (ST) path to Fusion has been proposed in R Stambaugh et al, Fus. Tech. 33 (1998) 1, and experiments on STs have already demonstrated feasibility of this approach. Advances in High Temperature Superconductor (HTS) technology (M Gryaznevich et al, Fus. Eng. & Design 88 (2013) 1593) allows significant increase in the Toroidal Field (TF) which was found to improve confinement in STs. The combination of the high beta, which has been achieved in STs, and high TF that can be produced by HTS TF magnets, opens a path to lower-volume fusion reactors, in accordance with the fusion power scaling ~ beta^2Bt^4V. High field spherical tokamak ST40 (design parameters: R=0.4-0.6m, R/a=1.6-1.8, Ipl=2MA, Bt=3T, k=2.5, pulse~1-2sec, 2MW NBI, 2MW ECRH/EBW, DD and DT operations) is the first prototype on this path and is now operating, Fig.1.
Plasma current > 0.5MA at 2T TF, electron and ion temperatures in a several-keV range produced using merging-compression formation, solenoid-assisted ramp-up and 1MW of 25kV NBH, and densities up to 2x10^20m^-3 have been achieved in the first experimental campaigns in 2018-2020. At the flat-top, measured Ti increases with TF, in agreement with observations on other STs. However, on ST40, at TF > 1 – 1.1T we observe sharp increase in Te, Ti and W(EFIT), Fig.2.
TF Cu magnet in ST40 and some PF coils are LN2 cooled and research is on-going on development of full-HTS magnets. HTS prototype magnet with 24.4 T (at 21 deg K) has been built, Fig.3, and now we are planning to increase the field. LN2 cooling of present Cu magnets, installation of the second 1MW 50kV beam and upgrades of power supplies are on-going and will allow increase of TF to 3T and the pulse duration from ~0.3sec (at present) to 1-2sec.
Experiments are carried out to study transport properties in ST at higher TF, higher heating power (up to 4MW), low collisionality, in aim to bring plasma parameters close to burning conditions. Transport simulations with ASTRA, NUBEAM and TSC codes have been performed to model ST40 parameters and to support the physics basis of the compact high field ST path to Fusion. We show that high confinement regimes with just collisional (neoclassical) transport can be expected even when only ohmically heated. In an auxiliary heating regime, we find a hot ion mode with Ti in the 10keV range to be achievable in ST40 with as low as 1MW of absorbed power, Fig.4.
Issues connected with specific features of the high field ST are discussed, i.e. limitations of applicability of confinement scalings for prediction of performance of ST40. However, we show that if the performance achieved on other spherical tokamaks can be extended to ST40 conditions, up to 1 MW of Fusion power can be expected in DT operations. Studies of fast ions and alpha particle transport, heating and current drive, torque deposition and momentum transport have been performed using ASCOT, NUBEAM, Monte Carlo code NFREYA and the Fokker - Planck code NFIFPC. Different NBI energies and launch geometries have been studied and optimised. The confinement of thermal alphas in ST40 3T/2MA scenario is studied with full orbit following (which is necessary because of the large, compared to the plasma size, alpha particle gyro radius). The first orbit losses are seen to be above 60% even in the high-performance scenario illustrating that the alpha confinement in a small device is very difficult even at the highest available fields and plasma currents. However, DT experiments on ST40 will provide useful information for verification of such simulations.
We are intensively working on the design of our next tokamak, ST-F1, with plasma volume ~0.5 of JET. This device is aimed to demonstrate Q>3, will have HTS magnets, tritium blanket, and the goal is to build it by 2025.
Speaker: Mr Mikhail Gryaznevich -
Operating a full tungsten actively cooled tokamak:
overview of WEST first phase of operation
J. Bucalossi and the WEST Team (http://west.cea.fr/WESTteam)
CEA, IRFM, F-13108 St-Paul-Lez-Durance, France.E-mail: jerome.bucalossi@cea.fr
WEST is a MA class superconducting, actively cooled, full tungsten (W) tokamak. Equipped with two up-down symmetric divertors, it operates at 3.7T, up to 1MA, with a plasma volume of 15 m3 and an aspect ratio between 5 and 6. CW RF power is installed: up to 9 MW of ICRH power and 7 MW of LHCD.
In support of ITER operation and DEMO conceptual activities, WEST aims at power exhaust studies in long and steady-state pulses, in various divertor configurations (LSN, USN, DN), in a full W environment. The lower divertor, partially made of ITER-grade Plasma Facing Units (PFU) complemented with inertially cooled W-coated elements in phase 1 (2017-2019, see Fig. 1), will be replaced by a complete ITER-grade divertor in phase 2, starting autumn 2020. This paper reports on the main findings from WEST phase 1, in terms of operational domain, plasma performance achieved, and first tests of the ITER grade PFU.
Initial phase of operation was hindered by the production of runaway electron beams, when the ohmic current failed to rise quickly enough. Interestingly, start-up runaway electrons have been avoided by reducing the prefill pressure. Severe damages on PFC were observed [Diez] and one runaway beam impact induced radiation even quenched one superconducting coil [Reux].
Additionally to 200°C baking and glow discharge cleaning, boronizations have been performed in the second campaign leading to long lasting improved breakdown conditions and higher density operational domain.
Apart from the few post-boronization pulses, the fraction of radiated power in LHCD, ICRH or LHCD+ICRH pulses remains high, around 50%. Tungsten is, in most cases, the major radiating species [Goniche]. Remarkably, in presence of W antenna limiters, the fraction of radiated power using LHCD or ICRH is similar [Colas], as long as the ICRH coupling conditions are optimized [Hillairet].
In the ohmic phase, W radiation can lead to central cooling hence deleterious (2,1) MHD modes [Maget]. Nitrogen injection during this early phase, by increasing the edge resistivity, leads to more peaked electron temperature, hence reduced MHD [Manas], allowing for higher performance of the RF-heated phase. Up to 9.2 MW of combined ICRH and LHCD power has been achieved and up ~5MW/1s separately [Hillairet, Liang].
In L-mode, the stored energy, WMHD, increases according to the ITER96 L-mode scaling law up to 350kJ. L-H transitions are observed after fresh boronization, when combining 4MW of LHCD with 1MW of ICRH [Goniche], for a power crossing the separatrix of the order of the Martin 2008 scaling law [Martin J. Phys. Conf. Ser. 2008]. It results in a significant increase of the particle confinement time (30% increase of plasma density with gas injection turned off). The Doppler reflectometry ExB velocity well gets deeper, reaching -5km/s [Vermare]. But, in most cases, the plasma radiation increases leading to an oscillatory regime.
On the actively cooled upper W divertor, long pulses lasting up to 55 s have been routinely achieved (see Fig. 2), with a loop voltage down to 90 mV [Goniche]. In these L mode, electron heated, torque free plasmas, no W accumulation is reported despite peaked density profiles attributed to dominant TEM turbulence [Manas].
The heat flux level and pattern on the lower and upper divertors have been characterized thanks to embedded thermal measurements, IR and flush-mounted Langmuir Probes. The maximum heat flux currently reported on the W-coated graphite components is slightly above 5 MW.m-2 [Gaspar]. This was obtained with a conducted power on the divertor of ~2MW in two different scenarios: 1/ with combined LH and ICRH power after boronization at low X-point height (dX = 40 mm), 2/ with LH power only and high X-point height (dX = 120 mm).
In SOLEDGE2D simulations, target temperature profiles measured by Langmuir probe, as well as the radiated power measured by bolometry, were well reproduced with 3% of Oxygen as effective medium Z charge [Ciraolo]. The simulated asymmetry of O between the inner and outer targets is in qualitative agreement with UV measurements. The force balance analysis shows that friction dominates over thermal gradient forces at the inner target, while, at the outer target, the repelling thermal gradient forces dominate.
On the ITER-grade PFU, cracking and local melting have been observed for misaligned PFU. In addition, optical hot spots, which have been predicted to occur in ITER at the projection of the toroidal gaps on the subsequent PFU, have been observed experimentally, even for PFU aligned within specifications [Diez].
Finally, a He campaign has been run to investigate interactions between He plasmas and W PFC, in particular the formation of W fuzz [Tsitrone, Pegourie, Douai]. More than a hundred of 20-30 s pulses were repetitively performed in LSN. The conditions for W fuzz formation have been reached in the outer strike point area on the inertial W divertor (Einc > 20 eV, fluence > 1e24 He.m-2, Tsurf > 700 °C). Articulated Inspection Arm inspections before and after the He campaign have shown no macroscopic sign of surface modification. Post mortem analysis of the W components is ongoing to characterize the He induced nanostructures formed.
WEST phase 2 will start in autumn 2020, to address long pulse / high fluence operation on the newly manufactured ITER-grade actively cooled divertor, up to 10 MW/1000 s.
References:
[Diez], [Reux], [Goniche], [Colas], [Hillairet], [Maget], [Manas], [Liang], [Gaspar], [Ciraolo], [Tsitrone], [Pegourié], [Douai], this conference. [Martin 2008] J. Phys.: Conf. Ser. 123 012033Speaker: Jerome Bucalossi (CEA) -
Spherical tokamak (ST) research in Japan 1 is being conducted as a nationally coordinated program of university-scale ST devices under the ST Research Coordination Subcommittee organized by National Institute for Fusion Science (NIFS). The roles of university ST research include: (1) unique and challenging research through creativity and innovation which might be considered too risky for large ST devices, (2) establishment of the scientific basis for achieving ultra-high beta and ultra-long pulse (Fig. 1),
![Japanese ST research strategy [1].](https://workshop.nifs.ac.jp/fec2020/image/101-Takase-image-fig_1.jpg)
(3) contribution to the scientific basis for practical and economically competitive fusion power, complementing the mainline tokamak research (JT-60SA, ITER, etc.), and (4) development and training of a future generation of world-leading tokamak scientists. Specific research topics include: (a) development of start-up, current drive, and control techniques without the use of the central solenoid (CS), (b) formation and sustainment of very high beta plasmas, and (c) demonstration of steady-state operation and the study of steady-state issues such as heat and particle control, divertor physics, and plasma-wall interaction.
(a)-1. Plasma current ($I_p$) start-up by RF waves: Electron cyclotron wave (ECW) at 2.45 GHz and 5 GHz are used to excite the electron Bernstein wave (EBW) via O-X-B mode conversion on LATE (Kyoto U.). Highly overdense ST plasmas (up to 7 times the plasma cutoff density) are formed when the fundamental EC resonance layer is located in the plasma core, and EBW is excited in the 1st frequency band ($\omega_{ce} < \omega < 2\omega_{ce}$). Whereas EBW in the 1st frequency band heats the bulk electrons, EBW in the 2nd frequency band is absorbed by high-energy electrons and drives $I_p$. Intermittent plasma ejections across the plasma boundary synchronized with poloidal field decrement were observed in highly overdense plasmas. Oscillations in the Alfven frequency range and potential increase were observed, suggesting the loss of high-energy electrons. The 28 GHz RF injection system on QUEST (Kyushu U.) can regulate wave polarization and parallel index of refraction ($N_{||}$) with a beam radius focused down to 50 mm. $I_p > 100\ kA$ was achieved by injecting X-mode with $N_{||} = 0.78$, assisted by poloidal field induction. The RF power is likely absorbed by energetic electrons. Electron temperature of up to 0.5 keV was obtained by injecting X-mode with $N_{||} = 0.1$, indicating that effective bulk heating is possible. In $I_p$ start-up experiments using the lower hybrid wave (LHW) on TST-2 (U. Tokyo), top-launch was found to be more efficient for Ip ramp-up than outboard-launch. A 2-dimensional phase space model explaining X-ray emission shows that LHW driven radial transport is the dominant loss mechanism of fast electrons, and higher density is preferable in the present situation [2]. This is the first clear demonstration of RF driven electron transport.
(a)-2. CS-less $I_p$ start-up by non-RF methods: In transient CHI, magnetic reconnection plays an important role in the formation of closed flux surfaces during $I_p$ start-up. Plasmoid-driven reconnection following the tearing instability of the elongated current sheet and associated ion heating in the presence of the toroidal guide field were investigated on HIST (U. Hyogo) [3]. Several small-scale plasmoids generated during the injection phase merge with each other to form one or two large-scale closed flux surfaces during the decay phase. Transient CHI is also investigated on QUEST (US-JA collaboration). Investigation of synergistic effects of electron beam injection and EBW current drive in overdense plasmas has begun on LATE [4].
(a)-3. Optimization of inductive plasma start-up: In low voltage inductive $I_p$ start-up with ECW pre-ionization on TST-2, application of a weak vertical field with positive decay index during breakdown was found to be beneficial at low pre-fill pressure and high ECW power. Application of ECW power extended the low pressure limit for breakdown as well as the high pressure limit for burn-through. An MHD equilibrium model with fast electron orbits taken into account, and a model to simulate electron diffusion in both velocity space and real space are being developed.
(b) Access to high temperature and/or beta regime: Reconnection heating for direct access to burning plasmas, being investigated on TS-3U, TS-4U, UTST (U. Tokyo), will be reported in Ref [5].
(c) Demonstration of steady state operation by high-temperature wall: A high-temperature wall plays an essential role in reducing wall-stored hydrogen and facilitates hydrogen recycling. A clear extension of pulse duration at the wall temperature of 473 K was observed on QUEST by water cooling, indicating that recycling can be controlled by wall temperature. During long duration discharges, a high concentration of neutral particles was achieved behind the bottom divertor plate [6].
Stabilization using helical field coils: Suppression of the oscillation and the outer displacement in the radial position was observed by applying the helical field to the tokamak plasma on TOKASTAR-2 (Nagoya U.), which is an ST-helical hybrid device equipped with parallelogram-shaped partial helical field coils [7].
ST research in Japan has produced many innovative results including (i) $I_p$ start-up by LHW (TST-2), EBW (LATE), ECW/EBW (QUEST), CHI (HIST, QUEST), electron beam (LATE); (ii) optimization of ECW-assisted inductive start-up and pre-ionization by AC operation of the Ohmic coil (TST-2); (iii) extension of ion heating by plasma merging (TS-3U, TS-4U, UTST); (iv) hydrogen recycling control with high-temperature wall (QUEST); and (v) radial position stabilization by superposed helical field (TOKASTAR-2).
1 Y. Takase et al., Nucl. Fusion 57, 102005 (2017).
[2] A. Ejiri et al., this conference; N. Tsujii et al., this conference.
[3] M. Nagata et al., this conference.
[4] H. Tanaka et al., this conference.
[5] Y. Ono et al., this conference.
[6] T. Onchi et al., this conference; K. Hanada et al., this conference.
[7] K. Yasuda et al., Plasma Fusion Res. 13, 3402072 (2018).Speaker: Yuichi Takase (University of Tokyo) -
The report provides an overview of the results obtained at the upgraded Globus-M2 spherical tokamak 1 since the last IAEA conference. The tokamak was designed to reach the toroidal magnetic field as high as BT =1 T and the plasma current Ip = 0.5 MA having a small plasma minor radius a = 0.22-0.23 m. Currently 80% of highest magnetic field and plasma current value are reached, so during the reported period the experiments were performed with the toroidal magnetic field up to 0.8 T and plasma current up to 0.4 MA. The plasma breakdown conditions were improved noticeably with regard to the Globus-M ones, 30% breakdown loop voltage decreasing was achieved. The discharge duration was increased due to higher central solenoid volt-second consumption. The plasma column magnetic configuration explored was the divertor lower null with the aspect ratio A = R/a = 1.5-1.6, triangularity up to δ~0.35 and elongation up to κ~2.2.
The first neutral beam heating experiments on Globus-M2 have demonstrated an increased efficiency, comparing with the Globus-M ones, at the same NBI parameters (deuterium beam with particle energy 28 keV and the heating power 0.8 MW). The electron and ion central plasma temperatures exceeded 1 keV at the central density as high as 1×10^20 m-3. The diamagnetically measured plasma thermal energy increased up to 10 kJ, which is nearly triple as high as in Globus-M (BT =0.4, Ip = 0.2 MA). NPA spectra demonstrating improved fast particle confinement are presented. The energy confinement time increased more than two times that is significantly higher than the IPB98(y,2) scaling predicts. The effect is due to the strong dependence of the energy confinement time on the toroidal magnetic field in accordance with the Globus-M experimental scaling that is found to be valid for a wider range of BT. The regression fit of the Globus-M/Globus-M2 data yields the following scaling for energy confinement time:
τE ~ Ip^(0.58)BT^(1.23)Pabs^(-0.66)ne^(0.63)
where Pabs is the absorbed heating power and ne is the line average density. The scaling confirms weak τE dependence on Ip that emphasizes the major role of BT on heat perpendicular transport in spherical tokamaks, Enhanced plasma parameters allowed us to obtain regimes with much lower collisionality. That make possible investigation of dependence of the normalized energy confinement time (BTτE) on collsionality (ν~ne/T^2) in the wide range of plasma collisionalities 0.018<ν< 0.23. This dependence turned out to be rather strong BTτE ~ ν^(-0.8) for a fixed values of safety factor q ~ BT/Ip, normalized ion gyroradius ρ ~ T^(0.5)/BT and parameter βT ~ W/BT^2. The power balance analysis carried out using ASTRA transport code indicates the reduction of both electron and ion heat diffusivity with collisionality decrease while the ion heat diffusivity remains near the neoclassical level.
Important results are related to non-inductive current drive. About 30% of the loop voltage drop was recorded during the NB injection, which indicates a noticeable amount of non-inductively (mainly bootstrap) driven current. For the first time in spherical tokamaks a non-inductively driven current was recorded during the launch of the electromagnetic waves of the lower hybrid (LH) range (2.45 GHz) with the help of toroidally oriented grill. The fraction of noninductively driven current has exceeded 30% in the discharge with the total current of 0.2 MA. The modelling results of the experimental data by means of the ASTRA transport code and Fast Ray Tracing Code incorporated to ASTRA 2 are presented.
Plasma scrape of layer (SOL) and divertor characteristics were investigated in new experimental conditions of enhanced magnetic field and plasma current. Heat and particle fluxes together with currents and potentials in SOL and divertor plate vicinity were measured with a divertor Langmuir probe array and movable Langmuir probe. The plasma parameters in SOL were also modelled with the fluid version of the SOLPS-ITER code. Currents and drifts were included in the simulations. Comparison of experimental and simulated heat flux power density decay length (λqt) in SOL with the well-known scalings is presented.
The study of Alfvén modes (АМ) was continued during the reported period. An increase in plasma parameters led to a change in the nature of AM and the expansion of their frequency spectrum (50–300 kHz). Together with the toroidal Alfvén eigenmodes (TAE), observed earlier on Globus-M, the so-called Alfvén cascades (AC or RSAE) were identified. Observation of ACs made it possible to apply the method of MHD spectroscopy to determine the evolution of qmin in a discharge. In experiments on current drive by the LH waves, modes with a frequency of about 1 MHz, excited by fast electrons, were detected. To study the spatial structure of AM, Doppler backscattering diagnostics was used [3] with application of a multi-channel microwave scheme. Using the neutral particle analyzer and a neutron detector, we studied the dependence of fast particle losses initiated by TAEs on the magnetic field and plasma current. It was shown that losses decrease significantly with increasing field and current, demonstrating dependence favorable for compact neutron sources.
Also presented are new diagnostics designed to fill in the missing data on plasma parameters and improve the quality of the simulation, such as: diagnostics Z eff, laser interferometer, charge-exchange resonance spectroscopy (CXRS), etc.
1. V.B. Minaev et al 2017 Nucl. Fusion 57 066047
2. A.D. Piliya, A.N. Saveliev, JET Joint Undertakin Abingdon, Oxfordshire, OX14 3EA, 1998
3. V.V. Bulanin et al 2019 Tech. Phys. Lett., v.45, 11 p.p. 1107-1110Speaker: Yury Petrov (Ioffe Institute) -
The ADITYA Upgrade (ADITYA-U) is a medium sized (R0 = 75 cm, a= 25 cm) tokamak having toroidal graphite limiter, configured to attain shaped-plasma operations with an open divertor in single and double-null configurations [1]. The foremost objective of ADITYA-U is to prepare the physics and the technological base for future larger tokamaks by expanding the ADITYA-U operating space and by performing dedicated experiments for validation of physics models. Since the 2018 IAEA-FEC Conference, ADITYA-U operations have been mainly devoted on realizing the plasma parameters close to the design parameters of circular plasmas in limiter configuration and also the initiation of shaped plasma operation. Emphasis has been given to novel experiments on runaway electrons (REs) and disruption control in the ADITYA-U [2]. Furthermore, experiments on radiative-improved modes using Ne, Ar gas injection, modulation of MHD modes [3] and edge turbulence using periodic fuel gas-puffs, density dependence of plasma toroidal rotation reversal [4], fuelling using SMBI etc.
For the typical discharges in ADITYA-U, in absence of any strong pre-ionization, the gas breakdown and successful plasma start-up is normally achieved with peak loop voltages of ~ 18 – 20 V (Electric field ~ 4.5 V/m). 42 GHz ECR [5] assisted low loop voltage (~10 – 12 V, Electric field ~2.1 V/m) start-up was successfully achieved with wave launched in fundamental O-mode from low field side. The toroidal magnetic field is ~ 1.4 T. Plasma discharges having plasma current ~ 170 kA, plasma duration ~ 330 ms, chord-averaged electron density ~ 2 – 6 x 10^19 m^-3 and central electron temperature ~ 300 – 500 eV has been achieved. The time evolution of typical high current, longer duration discharge of ADITYA-U is shown in Figure 1 and the overall progress of plasma current and duration enhancement during the year 2018-2019 are shown in Figure 2(a) and 2(b) respectively. Repeated cycles of vacuum vessel baking up to 135° C, followed by extensive wall conditioning using novel techniques [6] along with lithium coating [7] resulted in substantial reduction in partial pressures of various mass species and achievement of lower base vacuum of ~ 6 x 10^-9 Torr. Successful recovery of volt–sec along with adequate control of real-time horizontal plasma position and multiple gas puffing led to the achievement of longer discharge duration discharges in ADITYA-U.
_Fig.2(b)_RLTanna.jpg)
Over the last two years, significant progresses have been achieved in ADITYA-U experimental research including, 42GHz ECR assisted low loop voltage start-up and heating experiments, electromagnetically driven pellet impurity injector for injecting micron size particles at high velocity (~220m/s) to understand disruption mitigation, wall conditioning by using different techniques of lithium coating, formation of runaway beam and its avoidance, Neon and Argon impurity injection for radiative improved modes, suppression of electrostatic fluctuations using hydrogen gas puff and its correlation with hard X-rays, effect of positive edge electrode biasing on drift tearing modes and runaway transport and effect of SMBI with edge safety factor (qedge), toroidal rotation reversal threshold studies etc. Dependence of current quench time (CQT) on qedge during disruptions and relation of CQT on the prevailing MHD activities prior to disruption has been studied in detail. Discharges with low qedge showed a high CQT compared to those observed at high qedge. Exploring the RE formation mechanisms, controlled RE generation experiments have been carried out by lowering the plasma density and adjusting the vertical magnetic field as shown in Figure 3 (#33061).
Fast visible imaging video camera, used for 2D tangential viewing, captured images which showed features of the RE beam formation with high spatial and temporal resolution as shown in Figure 4. This RE beam generation in controlled fashion is very useful in studying the mitigation techniques of REs using different techniques as the prevention of such RE beam is of a vital importance in future tokamaks, especially in the ITER, because of its potential danger to the plasma facing components.
In another significant RE experiment carried out in ADITYA-U, correlated suppression of RE loss, evident from hard X-rays intensity measurements, has been observed with suppression of edge electrostatic fluctuations in discharges where magnetic fluctuation amplitudes are not sufficient for affecting REs. Multiple periodic gas puff, which are used for plasma fuelling, suppresses the electrostatic fluctuations in the floating potential as well as the density fluctuations (measured with Langmuir probes) in the edge region. Figure 5 shows the suppression of edge turbulence due to gas puff. This is also observed clearly in the spectrogram of density fluctuations. The hard X-rays flux is also seen to be highly correlated with the edge turbulence. The decrease in edge turbulence is accompanied by a decrease in HXR flux. This observation of correlated effect of electrostatic fluctuation on RE transport throws light on new mechanism of RE loss and may be exploited to design novel RE mitigation methods.
In another interesting experiment, the REs are confined by applying a voltage to an electrode placed at the edge region of the plasma prior to the disruption. Figure 6 shows the multiple pulses of biasing applied to an electrode placed ~ 2.5 cm inside the LCFS for the shot #33336. The last pulse of biasing coincides with the plasma disruption as shown in Figure 7, describes that in presence of biasing during the disruption, the HXR intensity as well as HXR flux persists for ~10 ms even after the termination of the plasma current. A possible mechanism of confinement is the Er x BΦ motion of REs after the disruption, where the Er is generated by electrode biasing.
For the first time, ADITYA-U has experimentally demonstrated the use of electromagnetically driven payloads for particle injection into the tokamak plasmas for disruption mitigation studies. The impurity particles reached the core of the plasma within ~ 1.25 ms and causes fast termination of plasma current and radiate the whole plasma energy in ~ < 2 ms. Furthermore, the preliminary experiments related to plasma shaping by charging the divertor coils during plasma current plateau, confinement improvement with ion cyclotron resonance (ICRH) assisted auxiliary heating and deuterium injection is undergoing and the results of the same will be presented. This paper summarizes the experimental research of ADITYA-U tokamak in the key areas of thermo-nuclear fusion over the last two years.References
[1] Tanna R. et al 2018 Plasma production and preliminary results from the ADITYA Upgrade tokamak Plasma Sci. Technol. 20 074002.
[2] Tanna R. et al 2019 Overview of operation and experiments in the ADITYA-U tokamak Nucl. Fusion 59 112006.
[3] Raj Harshita et al 2020 Effect of periodic gas-puff on drift-tearing modes in ADITYA/ADITYA-U tokamak discharges Nucl. Fusion 60 036012.
[4] Shukla G. et al 2019 Observations of toroidal plasma rotation reversal in the ADITYA-U tokamak Nucl. Fusion 59 106049.
[5] Shukla B.K. et al 2019 Commissioning of Electron Cyclotron Resonance Heating (ECRH) system on tokamak ADITYA-U Fusion Eng. Des. 146 2083 – 86.
[6] Jadeja K.A. et al 2019 Novel approach of pulsed-glow discharge wall conditioning in the ADITYA Upgrade tokamak Nucl. Fusion 59 086005.
[7] Jadeja K.A. et al Lithium Wall Conditioning Techniques in ADITYA-U Tokamak for Impurity and Fuel Control (In this conference).Speaker: RAKESH TANNA (Institute for Plasma Research) -
Using its unique flexibility and advanced plasma diagnostics, the TJ-II stellarator is contributing to the understanding and solution of critical challenges in fusion plasmas. Next, we highlight some of the most relevant recent results in the framework of its research programme.
Towards validation of gyrokinetic and neoclassical simulations. Aiming at the validation of the instability properties predicted by gyrokinetic (GK) simulations and of the electrostatic potential variations on the flux surface, φ1, calculated by neoclassical (NC) codes, dedicated experiments have been carried out in TJ-II for a systematic characterization of turbulence wavenumber spectra and perpendicular rotation velocity measured by Doppler Reflectometry (DR) at poloidally separated positions on the same flux surface [1]. Poloidal asymmetries in the intensity of the wave number spectrum that depend on plasma conditions have been characterized and compared with global linear GK simulations by the code EUTERPE. Model and experiment qualitatively agree in the radial dependence of the turbulence intensity, in the turbulence dispersion relation and in showing a poloidal asymmetry that depends on the magnetic configuration. Recent experiments exploring configurations with different magnetic ripple have shown a reduction in the turbulence asymmetry at configurations with reduced ripple. Besides, the influence of base ion mass has been investigated in hydrogen and deuterium plasmas. The ion mass in TJ-II plasmas does not affect the properties of the turbulence, neither the amplitude nor the spectral shape or the poloidal asymmetry. The lack of dependence of the turbulence spectrum on the ion mass is also found in GK simulations. Model validation will also benefit from the effort of verification of GK simulations in different computational domains [2] as well as from the application of the recently developed GK code stella to multispecies turbulent transport calculations in TJ-II [3].
Poloidal asymmetries in radial electric field, Er, are found that depend on the plasma collisionality. These results have been compared with the contribution to Er arising from −φ'1 as calculated with the NC version of the code EUTERPE. These results show variations in Er comparable in size to those found in the experiments, but there is a disagreement regarding the sign of the Er correction. Recent simulations performed with the newly developed NC code KNOSOS [4] show that the effect of kinetic electrons on φ1 has to be taken into account due to the strong Te dependence of the electron contribution to φ1 when the electrons are in the 1/ν regime [Fig. 1]. KNOSOS (KiNetic Orbit-averaging SOlver for Stellarators) is a freely available, open-source code that calculates neoclassical transport in low-collisionality plasmas of three-dimensional magnetic confinement devices by solving the radially local drift-kinetic and quasineutrality equations.
Experiments in TJ-II with cryogenic and TESPEL pellets show that post-injection particle radial redistributions can be understood qualitatively from neoclassical predictions while also providing a means to benchmark the HPI2 code [5]. TJ-II measurements provide 2-D maps of plasma potential and fluctuations to address the question of how hollow density profiles, created by pellets, affect turbulent transport. Fluctuations are stronger in the negative density gradient than in the positive one in consistency with TEM linear simulations [6].
Towards the validation of fast ion induced stabilization and the identification of Alfvén Eigenmode actuators. An ambitious research programme is in progress to investigate the relation between zonal structures and Alfvén eigenmodes (AE) and its role on the nonlinear dynamics of AEs and transport as well as to develop and demonstrate AE control strategies using ECRH and ECCD in TJ-II. The unique TJ-II experimental set-up using a dual HIBP has shown that, in some conditions, long range correlations (LRC) are detected both at the AE frequencies and low frequencies (<10 kHz) [Fig. 2]. LRC are observed in plasma potential fluctuations but not in density fluctuations as expected in zonal flow structures [Fig. 2]. It is an open question whether those zonal structures are directly driven by fast particle effects or/and are the consequence of plasma scenarios with reduced damping of zonal flows. Experiments in TJ-II have demonstrated the effectiveness of ECRH and ECCD actuators to modify AE activity [7].
Towards the characterization of the interaction between neoclassical and turbulent transport mechanisms. We have investigated the impact of Er on turbulence propagation and the coupling between the plasma edge and the scrape-off layer (SOL) during electron–ion root transitions where Er is changed in a controlled manner from positive to negative values. It is shown that Er does not only affects the radial turbulence correlation length but it is also capable of reducing the propagation of turbulence from the edge into the SOL. This result was obtained using a technique based on the transfer entropy, which quantifies the propagation of information [8]. These observations are highly relevant for the understanding of the mechanisms that determine the SOL width.
The interplay between of NC radial electric fields, Reynolds stress gradients and LRC has been investigated in different plasma scenarios in TJ-II. Turbulent driven acceleration alone cannot explain the dynamics of zonal flows whose radial width is affected by the isotope mass [9]. These results are in line with the expectation that the interplay between turbulent and neoclassical mechanisms is an important ingredient of the dynamics of edge zonal flows.
Power exhaust physics: liquid metals. Solid and liquid samples of Li/LiSn/Sn, in a Capillary Porous System (CPS) arrangement, have been exposed to the edge plasma [10]. A simple 1D model was applied to the data, allowing for the evaluation of the kinetic energy (Ek) of ejected atomic species while their residence time at the edge was determined by monitoring the ratio of first ion/neutral emission light intensities. A clear evolution of Ek with sample temperature was deduced for Li atoms, this being associated to the different relative contributions of sputtered/evaporated atoms. SnI emission into the plasma has also been measured with radial and toroidal resolution. The deduced mean free paths for the ejected Sn atoms under sputtering conditions (low T) imply unrealistic high energies if the bibliographic data for the ionization rate constant of Sn are assumed. For the LiSn case, Li as well as Sn emissions were simultaneously detected and analysed. Plasmas under cut-off (collapsing) conditions were also investigated to check for the sensitivity of the recorder line intensities and ionization rates to the edge electron temperature.
[1] T. Estrada et al., Nuclear Fusion 59, 076021 (2019)
[2] E. Sánchez et al. Gyrokinetic simulations in stellarators using different computational domains, 28th IAEA Nice 2020
[3] J. M. García-Regaña et al., Turbulent transport of impurities in 3D devices, 28th IAEA Conf. Nice 2020.
[4] J. L. Velasco et al.,KNOSOS, a fast neoclassical code for three-dimensional configurations, 28th IAEA Nice 2020
[5] K. McCarthy et al., Pellet studies in TJ-II, 28th IAEA Nice 2020
[6] A. Melnikov et al., 2-D mapping of fluctuations and plasma profiles in TJ-II, 28th IAEA Nice 2020.
[7] A. Cappa et al., AE control strategies in TJ-II, 28th IAEA Conf., Nice 2020.
[8] G. Grenfell et al., Nucl. Fusion 60 014001 (2020)
[9] R. Gerrú et al., Nucl. Fusion 59 106054 (2019)
[10] F. Tabarés et al., Liquid metal studies in TJ-II, 28th IAEA Conf., Nice 2020Speaker: Prof. Carlos Hidalgo (CIEMAT) -
Achieving ignition and high fusion yield in the laboratory is a central goal of the U.S. Inertial Confinement Fusion (ICF) Program. Three major and credible approaches are currently being pursued: laser indirect-drive (LID), laser direct-drive (LDD), and magnetic direct-drive (MDD). While the three approaches use very different means for driving a spherical or cylindrical implosion that can compress and heat a mass of deuterium-tritium (DT) fuel (by laser-generated radiation drive, direct laser irradiation, or direct magnetic acceleration, respectively), they share many common challenges, including how to efficiently couple the kinetic energy of the implosion to internal energy of the fuel at stagnation, and how to assemble the fuel at the necessary conditions, in pressure, temperature, and areal density to initiate self-heating and ignition.
Significant progress has been made in each of these approaches in recent ICF experiments on the National Ignition Facility (NIF) at Lawrence Livermore National Laboratory, the OMEGA laser facility at the Laboratory for Laser Energetics, and the Z pulsed power facility at Sandia National Laboratories. This includes progress in both advancing the absolute fusion output and, as importantly, improving our knowledge and understanding of the implosion behavior and causes of deviations from ideal or theoretical performance.
In this overview, we will review some of the major results from each facility, with a particular focus on recent advances in diagnostic measurement techniques and analysis that have improved our understanding of the states of the assembled fuel and confining shell at stagnation. Particular emphasis is placed on recent advances in determining the 3D spatial morphology and thermodynamic properties of the fusion fuel, including pressure, temperature, areal density, and mix, that determine the degree of alpha-particle production, confinement, and self-heating, and proximity to ignition.
As an example, Fig. 1 shows the hot spot temperatures and areal densities inferred from some of the principal LID experimental campaigns performed on the NIF since 2011. Improved performance can be observed as the implosion design evolved from the original low-adiabat† (a~1.6) CH ablator design (2011-2013), to a mid-adiabat (a~2.3) CH design (2013-2015), and then to a mid-adiabat (a~2.7) high-density-carbon (HDC) ablator design (2016-2019) [1-3]. The principal gains resulted from improved hydro-instability at higher adiabat, and improved energy coupling and implosion symmetry with HDC designs. As of today, the highest performing HDC implosions have achieved hot spot temperatures of ~4.5 keV, hot spot areal densities of ~0.3 g/cm2, hot spot pressures of ~350 Gbar, and fusion energy output of >50 kJ. At these conditions, the self-heating of the hot spot by alpha-particle deposition is estimated to be amplifying the total fusion output by a factor of ~2.5-3x. As can be seen in Fig. 1, these conditions are now quite close to the static self-heating boundary, dT/dt>0, where for a static hot spot at the time of peak compression, the time derivative of temperature is positive due to alpha-particle heating exceeding energy losses from radiation and thermal conduction. Whilst reaching this boundary will be a very noteworthy physics achievement, it is not a sufficient condition for ignition. For ignition, the alpha-heating rate must exceed all losses including mechanical work from expansion of the hot spot after peak compression. This condition, equivalent to the requirement d2T/dt2>0, depends on the shell confinement and necessitates on the order of a ~1 keV higher hot spot temperature [4,5].
In the LDD approach, remarkable progress has been made using a data-driven statistical mapping model to optimize input laser and target parameters that has led to a 3x increase in neutron yield over the past two years on OMEGA [6]. The best-performing spherical direct drive (SDD) implosion has a hydrodynamically-scaled predicted fusion yield in the 0.5 MJ range at NIF laser energy, approaching the burning plasma regime. Polar direct drive (PDD) implosions and laser-plasma coupling research are performed on OMEGA and the NIF for the LDD approach [7,8]. In the MDD approach, recent magnetized liner inertial fusion (MagLIF) experiments on Z have produced record yields and stagnation parameters through steady enhancements in the target initial magnetization, laser preheat energy delivered to the fuel, and electrical current delivered to the imploding shells [9]. In addition to record fusion performance, MagLIF implosions have also demonstrated significant potential for DT alpha-particle confinement and stopping via the strong magnetic fields entrained in the fuel [10,11], a necessary requirement for self-heating and ignition.
This work was performed under the auspices of Lawrence Livermore National Security, LLC (LLNS) under Contract DE-AC52-07NA27344, the Sandia National Laboratories, a multi-mission laboratory managed and operated by National Technology & Engineering Solutions of Sandia, LLC, a wholly owned subsidiary of Honeywell International Inc., for the U.S. DOE NNSA under contract DE-NA0003525, and upon work supported by the U.S. DOE NNSA under Award Number DE-NA0003856, the University of Rochester, and the New York State Energy Research and Development Authority.
† The adiabat is the ratio of the pressure of the DT fuel to the Fermi degenerate pressure.
References
1. J. Lindl et al., Phys. Plasmas 21, 020501 (2014)
2. O. A. Hurricane et al., Nature 506, 343 (2014)
3. S. Le Pape et al., Phys. Rev. Lett. 120, 245003 (2018)
4. P. T. Springer et al., Nucl. Fusion 59, 032009 (2019)
5. P. K. Patel et al., Phys. Plasmas (submitted, 2020)
6. V. Gopalaswamy et al., Nature 565, 581 (2019)
7. F. J. Marshall et al., Phys. Plasmas 23, 012711 (2016)
8. P. B. Radha et al., Phys. Plasmas 23, 056305 (2016)
9. M. R. Gomez et al., manuscript in preparation (2020)
10. P. F. Schmit et al., Phys. Rev. Lett. 113, 155004 (2014)
11. P. F. Knapp et al., Phys. Plasmas 22, 056312 (2015)Speaker: Pravesh Patel (Lawrence Livermore National Laboratory) -
The mission of the spherical tokamak NSTX-U is to advance the physics basis and technical solutions required for optimizing the configuration of next-step tokamak fusion devices, and to advance the development of the ST concept towards a compact, low-cost Pilot Plant 1. NSTX-U will operate at up to 2 MA and 1 T with up to 10 MW of Neutral Beam Injection (NBI) power for 5 seconds and 4 MW of High Harmonic Fast Wave (HHFW) power. NSTX-U has three main objectives: to explore confinement and stability at low aspect ratio and high beta at low collisionality, to develop the physics understanding and control tools to ramp-up and sustain high performance plasmas in a fully-non-inductive regime, and to develop and evaluate conventional and innovative power and particle handling techniques to optimize plasma exhaust in high performance scenarios. Following the initial 2016 NSTX-U run campaign, analysis has continued during NSTX-U Recovery to address physics issues to develop understanding and capabilities that, once operation commences, will aid in achieving these three objectives.
Stability and 3D physics: The resistive DCON model for calculating tearing mode stability ($\Delta'$) has been developed, benchmarked against extended-MHD (M3D-C1) simulations, and used to identify regions in $q_{95}-\beta_N$ that are simultaneously stable to both 2/1 tearing modes (Fig.1) and n=1 ideal kink.
M3D-C1 has been extended to predict Te and Ti independently in the presence of impurities during a thermal quench 2. The simulations show a contraction of the current channel that is sensitive to temperature, and a fast stochastization of the B-field, highlighting the difficulty in cooling the plasma while avoiding a thermal quench using impurity injection. Continued analysis on 3D error fields have shown that misalignment of the toroidal field (TF) was the largest source of the error field on NSTX-U during initial operation in 2016, and these calculations, guided by constraints on the field line pitch at the divertor plates in order to mitigate potentially high heat fluxes, were used to drive the engineering tolerances for TF shift and tilt for NSTX-U Recovery [3]. Additional calculations show that misalignments in other PF coils lead to extended divertor footprints but which are contained within the divertor region designed to handle high heat fluxes [4]. Energetic particle (EP) physics: The phase-space-resolved reduced EP transport kick model has been extended to include non-Alfvénic low-frequency perturbations, reproducing observations of (i) large fast ion losses due to synergistic effects when TAE and fishbones/kink instabilities occur simultaneously, and (ii) enhanced fast ion loss due to NTMs when the island width exceeds a threshold [5]. The stability and scaling of global Alfven eigenmodes (GAEs), previously correlated with central Te flattening in NSTX, has been predicted using hybrid MHD/kinetic-fast-ion simulations [6] and also newly derived analytic instability conditions [7], revealing a previously unidentified instability regime necessary to explain observed GAE excitation and stabilization. The chirping and avalanche behavior of Alfven eigemodes that can influence fast-ion losses in NSTX-U (and burning plasma $\alpha$ losses) has been predicted using a guiding center code with a delta-f formalism [8]. New analysis from NSTX and NSTX-U data has provided a detailed picture of ion cyclotron emission (ICE), being considered as a possible diagnostic of confined $\alpha$’s in burning plasma experiments [9]. A self-consistent resonance-broadened quasi-linear (QL) model has been developed for relaxation of fast ion distribution function by Alfvenic modes [10].
Transport physics: Analysis in moderate $\beta$ NSTX scenarios using gyrokinetic simulations coupled with a novel “synthetic diagnostic” identify conditions where both electron thermal transport and turbulence measured by high-k microwave scattering are explained entirely by short-wavelength electron-scale ETG turbulence [11]. Gyrokinetic analysis in high-$\beta_{pol}$ scenarios envisioned for high non-inductive fraction operation indicates the deep-core profiles (with relatively flat Te) sit very near KBM (or EPM) limits when including only thermal ions (or thermal + fast ion species), suggesting core profiles may ultimately be constrained by $\nabla p$-limited ballooning modes. Analysis of enhanced Pedestal (EP) H-modes demonstrates that this high confinement (H98$\leq$1.8), wide-pedestal ($\Delta\psi_{N,ped}\leq0.4$), ELM-free regime is accessed at low edge ion collisionality, e.g. via reduced wall recycling with lithium wall coatings. While MTM, TEM and ETG instabilities predicted in this region may account for electron thermal losses, it is hypothesized that peeling-ballooning and near-marginal KBM instabilities (predicted via MHD and gyrokinetic simulations, respectively, Fig. 2) may enhance edge particle transport during the evolution to the EPH phase that helps to sustain the low-density, low-collisionality state.
RF physics: A new 2D full wave code (FW2D) has been updated to predict the sensitivity of SOL losses to variations in the realistic boundary shape for high harmonic fast wave (HHFW) heating [12]. Losses are further predicted to be minimized when the SOL density is near the critical density for fast wave cutoff, as found in experiments. Calculations of HHFW deposition in the presence of NBI using AORSA identify a competition between electron and fast ions absorption [13]. Additional simulations show that a sufficient concentration of H+ minority species could open up new HHFW heating scenarios in NSTX-U without NBI.
Scenarios & Control: A unified, physics-based reduced model for direct inductive startup, which computes the timing of plasma initiation and duration needed for plasma density buildup, has been developed for both NSTX-U and MAST-U [14]. Physics-based control-oriented models have been used to develop advanced-control approaches to determine neutral-beam current-drive requirements and evolutions to track prescribed current profiles in closed-loop [15], as well as to iteratively optimize the access to high-performance scenarios [16]. Aided by this is the development of a neural network-based description of beam heating and current drive profiles, well suited for rapid calculations and real-time application [17]. Development of a physics-based algorithm for closed loop feedback control of snowflake divertor configurations will provide real-time tracking and control capabilities [18].
The NSTX-U Recovery planning is ongoing, with NSTX-U targeting resumption of operations in 2021 [19].
This work was supported by US Department of Energy Contract No. DE-AC02- 09CH11466.
References
1 Menard, J.E., et al., this conference.
2 Ferraro, N.M., et al., Nucl. Fusion 59 016001 (2019).
[3] Ferraro, N.M., et al., Nucl. Fusion 59 086021 (2019).
[4] Munaretto, S., et al., Nucl. Fusion 59 076039 (2019).
[5] Podesta, M, et al., this conference.
[6] Belova, E., et al., this conference.
[7] J. B. Lestz et al., Physics of Plasmas 27, 022512 & 022513 (2020);
[8] White, R., et al., this conference.
[9] Fredrickson, E., et al., this conference.
[10] Gorelenkov, et al., this conference.
[11] Ruiz Ruiz, J., et al., Plasma Phys. Cont. Fusion 61 115015 (2019)
[12] Kim, E.-H., et al., Phys. Plasmas 26 062501 (2019)
[13] Bertelli, N., et al., Nucl. Fusion 59 086006 (2019)
[14] Battaglia, D.J., et al., Nucl. Fusion 59 126016 (2019)
[15] Ilhan, Z.O., et al., Fusion Eng. and Design 146 555 (2019)
[16] Wehner, W.P., et al., Fusion Eng. and Design 146 547 (2019)
[17] Boyer, M.D., et al., this conference.
[18] Vail, P.J., et al., Plasma Phys. Cont. Fusion 61 035005 (2019)
[19] Gerhardt, S.P., et al., this conference.Speaker: Walter Guttenfelder (Princeton Plasma Physics Laboratory) -
In the recent deuterium experiment on the Large Helical Device (LHD), we have succeeded to expand the temperature domain to higher region both in electron and ion temperatures as shown by the red region in Fig.1. We found a clear isotope effect in the formation of Internal Transport Barrier (ITB) in high temperature plasmas. In the deuterium plasmas, we have also succeeded to realize the formation of the Edge Transport Barrier (ETB) and the divertor detachment, simultaneously. It is found that the Resonant Magnetic Perturbation (RMP) has an important role in the simultaneous formation of the ETB and the detachment. A new technique to measure the hydrogen isotope fraction was developed in LHD in order to investigate the behavior of isotope mixed fuel ions. The technique revealed that the non-mixing state and the mixing state of hydrogen isotopes can be realized in plasmas.
One of the objective of deuterium experiment is the achievement of the reactor relevant high temperature plasmas in order to investigate the problems arising in the helical type fusion reactor. As shown in Fig.1, the ion temperature of 10keV was achieved at the electron temperature of around 4keV. This was reported at the last IAEA-FEC(FEC2018). The operational domain was extended to the direction of higher electron temperature range. The electron temperature of 6.6keV was achieved keeping the ion temperature of 10keV. For lower ion temperature plasmas (Ti0=~7keV), the electron temperature range was extended over ~12keV. It is found that the extension in higher electron temperature region of high ion temperature plasmas is an effective way to suppress the Energetic particle driven InterChange mode (EIC), which often prevents the access of high ion temperature plasmas in LHD. This is preferable feature for the future reactor scenario development 1.
The isotope effect is a long underlying mystery in plasma physics because the most of experimental observations in tokamak plasmas show favorable effect by the ion mass on the energy confinement time (tau_E) while the theoretical prediction based on the gyro-Bohm model shows unfavorable effect by the mass, i.e., tau_E~ M^-1/2. A series of dedicated experiment using dimensionally similar L-mode hydrogen and deuterium plasmas in LHD showed that the energy confinement time scaling for L-mode plasmas had a non-significant dependence on mass (M^0.01) 2. On the other hand, a clear isotope effect was observed for the formation of Internal Transport Barrier (ITB) in LHD as shown in Fig.2. Here, the profile gain factor G1.0 is defined as
and TLref is the expected temperature profile for L-mode plasmas with same heating condition 3. As shown in the figure, the ITB intensity, i.e., G1.0, is clearly larger for deuterium at the density range below 2x10^19 [m^-3].
The realization of both divertor heat load mitigation and the good core confinement property is an important issue in developing future reactor scenarios. For divertor detached deuterium plasmas which are realized by the RMP application, we have succeeded to improve the confinement of 38% with a formation of ETB. Thus, we call this improved confinement mode as the RMP induced H-mode 4. As shown in Fig.3, a steep pressure gradient was formed in deuterium plasmas at the edge region inside the m/n=1/1 island which was produced by the RMP. This formation of the steep edge pressure gradient with the detachment has never been observed in hydrogen plasmas. The thermal transport analysis based on TASK3D-code shows significant reduction in the thermal diffusivity at the edge region (r/a=~0.8) which indicates the formation of ETB.
In the future fusion reactor, the deuterium (D) and tritium (T) ions co-exist in plasmas and are assumed to be uniformly mixed with the ratio of D/T=1. On the other hand, it is not clear whether this assumption is always valid. To clarify the assumption, experiments to investigate the isotope mixing state were performed using Deuterium and Hydrogen mixture plasmas. It was found that the non-mixing state of isotopes can be realized when the centrally fuelled ion species are different from the ion species fuelled at the edge 5. Figure 4 shows a typical example of the non-mixing state where the tangential NBI fuels Hydrogen ions in the central region while the Deuterium is supplied at the peripheral region by a wall recycling (black symbols and curves). It was found this state can be changed to mixing state by an injection of ice pellet of hydrogen isotopes, i.e. either of Hydrogen or Deuterium. It was also found that a formation of hollow density profile due to the pellet injection plays an important role for changing the state.
1 H. Takahashi et al., this conference
2 H. Yamada et al., Phys. Rev. Lett. 123 (2019) 185001
3 T. Kobayashi et al., this conference
4 M. Kobayashi et al., this conference
5 K. Ida et al., this conferenceSpeaker: Prof. Masaki Osakabe (NIFS)
-
-
-
Operating a full tungsten actively cooled tokamak:
overview of WEST first phase of operation
J. Bucalossi and the WEST Team (http://west.cea.fr/WESTteam)
CEA, IRFM, F-13108 St-Paul-Lez-Durance, France.E-mail: jerome.bucalossi@cea.fr
WEST is a MA class superconducting, actively cooled, full tungsten (W) tokamak. Equipped with two up-down symmetric divertors, it operates at 3.7T, up to 1MA, with a plasma volume of 15 m3 and an aspect ratio between 5 and 6. CW RF power is installed: up to 9 MW of ICRH power and 7 MW of LHCD.
In support of ITER operation and DEMO conceptual activities, WEST aims at power exhaust studies in long and steady-state pulses, in various divertor configurations (LSN, USN, DN), in a full W environment. The lower divertor, partially made of ITER-grade Plasma Facing Units (PFU) complemented with inertially cooled W-coated elements in phase 1 (2017-2019, see Fig. 1), will be replaced by a complete ITER-grade divertor in phase 2, starting autumn 2020. This paper reports on the main findings from WEST phase 1, in terms of operational domain, plasma performance achieved, and first tests of the ITER grade PFU.Initial phase of operation was hindered by the production of runaway electron beams, when the ohmic current failed to rise quickly enough. Interestingly, start-up runaway electrons have been avoided by reducing the prefill pressure. Severe damages on PFC were observed [Diez] and one runaway beam impact induced radiation even quenched one superconducting coil [Reux].
Additionally to 200°C baking and glow discharge cleaning, boronizations have been performed in the second campaign leading to long lasting improved breakdown conditions and higher density operational domain.
Apart from the few post-boronization pulses, the fraction of radiated power in LHCD, ICRH or LHCD+ICRH pulses remains high, around 50%. Tungsten is, in most cases, the major radiating species [Goniche]. Remarkably, in presence of W antenna limiters, the fraction of radiated power using LHCD or ICRH is similar [Colas], as long as the ICRH coupling conditions are optimized [Hillairet].
In the ohmic phase, W radiation can lead to central cooling hence deleterious (2,1) MHD modes [Maget]. Nitrogen injection during this early phase, by increasing the edge resistivity, leads to more peaked electron temperature, hence reduced MHD [Manas], allowing for higher performance of the RF-heated phase. Up to 9.2 MW of combined ICRH and LHCD power has been achieved and up ~5MW/1s separately [Hillairet, Liang].
In L-mode, the stored energy, WMHD, increases according to the ITER96 L-mode scaling law up to 350kJ. L-H transitions are observed after fresh boronization, when combining 4MW of LHCD with 1MW of ICRH [Goniche], for a power crossing the separatrix of the order of the Martin 2008 scaling law [Martin J. Phys. Conf. Ser. 2008]. It results in a significant increase of the particle confinement time (30% increase of plasma density with gas injection turned off). The Doppler reflectometry ExB velocity well gets deeper, reaching -5km/s [Vermare]. But, in most cases, the plasma radiation increases leading to an oscillatory regime.
On the actively cooled upper W divertor, long pulses lasting up to 55 s have been routinely achieved (see Fig. 2), with a loop voltage down to 90 mV [Goniche]. In these L mode, electron heated, torque free plasmas, no W accumulation is reported despite peaked density profiles attributed to dominant TEM turbulence [Manas].
The heat flux level and pattern on the lower and upper divertors have been characterized thanks to embedded thermal measurements, IR and flush-mounted Langmuir Probes. The maximum heat flux currently reported on the W-coated graphite components is slightly above 5 MW.m-2 [Gaspar]. This was obtained with a conducted power on the divertor of ~2MW in two different scenarios: 1/ with combined LH and ICRH power after boronization at low X-point height (dX = 40 mm), 2/ with LH power only and high X-point height (dX = 120 mm).
In SOLEDGE2D simulations, target temperature profiles measured by Langmuir probe, as well as the radiated power measured by bolometry, were well reproduced with 3% of Oxygen as effective medium Z charge [Ciraolo]. The simulated asymmetry of O between the inner and outer targets is in qualitative agreement with UV measurements. The force balance analysis shows that friction dominates over thermal gradient forces at the inner target, while, at the outer target, the repelling thermal gradient forces dominate.
On the ITER-grade PFU, cracking and local melting have been observed for misaligned PFU. In addition, optical hot spots, which have been predicted to occur in ITER at the projection of the toroidal gaps on the subsequent PFU, have been observed experimentally, even for PFU aligned within specifications [Diez].
Finally, a He campaign has been run to investigate interactions between He plasmas and W PFC, in particular the formation of W fuzz [Tsitrone, Pegourie, Douai]. More than a hundred of 20-30 s pulses were repetitively performed in LSN. The conditions for W fuzz formation have been reached in the outer strike point area on the inertial W divertor (Einc > 20 eV, fluence > 1e24 He.m-2, Tsurf > 700 °C). Articulated Inspection Arm inspections before and after the He campaign have shown no macroscopic sign of surface modification. Post mortem analysis of the W components is ongoing to characterize the He induced nanostructures formed.
WEST phase 2 will start in autumn 2020, to address long pulse / high fluence operation on the newly manufactured ITER-grade actively cooled divertor, up to 10 MW/1000 s.
References:
[Diez], [Reux], [Goniche], [Colas], [Hillairet], [Maget], [Manas], [Liang], [Gaspar], [Ciraolo], [Tsitrone], [Pegourié], [Douai], this conference. [Martin 2008] J. Phys.: Conf. Ser. 123 012033Speaker: Jerome Bucalossi (CEA) -
KSTAR$^{1,2}$ program has been focused on resolving the key physics and engineering issues for ITER and future fusion reactors utilizing unique capabilities of KSTAR. First of all, a new advanced scenario was developed targeting steady-state operation based on the early diverting and heating during the ramp-up phase of plasma current and significant progress has been made in shape control to address the MA level of plasma current and stationary ITER-similar shape (ISS). It is demonstrated effective use of the H&CD with instrumented plasma control and shaping parameters became a key to access to the advanced operation scenarios such as high $β_p$, high $l_i$, high $q_{min}$, hybrid, internal transport barrier (ITB) and low $q_{95}$ operation. The examples of advanced scenarios are shown in Figure 1. The stationary ITB (fig 1a) is successfully reproduced with comparable confinement as H-mode level ($H_{89}$ ~ 2) both in limited and USN configuration, a low qmin scenario (fig 1b) is developed based on early diverting and delayed core heating approach and finally stable long-pulse H-mode operation (fig 1c) was extended upto 88 sec.
Recent KSTAR 3D experiments have focused on several ITER-relevant issues, such as divertor heat flux broadening in 3-row vs 2-row resonant magnetic perturbations (RMPs) on ELM-crash suppression, RMP-driven ELM-crash-suppression on ITER-like low $q_{95}$ (~3.2-3.4) and the characterization of ELM-crash suppression window in terms of normalized electron collisionality ($\nu^*_e$) and plasma toroidal rotation ($V_{tor}$) at pedestal top. Strong up-down asymmetry in 3-row configuration was identified and effect of the kink/anti-kink configuration was also clarified for ELM suppression in LSN plasmas. We have demonstrated the ISS-compatible RMP control in KSTAR using n=2, +900 phasing RMP, although the ISS has been more vulnerable to mode-locking than typical KSTAR configuration. A detailed study of the KSTAR database (where RMP configuration of all the discharges belongs to n=1, +900 phasing) showed that the ELM-crash suppression phase in KSTAR is in the range of 0.2 < $\nu^*_e$ < 1.2 and $V_{tor}$> 40 km/s. During the ELM suppression phase, coexistence of filamentary mode and smaller scale turbulent eddies at pedestal with broad-range of wave number ($k_θ$<1.1 $cm^{-1}$ and frequency (f<100 kHz) is identified by ECE imaging (ECEI) and strong energy exchange of the filamentary and turbulent modes was measured. The bicoherence analysis of the edge harmonic oscillations (EHOs) at natural ELM-less mode shows that there is a strong nonlinear interaction between EHOs, and the nonlinear interaction of EHOs has a significant effect on the ELM structure and dynamics.
Cross-validation between the advanced diagnostics and the modeling provides new insight on the basic transport process at KSTAR. For example, in the recent MHD-quiescent KSTAR plasmas non-diffusive avalanche-like electron heat transport events are observed by the ECEI and these observations have been successfully reproduced by gyrokinetic simulations indicating the broad range of spatial scales up to the minor radius. In addition, various studies utilizing the KSTAR fluctuation diagnostics demonstrated the importance of the turbulence characteristics in plasma rotation and confinement. The extensive study of the intrinsic rotation in Ohmic plasmas found a clear link between the counter-current toroidal rotation direction and the quasi coherent mode (QCM) which is measured by the Microwave Imaging (MIR). The improved confinement in the low rotation experiment was correlated with the suppression of the broadband (~200 kHz) ECEI fluctuations, and Collective Thomson Scattering provides a detailed measurement on the high-k density turbulence which is suppressed during the typical LH transition. Finally, strong interaction between fast-ion and EP driven MHD mode was identified with Fast ion $D_α$ (FIDA) diagnostics.
KSTAR provided unique demonstration on the performance of symmetric multiple Shattered Pellet Injections (SPIs) which is the main strategy of ITER for disruption mitigation. It was shown successfully the current quench rate changes proportionally as the time difference varies from several percent to several tens of percent of the thermal quench (TQ) duration (1~2 ms) and it was demonstrated that peak density was increased twice with dual SPIs compared with a single SPI and energy can be radiated when multiple SPIs are injected simultaneously, as planned in ITER.
Lastly, the research plan in near term will be addressed with the machine upgrades. KSTAR will focus on the development of the DEMO/ITER relevant operational scenario, i.e., high-beta steady-state operation with benign MHD activities which will require robust plasma control in strong shaping, control of MHD modes and thorough analysis of the fundamental physics processes. In these regards, KSTAR upgrades will includes extensive NBI (off-axis, 6MW) & RF (Helicon CD, 4MW) heating & current drive capabilities and the installation of new tungsten divertors with active cooling.References:
$^1$G.S. Lee et al, Nucl. Fusion 40 575 (2000) 575
$^2$H. K. Park et al, Nucl. Fusion 59 (2019) 112020 (13pp)Speaker: Si-Woo Yoon (National Fusion Research Institute) -
Construction of JT-60SA is progressing on schedule towards completion of assembly in March 2020 and the first plasma in September 2020. As of January 2020, manufacture and assembly of all the main tokamak components have been successfully completed satisfying technical requirements including functional performances and dimensional accuracies. Development of plasma actuators and diagnostics is also going well such as achievement of long sustainment of high energy intense negative ion beam. Commissioning of the power supply and the cryoplant has also satisfied requirements. Development of all the control systems and evaluation procedures of tokamak operation has been completed towards the Integrated Commissioning starting in April 2020, and plasma operation scenarios in the first plasma phase have been established. Unique importance of JT-60SA for H-mode and high-beta steady-state plasma research has been confirmed using advanced integrated modellings. These experiences of assembly, integrated commissioning and plasma operation of JT-60SA contribute to ITER risk mitigation and efficient implementation of ITER operation.
Introduction
The JT-60SA (R/a =3m/1.2m, Ip-max =5.5MA, heating power = 41MW x 100s) project [ref.1] was initiated in 2007 under the framework of the Broader Approach agreement by EU and Japan for early realization of fusion energy by conducting supportive and complementary works for ITER towards DEMO. Construction of JT-60SA is progressing successfully towards completion of assembly in Mar. 2020 and the first plasma in Sep. 2020 by the very close collaboration between QST in Japan, F4E in Europe, EU Voluntary Contributors and EUROfusion. The JT-60SA Research Plan [ref.2] covering its machine lifetime of ~ 20 years coordinated with ITER and DEMO schedules has been established with variety of plasma prediction using integrated modeling codes [ref.3]. Recently in Nov. 2019, a new collaboration arrangement between ITER and JT-60SA was signed which covers assembly, integrated commissioning and operation/experiments for finalization of ITER component design, risk mitigation and efficient implementation of ITER operation.Tokamak Construction
After the last IAEA FEC [ref.1], manufacture of all remaining tokamak components has been completed successfully including, superconducting Centre Solenoid (CS), thermal shields, Cryostat Top Lid, Cryolines, etc. As of Dec. 2019, the closure of the vacuum vessel has been accomplished, and the tokamak has been covered by the Cryostat Vessel Body (Fig.1). All the tokamak components have been assembled with excellent dimensional accuracy of ±1mm thanks to careful and smooth positioning using specially designed jigs, high accuracy measurement by Laser trackers, and fine adjustment utilizing sims. The magnetic field error is now expected below 10-4 Bt as designed. Commissioning operation of all large power supply systems, the Quench Protection Cirquit, the Switching Network Units and Super Conducting Magnet Power Supplies, has also been progressed with few residual commissioning activities still ongoing. The commissioning operation of the Cryoplant (equivalent refrigeration capacity of 9 kW at 4.4K) has also been successfully completed by satisfying the required performances.Plasma Hating Systems
For the heating systems, Positive-ion source NBs (85keV, 100sec, 20MW by 12 unit), Negative-ion source NBs (500keV, 100sec, 10MW by 2 units), ECH with multiple frequency Gyrotron (110GHz & 138GHz for 100s and 82GHz for 1 sec) and movable launchers, R&D have been steadily progressing and the targets of their development have been achieved. In particular, high energy intense hydrogen negative ion beams with 500 keV, 154 A/m2 for 118 s, which exceeds the requirement for JT-60SA, has been demonstrated by using a semi- cylindrical negative ion source with a three-stage accelerator. This result was realized by integration of i) stable voltage insulation by suppression of arching, ii) precise beam control and iii) stable negative ion production by maintaining the temperature balance in the negative ion source.
Integrated Commissioning and Control Systems
From April 2020 to Feb. 2021, the integrated commissioning is planned with the first plasma in Sep. 2020 and subsequent 5 months of machine commissioning with plasmas (‘the first plasma phase’). In this phase, the goal of plasma operation is to demonstrate equilibrium controllability of MA-class (<2.5MA) diverted plasmas with the full performance superconducting coil systems, 1.5MW ECH and upper divertor. For such tokamak operations, the Supervisory Control System and Data Acquisition System (SCSDAS) has been developed having the roles of (a) plant monitoring and machine state management, (b) discharge sequence management, (c)real-time plasma control, (d) device protection and human safety, (e) data storage, archive, and database management etc. For plasma controls, we have simulated operation scenarios of the first plasma phase with a newly developed advanced codes with control logics, such as pre-magnetic optimization scheme, plasma equilibrium control with iso-flux control method, control gain optimization method, and strategies for accessing stable operational regimes. Figure 2 shows the discharge scenario at Ip=2.5MA. EC wall cleaning operations and EC-assisted breakdown are also explored with optimized EC injection and toroidal / horizontal field. These results in the JT-60SA first plasma phase will contribute to highly valued subjects in ITER first plasma/subsequent operations.![Feedback-controlled plasma current wave form at Ip=2.5MA with upper divertor.
Scenario Development for ITER and DEMO and Risk Mitigation of ITER
After machine enhancements in 2021-2022, physics experiments will start in 2023 using in-vessel coils, particle fueling and pumping with lower divertor, enhanced diagnostics and high heating power of 26MW at Ip up to 5.5 MA. Toward this phase, variety of predictions of H-mode and high-beta steady-state plasmas covering divertor-SOL-Pedestal-Core have been progressing using advanced integrated modellings including newly-developed globally optimized steady-state transport solver GOTRESS coupled with turbulence models and pedestal models, Gyrokinetic theory based neural-network transport modeling DeKANIS, etc. These studies have confirmed the unique and important characteristics of JT-60SA (highly-shaped, high-beta, 500keV high energy ions, electron heating, controllable rotation etc.) for study of fusion plasma physics such as impacts of fast ions and plasma shape on microturbulence. As for operation scenario development of high-beta steady-state with controlled divertor heat load, an important result has been achieved using the integrated divertor code SONIC upgraded to treat multiple impurity species simultaneously. The result has shown that ‘mixture-seeding of Ar with small amount of Ne’ can keep the peak heat load below allowable 10MW/m2 together with smaller Ar concentration in the SOL and core plasmas than an Ar-only case. These studies have also confirmed significant roles of JT-60SA for ITER risk mitigation (disruption and ELM mitigation) including magnetic perturbation effect on both transient and stationary heat load, vertical displacement event, plasma response to massive gas injection, pedestal and ELM stability and control with Pellet and RMP.[ref.1] P. Barabaschi, Y. Kamada, H. Shirai and JT-60SA Intrgrated Project Team,
Nucl. Fusion 59 (2019) 112005.
[ref.2] JT-60SA Research Plan - Version 4.0, Sept. 2018,
http://www.jt60sa.org/pdfs/JT-60SA_Res_Plan.pdf
[ref.3] G. Giruzzi, M. Yoshida et al., Plasma Phys. Control. Fusion, 62 (2020) 014009.Speaker: Yutaka Kamada (National Institutes for Quantum and Radiological Science and Technology) -
A new era of predictive integrated modeling has begun. The successful validation of theory-based models of transport, MHD stability, heating and current drive, with tokamak measurements over the last 20 years, has laid the foundation for a new era where these models can be routinely used in a "predict first" approach to design and predict the outcomes of experiments on tokamaks today. The capability to predict the plasma confinement and core profiles with a quantified uncertainty, based on a multi-machine, international, database of experience, will provide confidence that a proposed discharge will remain within the operational limits of the tokamak. Developing this predictive capability for the first generation of burning plasma devices, beginning with ITER, and progressing to tokamak demonstration reactors, is a critical mission of fusion energy research. Major advances have been made using this predict first methodology. Extensive predictive modeling has informed the planning for the JET D-T campaign. This includes integrated modeling of JET hybrid regimes with newly upgraded heating sources, for various concentrations of deuterium D and tritium T. The self-consistent profiles of tungsten, ion and electron temperature, toroidal rotation and densities, have been predicted using theory-based turbulence and neoclassical transport models. The EPED model predicts it is possible to access the, higher pressure, super-H pedestal regime for JET achievable shapes. This prediction has been confirmed with DIII-D experiments. Super-H experiments on JET are planned. A new high accuracy neural network fit to the QuaLiKiz transport model has been completed, opening the way to time dependent predictions, at near real time speed, of complete tokamak discharges. Neural network fits to the TGLF and Multi-Mode models are progressing. The EAST tokamak is using predictive modeling to optimize the high bootstrap fraction regime for fully non-inductive operation and to plan future upgrades of power and current drive systems. A new integrated modeling workflow called TRIASSIC is being developed and tested on the KSTAR tokamak. Predictive modeling of CFETR is informing the design activity. ITER is using predictive modeling to simulate phases of the experimental operations plan. An overview of several of these recent advances will be presented, providing the integrated modeling foundations of experimental successes, as well as progress towards the goal of integrated predictive modeling for experimental design. Two examples, selected from the many advances in the prediction of tokamak experiments, are summarized in this synopsis.
1st example: The fast response of cold pulses due to impurity injection in tokamaks, with an inversion of the inward electron temperature pulse from decrease to increase, has long been argued to be inconsistent with a local transport paradigm [1]. The first demonstration that the cold pulse temperature response could be captured by a local turbulence transport model (TGLF [2]) was performed for the C-MOD tokamak [3]. Only electron and ion temperatures were predicted in these cases, with the density profile being evolved in a prescribed way. It was found that the inversion of the electron temperature pulse from decrease to increase was caused by the stabilization of the trapped electron mode (TEM) by the flattening of the electron density profile. In discharges where the TEM mode was not dominant there was no inversion in agreement with experiment. The transport model was then used in predict first method to simulate the cold pulse response [4] in the DIII-D tokamak. The very fast, high spatial resolution, density profile data on DIII-D confirmed the speed of the prescribed density response and the electron temperature response predictions were confirmed. The final step was to prove that the TGLF model could predict the fast density response to the impurity injection. This required adding the injected impurity density to the transport modeling. This integrated modeling was performed for experiments on the ASDEX Upgrade tokamak [5]. The predicted electron temperature response is compared with data in. It was found that the destabilization of the ion temperature gradient mode (ITG) by the transiently hollow impurity profile increased the speed of propagation of the electron density pulse into the core. Thus, the speed of the combined electron, ion, and impurity, temperature and density pulses were accurately modeled and new physics insights were discovered. This is a convincing proof that local turbulence transport can account for the paradoxical cold pulse phenomenon.
2nd example: The new upgrades to off-axis NBI current drive capability on DIII-D were preceded by state of the art integrated modeling [6] illustrated by the advanced tokamak predictions in. The profiles in Fig. 2 are a steady state self-consistent solution of the pedestal structure (height and width), core transport, MHD equilibrium and heating and current drive using validated theory-based models. An iterative high performance workflow IPS-FASTRAN was developed to find the integrated optimum solution [6]. Well validated theory-based models for MHD equilibrium (EFIT) and stability (DCON), turbulent transport (TGLF), pedestal structure (EPED1), neutral beam heating and current drive (NUBEAM) and electron cyclotron heating and current drive (TORAY-GA) were integrated. The IPS-FASTRAN modeling predictions have been confirmed with experiments showing good agreement that will be reported at the FEC 2020 conference. Verification of the accuracy of these predict first method simulations are a valuable test of the new capabilities. The same integrated modeling workflow is being used in the design of the CFETR, CAT Fusion Pilot Plant and SPARC tokamaks and to predict ITER plasmas.
This work was supported by the US Department of Energy under DE-FG02-95ER54309, DE-FC02-04ER54698, DE-SC0019736
[1] K. W. Gentle, W. L. Rowan, R. V. Bravenec, G. Cima, T. P. Crowley, et al., Phys. Rev. Lett. 74 (1995) 3620
[2] G. M. Staebler, J. E. Kinsey, and R. E. Waltz, Phys. Plasmas 14, (2007) 055909.
[3] P. Rodriguez-Fernandez, A. E. White, N. T. Howard, B. A. Grierson, G. M. Staebler, et al., Phys. Rev. Lett. 120 (2018) 075001.
[4] P. Rodriguez-Fernandez, A. E. White, N. T. Howard, B. A. Grierson, L. Zheng, et al., Phys. Plasmas 26, (2019) 062503.
[5] C. Angioni, E. Fable, F. Ryter, P. Rodriguez-Fernandez, T. Putterich, and the ASDEX Upgrade team, Nuclear Fusion 59 (2019) 106007.
[6] J. M. Park, J. R. Ferron, C. T. Holcomb, R. J. Buttery, W. M. Solomon, et al., Phys. Plasmas 25 (2018) 012506.Disclaimer-This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights. Reference herein to any specific commercial product, process, or service by trade name, trademark, manufacturer, or otherwise, does not necessarily constitute or imply its endorsement, recommendation, or favoring by the United States Government or any agency thereof. The views and opinions of authors expressed herein do not necessarily state or reflect those of the United States Government or any agency thereof.
Speaker: Gary Staebler (General Atomics)
-
-
-
Achieving ignition and high fusion yield in the laboratory is a central goal of the U.S. Inertial Confinement Fusion (ICF) Program. Three major and credible approaches are currently being pursued: laser indirect-drive (LID), laser direct-drive (LDD), and magnetic direct-drive (MDD). While the three approaches use very different means for driving a spherical or cylindrical implosion that can compress and heat a mass of deuterium-tritium (DT) fuel (by laser-generated radiation drive, direct laser irradiation, or direct magnetic acceleration, respectively), they share many common challenges, including how to efficiently couple the kinetic energy of the implosion to internal energy of the fuel at stagnation, and how to assemble the fuel at the necessary conditions, in pressure, temperature, and areal density to initiate self-heating and ignition.
Significant progress has been made in each of these approaches in recent ICF experiments on the National Ignition Facility (NIF) at Lawrence Livermore National Laboratory, the OMEGA laser facility at the Laboratory for Laser Energetics, and the Z pulsed power facility at Sandia National Laboratories. This includes progress in both advancing the absolute fusion output and, as importantly, improving our knowledge and understanding of the implosion behavior and causes of deviations from ideal or theoretical performance.
In this overview, we will review some of the major results from each facility, with a particular focus on recent advances in diagnostic measurement techniques and analysis that have improved our understanding of the states of the assembled fuel and confining shell at stagnation. Particular emphasis is placed on recent advances in determining the 3D spatial morphology and thermodynamic properties of the fusion fuel, including pressure, temperature, areal density, and mix, that determine the degree of alpha-particle production, confinement, and self-heating, and proximity to ignition.
As an example, Fig. 1 shows the hot spot temperatures and areal densities inferred from some of the principal LID experimental campaigns performed on the NIF since 2011. Improved performance can be observed as the implosion design evolved from the original low-adiabat† (a~1.6) CH ablator design (2011-2013), to a mid-adiabat (a~2.3) CH design (2013-2015), and then to a mid-adiabat (a~2.7) high-density-carbon (HDC) ablator design (2016-2019) [1-3]. The principal gains resulted from improved hydro-instability at higher adiabat, and improved energy coupling and implosion symmetry with HDC designs. As of today, the highest performing HDC implosions have achieved hot spot temperatures of ~4.5 keV, hot spot areal densities of ~0.3 g/cm2, hot spot pressures of ~350 Gbar, and fusion energy output of >50 kJ. At these conditions, the self-heating of the hot spot by alpha-particle deposition is estimated to be amplifying the total fusion output by a factor of ~2.5-3x. As can be seen in Fig. 1, these conditions are now quite close to the static self-heating boundary, dT/dt>0, where for a static hot spot at the time of peak compression, the time derivative of temperature is positive due to alpha-particle heating exceeding energy losses from radiation and thermal conduction. Whilst reaching this boundary will be a very noteworthy physics achievement, it is not a sufficient condition for ignition. For ignition, the alpha-heating rate must exceed all losses including mechanical work from expansion of the hot spot after peak compression. This condition, equivalent to the requirement d2T/dt2>0, depends on the shell confinement and necessitates on the order of a ~1 keV higher hot spot temperature [4,5].
In the LDD approach, remarkable progress has been made using a data-driven statistical mapping model to optimize input laser and target parameters that has led to a 3x increase in neutron yield over the past two years on OMEGA [6]. The best-performing spherical direct drive (SDD) implosion has a hydrodynamically-scaled predicted fusion yield in the 0.5 MJ range at NIF laser energy, approaching the burning plasma regime. Polar direct drive (PDD) implosions and laser-plasma coupling research are performed on OMEGA and the NIF for the LDD approach [7,8]. In the MDD approach, recent magnetized liner inertial fusion (MagLIF) experiments on Z have produced record yields and stagnation parameters through steady enhancements in the target initial magnetization, laser preheat energy delivered to the fuel, and electrical current delivered to the imploding shells [9]. In addition to record fusion performance, MagLIF implosions have also demonstrated significant potential for DT alpha-particle confinement and stopping via the strong magnetic fields entrained in the fuel [10,11], a necessary requirement for self-heating and ignition.
This work was performed under the auspices of Lawrence Livermore National Security, LLC (LLNS) under Contract DE-AC52-07NA27344, the Sandia National Laboratories, a multi-mission laboratory managed and operated by National Technology & Engineering Solutions of Sandia, LLC, a wholly owned subsidiary of Honeywell International Inc., for the U.S. DOE NNSA under contract DE-NA0003525, and upon work supported by the U.S. DOE NNSA under Award Number DE-NA0003856, the University of Rochester, and the New York State Energy Research and Development Authority.
† The adiabat is the ratio of the pressure of the DT fuel to the Fermi degenerate pressure.References
1. J. Lindl et al., Phys. Plasmas 21, 020501 (2014)
2. O. A. Hurricane et al., Nature 506, 343 (2014)
3. S. Le Pape et al., Phys. Rev. Lett. 120, 245003 (2018)
4. P. T. Springer et al., Nucl. Fusion 59, 032009 (2019)
5. P. K. Patel et al., Phys. Plasmas (submitted, 2020)
6. V. Gopalaswamy et al., Nature 565, 581 (2019)
7. F. J. Marshall et al., Phys. Plasmas 23, 012711 (2016)
8. P. B. Radha et al., Phys. Plasmas 23, 056305 (2016)
9. M. R. Gomez et al., manuscript in preparation (2020)
10. P. F. Schmit et al., Phys. Rev. Lett. 113, 155004 (2014)
11. P. F. Knapp et al., Phys. Plasmas 22, 056312 (2015)Speaker: Pravesh Patel (Lawrence Livermore National Laboratory) -
We present here recent highlights from Wendelstein 7-X (W7-X), the most advanced and largest stellarator in the world, in particular stable detachment with good particle exhaust, low impurity content, and energy confinement times exceeding 100 ms, maintained for tens of seconds, as well as proof that the reduction of neoclassical transport through magnetic field optimization is successful. W7-X, which has a magnetic field strength of 2.5 T and a plasma volume of 30 m$^3$, started operation in 2015 [1-5]. Following the installation of a full set of in-vessel components, in particular 10 passively cooled fine-grain graphite test divertor units, it was operated again in 2017 and 2018. Plasma pulses up to 100 s were successfully sustained [6], despite the lack of active cooling. Stable and complete detachment was achieved routinely. The pumping efficiency was initially relatively low [7-9] but it significantly improved later. Detachment with high pumping efficiency was achieved for up to 28 seconds at a heating power of 5 MW with a very low impurity content [10] (Figure 1), indicating control of divertor-heat-flux, plasma density, and impurity content, and giving confidence for reaching the foreseen high-performance, quasi-steady-state (30 minutes) discharges in the future [11]. The performance of the W7-X divertor, and the behavior and parameters of the edge- and scrape-off-layer plasma are now understood in quite some detail [eg. 12], thanks to measurements from a suite of diagnostics [eg. 13-17].
The earlier-reported stellarator triple-product record discharge [18] has now been shown to provide proof that the optimization for reduced neoclassical transport in W7-X was successful, Figure 2 [19]: The high temperature (appr. 3.5 keV for both ions and electrons in the center) and high hydrogen ion density (appr. 7x1019 m-3 in the core) were achieved with 5 MW of heating, and an energy confinement time of 0.22 s corresponding to about 1.4 times the energy confinement time expected from the ISS04 stellarator scaling [20]. For other, less optimized stellarators scaled to the W7-X size and magnetic field strength, similar plasma temperature and density profiles would have required significantly higher heating power to balance neoclassical transport, in particular in the mid-radius (strong gradient) region.
A number of discharges with similar performance to the triple-product-record discharge have since been achieved. These are generally characterized by core density peaking, and a reduction of turbulent density fluctuations. Without such turbulence reduction, the central ion temperature appears to be clamped to appr. 2 keV [21]. These findings are consistent with W7-X transport usually being dominated by ITG turbulence, but stabilized by strong density gradients in a so-called stability valley [22], as exemplified in Figure 3 for the W7-X standard configuration.
During turbulence-dominated phases, impurity confinement times are low (of order the energy confinement time) and no impurity accumulation is seen, but they can be very large if the turbulence is suppressed, and this then leads to impurity accumulation [23]. Recent findings from the Large Helical Device (LHD) show that ITG-dominated discharges readily mix hydrogen isotopes, whereas electron-scale (trapped-electron mode) turbulence does not [24]. It is tentatively concluded that a non-negligible amount of ITG turbulence is beneficial for impurity control as well as for fuel (isotope) exchange and helium exhaust in a stellarator fusion reactor, whereas too much ITG turbulence could potentially clamp the ion temperature below the burn point. These and other recent results [see eg. 25-31] will be put into the context of future goals for the W7-X, the world stellarator program, and the magnetic confinement fusion program in general.
References
1 T. Klinger et al, Plasma Phys. Controlled Fusion 59(1) 014018 (2017)
2 H.-S. Bosch et al, Nuclear Fusion 57, 116015 (2017)
3 R. C. Wolf et al, Nuclear Fusion 57 102020 (2017)
[4] T. Sunn Pedersen et al, Physics of Plasmas 24 055503 (2017)
[5] A. Dinklage et al, Nature Physics (2018) https://doi.org/10.1038/s41567-018-0141-9
[6] T. Klinger et al, Nuclear Fusion 59 112004 (2019)
[7] T. Sunn Pedersen et al, Nuclear Fusion 59 096014 (2019)
[8] D. Zhang et al, Phys. Rev. Lett. 123, 025002 (2019)
[9] D. Zhang et al, this conference (2020)
[10] M. Jakubowski et al, submitted to Phys. Rev. Lett. (2020)
[11] M. Jakubowski et al, this conference (2020)
[12] F. Reimold et al, this conference (2020)
[13] V. Perseo et al, Nuclear Fusion 59 124003 (2019)
[14] V. Perseo et al, this conference (2020)
[15] T. Barbui et al, JINST 14 C07014 (2019)
[16] C. Killer et al, Plasma Phys. Control. Fusion 61 125014 (2019)
[17] C. Killer et al, this conference (2020)
[18] T. Sunn Pedersen et al, Plasma Phys. Control. Fusion 61 014035 (2019)
[19] C. Beidler et al, in preparation (2020)
[20] H. Yamada et al Nuclear Fusion 45 1684 (2005)
[21] M. Beurskens et al, this conference (2020)
[22] J. A. Alcusón et al, Plasma Phys. Control. Fusion 62 035005 (2020)
[23] A. Langenberg et al, this conference (2020)
[24] K. Ida et al, Phys. Rev. Lett. 124, 025002 (2020)
[25] J. Geiger et al, this conference (2020)
[26] Y. Feng et al, this conference (2020)
[27] G. Fuchert et al, this conference (2020)
[28] K. Aleynikova et al, this conference (2020)
[29] H. Laqua et al, this conference (2020)
[30] S. Lazerson et al, this conference (2020)
[31] A. Dinklage et al, this conference (2020)Speaker: Prof. Thomas Sunn Pedersen (Max Planck Institute for Plasma Physics) -
In the recent deuterium experiment on the Large Helical Device (LHD), we have succeeded to expand the temperature domain to higher region both in electron and ion temperatures as shown by the red region in Fig.1. We found a clear isotope effect in the formation of Internal Transport Barrier (ITB) in high temperature plasmas. In the deuterium plasmas, we have also succeeded to realize the formation of the Edge Transport Barrier (ETB) and the divertor detachment, simultaneously. It is found that the Resonant Magnetic Perturbation (RMP) has an important role in the simultaneous formation of the ETB and the detachment. A new technique to measure the hydrogen isotope fraction was developed in LHD in order to investigate the behavior of isotope mixed fuel ions. The technique revealed that the non-mixing state and the mixing state of hydrogen isotopes can be realized in plasmas.
One of the objective of deuterium experiment is the achievement of the reactor relevant high temperature plasmas in order to investigate the problems arising in the helical type fusion reactor. As shown in Fig.1, the ion temperature of 10keV was achieved at the electron temperature of around 4keV. This was reported at the last IAEA-FEC(FEC2018). The operational domain was extended to the direction of higher electron temperature range. The electron temperature of 6.6keV was achieved keeping the ion temperature of 10keV. For lower ion temperature plasmas (Ti0=~7keV), the electron temperature range was extended over ~12keV. It is found that the extension in higher electron temperature region of high ion temperature plasmas is an effective way to suppress the Energetic particle driven InterChange mode (EIC), which often prevents the access of high ion temperature plasmas in LHD. This is preferable feature for the future reactor scenario development 1. The isotope effect is a long underlying mystery in plasma physics because the most of experimental observations in tokamak plasmas show favorable effect by the ion mass on the energy confinement time (tau_E) while the theoretical prediction based on the gyro-Bohm model shows unfavorable effect by the mass, i.e., tau_E~ M^-1/2. A series of dedicated experiment using dimensionally similar L-mode hydrogen and deuterium plasmas in LHD showed that the energy confinement time scaling for L-mode plasmas had a non-significant dependence on mass (M^0.01) 2. On the other hand, a clear isotope effect was observed for the formation of Internal Transport Barrier (ITB) in LHD as shown in Fig.2. Here, the profile gain factor G1.0 is defined as
and TLref is the expected temperature profile for L-mode plasmas with same heating condition 3. As shown in the figure, the ITB intensity, i.e., G1.0, is clearly larger for deuterium at the density range below 2x10^19 [m^-3].The realization of both divertor heat load mitigation and the good core confinement property is an important issue in developing future reactor scenarios. For divertor detached deuterium plasmas which are realized by the RMP application, we have succeeded to improve the confinement of 38% with a formation of ETB. Thus, we call this improved confinement mode as the RMP induced H-mode 4. As shown in Fig.3, a steep pressure gradient was formed in deuterium plasmas at the edge region inside the m/n=1/1 island which was produced by the RMP. This formation of the steep edge pressure gradient with the detachment has never been observed in hydrogen plasmas. The thermal transport analysis based on TASK3D-code shows significant reduction in the thermal diffusivity at the edge region (r/a=~0.8) which indicates the formation of ETB.
In the future fusion reactor, the deuterium (D) and tritium (T) ions co-exist in plasmas and are assumed to be uniformly mixed with the ratio of D/T=1. On the other hand, it is not clear whether this assumption is always valid. To clarify the assumption, experiments to investigate the isotope mixing state were performed using Deuterium and Hydrogen mixture plasmas. It was found that the non-mixing state of isotopes can be realized when the centrally fuelled ion species are different from the ion species fuelled at the edge 5. Figure 4 shows a typical example of the non-mixing state where the tangential NBI fuels Hydrogen ions in the central region while the Deuterium is supplied at the peripheral region by a wall recycling (black symbols and curves). It was found this state can be changed to mixing state by an injection of ice pellet of hydrogen isotopes, i.e. either of Hydrogen or Deuterium. It was also found that a formation of hollow density profile due to the pellet injection plays an important role for changing the state.
1 H. Takahashi et al., this conference
2 H. Yamada et al., Phys. Rev. Lett. 123 (2019) 185001
3 T. Kobayashi et al., this conference
4 M. Kobayashi et al., this conference
5 K. Ida et al., this conferenceSpeaker: Prof. Masaki Osakabe (NIFS) -
Using its unique flexibility and advanced plasma diagnostics, the TJ-II stellarator is contributing to the understanding and solution of critical challenges in fusion plasmas. Next, we highlight some of the most relevant recent results in the framework of its research programme.
Towards validation of gyrokinetic and neoclassical simulations. Aiming at the validation of the instability properties predicted by gyrokinetic (GK) simulations and of the electrostatic potential variations on the flux surface, φ1, calculated by neoclassical (NC) codes, dedicated experiments have been carried out in TJ-II for a systematic characterization of turbulence wavenumber spectra and perpendicular rotation velocity measured by Doppler Reflectometry (DR) at poloidally separated positions on the same flux surface [1]. Poloidal asymmetries in the intensity of the wave number spectrum that depend on plasma conditions have been characterized and compared with global linear GK simulations by the code EUTERPE. Model and experiment qualitatively agree in the radial dependence of the turbulence intensity, in the turbulence dispersion relation and in showing a poloidal asymmetry that depends on the magnetic configuration. Recent experiments exploring configurations with different magnetic ripple have shown a reduction in the turbulence asymmetry at configurations with reduced ripple. Besides, the influence of base ion mass has been investigated in hydrogen and deuterium plasmas. The ion mass in TJ-II plasmas does not affect the properties of the turbulence, neither the amplitude nor the spectral shape or the poloidal asymmetry. The lack of dependence of the turbulence spectrum on the ion mass is also found in GK simulations. Model validation will also benefit from the effort of verification of GK simulations in different computational domains [2] as well as from the application of the recently developed GK code stella to multispecies turbulent transport calculations in TJ-II [3].
Poloidal asymmetries in radial electric field, Er, are found that depend on the plasma collisionality. These results have been compared with the contribution to Er arising from −φ'1 as calculated with the NC version of the code EUTERPE. These results show variations in Er comparable in size to those found in the experiments, but there is a disagreement regarding the sign of the Er correction. Recent simulations performed with the newly developed NC code KNOSOS [4] show that the effect of kinetic electrons on φ1 has to be taken into account due to the strong Te dependence of the electron contribution to φ1 when the electrons are in the 1/ν regime [Fig. 1]. KNOSOS (KiNetic Orbit-averaging SOlver for Stellarators) is a freely available, open-source code that calculates neoclassical transport in low-collisionality plasmas of three-dimensional magnetic confinement devices by solving the radially local drift-kinetic and quasineutrality equations.
Experiments in TJ-II with cryogenic and TESPEL pellets show that post-injection particle radial redistributions can be understood qualitatively from neoclassical predictions while also providing a means to benchmark the HPI2 code [5]. TJ-II measurements provide 2-D maps of plasma potential and fluctuations to address the question of how hollow density profiles, created by pellets, affect turbulent transport. Fluctuations are stronger in the negative density gradient than in the positive one in consistency with TEM linear simulations [6].
Towards the validation of fast ion induced stabilization and the identification of Alfvén Eigenmode actuators. An ambitious research programme is in progress to investigate the relation between zonal structures and Alfvén eigenmodes (AE) and its role on the nonlinear dynamics of AEs and transport as well as to develop and demonstrate AE control strategies using ECRH and ECCD in TJ-II. The unique TJ-II experimental set-up using a dual HIBP has shown that, in some conditions, long range correlations (LRC) are detected both at the AE frequencies and low frequencies (<10 kHz) [Fig. 2]. LRC are observed in plasma potential fluctuations but not in density fluctuations as expected in zonal flow structures [Fig. 2]. It is an open question whether those zonal structures are directly driven by fast particle effects or/and are the consequence of plasma scenarios with reduced damping of zonal flows. Experiments in TJ-II have demonstrated the effectiveness of ECRH and ECCD actuators to modify AE activity [7].
Towards the characterization of the interaction between neoclassical and turbulent transport mechanisms. We have investigated the impact of Er on turbulence propagation and the coupling between the plasma edge and the scrape-off layer (SOL) during electron–ion root transitions where Er is changed in a controlled manner from positive to negative values. It is shown that Er does not only affects the radial turbulence correlation length but it is also capable of reducing the propagation of turbulence from the edge into the SOL. This result was obtained using a technique based on the transfer entropy, which quantifies the propagation of information [8]. These observations are highly relevant for the understanding of the mechanisms that determine the SOL width.
The interplay between of NC radial electric fields, Reynolds stress gradients and LRC has been investigated in different plasma scenarios in TJ-II. Turbulent driven acceleration alone cannot explain the dynamics of zonal flows whose radial width is affected by the isotope mass [9]. These results are in line with the expectation that the interplay between turbulent and neoclassical mechanisms is an important ingredient of the dynamics of edge zonal flows.
Power exhaust physics: liquid metals. Solid and liquid samples of Li/LiSn/Sn, in a Capillary Porous System (CPS) arrangement, have been exposed to the edge plasma [10]. A simple 1D model was applied to the data, allowing for the evaluation of the kinetic energy (Ek) of ejected atomic species while their residence time at the edge was determined by monitoring the ratio of first ion/neutral emission light intensities. A clear evolution of Ek with sample temperature was deduced for Li atoms, this being associated to the different relative contributions of sputtered/evaporated atoms. SnI emission into the plasma has also been measured with radial and toroidal resolution. The deduced mean free paths for the ejected Sn atoms under sputtering conditions (low T) imply unrealistic high energies if the bibliographic data for the ionization rate constant of Sn are assumed. For the LiSn case, Li as well as Sn emissions were simultaneously detected and analysed. Plasmas under cut-off (collapsing) conditions were also investigated to check for the sensitivity of the recorder line intensities and ionization rates to the edge electron temperature.
[1] T. Estrada et al., Nuclear Fusion 59, 076021 (2019)
[2] E. Sánchez et al. Gyrokinetic simulations in stellarators using different computational domains, 28th IAEA Nice 2020
[3] J. M. García-Regaña et al., Turbulent transport of impurities in 3D devices, 28th IAEA Conf. Nice 2020.
[4] J. L. Velasco et al.,KNOSOS, a fast neoclassical code for three-dimensional configurations, 28th IAEA Nice 2020
[5] K. McCarthy et al., Pellet studies in TJ-II, 28th IAEA Nice 2020
[6] A. Melnikov et al., 2-D mapping of fluctuations and plasma profiles in TJ-II, 28th IAEA Nice 2020.
[7] A. Cappa et al., AE control strategies in TJ-II, 28th IAEA Conf., Nice 2020.
[8] G. Grenfell et al., Nucl. Fusion 60 014001 (2020)
[9] R. Gerrú et al., Nucl. Fusion 59 106054 (2019)
[10] F. Tabarés et al., Liquid metal studies in TJ-II, 28th IAEA Conf., Nice 2020Speaker: Prof. Carlos Hidalgo (CIEMAT)
-
-
-
-
-
For the first time, experiments on the DIII-D tokamak have demonstrated electron cyclotron current drive (ECCD) with more than double the efficiency of the conventional outside launch by using a novel top launch geometry (figure 1), as predicted by linear ray tracing and quasi-linear Fokker-Planck simulations. Studies have shown that off-axis current drive is a requirement for a steady-state reactor in the Advanced Tokamak (AT) regime$^{1,2}$; however, driving current off-axis efficiently remains a challenge. Launching electron cyclotron waves from the high field side of the plasma, but the low field side of the resonance, with large toroidal steering in a plane nearly parallel to the resonance layer (illustrated in figure 2) is found to greatly increase the ECCD efficiency at mid-radii compared to conventional outside launch. The higher ECCD efficiency is due to 1) selective EC wave damping on higher v$_{||}$ electrons, and 2) longer absorption path lengths to compensate for inherently weaker absorption at higher v$_{||}$. DIII-D experiments using a prototype top launch system with a fixed mirror have established these two tenets through scanning v$_{||}$ of the wave-particle interaction by varying the magnetic field B$_T$. Power deposition measurements show that the absorbed EC power decreases for higher v$_{||}$ interaction (lower B$_T$), giving rise to a “sweet spot” (optimal B$_T$) for maximum ECCD efficiency at $\rho$~0.5 (figure 3) where the higher current drive efficiency for higher v$_{||}$ is balanced by sufficient absorption. Simulations of ‘top launch’ ECCD for FNSF, DEMO and CFETR support it as an improved efficiency off-axis current drive technique for future fusion reactors$^{3-5}$.
Top launch ECCD with a long wave-electron interaction zone and a large Doppler shift ensures strong damping on tail electrons leading to higher ECCD. Top launch ECH experiments have been done previously on TCV using radial launch and 3$^{rd}$ harmonic X-mode to heat high density plasmas; the top launch experiments on DIII-D have the different goal of efficient off-axis current drive, and the launch scheme has been uniquely optimized for this purpose. As illustrated in figure 2, EC wave from this top launch 1) propagates nearly parallel to the resonance plane and only gradually approaches the resonance, resulting in a longer absorption path; 2) suffers less trapping effects by being on HFS of the axis; and 3) allows a larger Doppler shift, thus interacting with higher energy (less collisional) electrons. To experimentally validate and characterize this approach, a prototype top launch system is installed on DIII-D with a fixed mirror angle utilizing 2$^{nd}$ harmonic damping of either a single 110GHz or 117.5GHz gyrotron with injected power between 0.5-0.6MW.The longer absorption path of top launch predicted by the TORAY ray-tracing code is measured by modulating the ECCD power and observing the electron temperature oscillations with an ECE radiometer. The measured power deposition location generally agrees with TORAY. The vertical path is verified via comparison of X-mode and O-mode deposition, where the predicted location shift between X and O (reflecting the different vertical paths due to the different absorptions) is confirmed. A much longer (i.e., three times) absorption zone for top launch ECCD compared to outside launch is also measured, consistent with TORAY. The longer interaction zone usually results in a broader deposition profile when mapped onto $\rho$ space but not always. The more parallel the EC ray is to the flux surface, the narrower the deposition profile.
Selective damping on electrons with different v$_{||}$ via top launch ECCD geometry is evidenced by the reduced absorption measured with lower B$_T$ in DIII-D experiments, as predicted by TORAY. The cold resonance moves to higher v$_{||}$ at lower Bt and the wave-electron interaction follows. Damping on tail electrons that are less collisional and drive current more efficiently is crucial for high ECCD efficiency; however, because the electron population decreases with increasing energy, the total absorption can drop far below 100%. Reduced total absorption at extreme low B$_T$ (i.e., high v$_{||}$) is observed in both L-mode and H-mode (figure 3(a)) plasmas in DIII-D.
The highest top launch ECCD efficiency is predicted and achieved when balancing higher v$_{||}$ interaction and sufficient total absorption. The ECCD profile is determined from the change in the magnetic field pitch angles measured by motional Stark effect (MSE) polarimetry. As illustrated in figure 3(b), at high Bt (low v$_{||}$) the ECCD efficiency is low despite full absorption. The measured ECCD efficiency increases with decreasing B$_T$ (increasing v$_{||}$) until the curve rolls over when too little wave energy is absorbed by too few high v|| electrons. In H-mode plasmas, the ‘sweet spot’ for highest top launch ECCD is predicted and measured at Bt~1.55T, where the driven ECCD at $\rho$~0.5 is double for top launch compared to outside launch (figure 1), consistent with the predictions from TORAY and quasi-linear Fokker-Planck code CQL3D.
Studies are underway to evaluate whether the high-beta, steady-state goals of the AT program on DIII-D can be achieved using four top-launch gyrotrons and four outside-launch gyrotrons. Initial FASTRAN simulations with self-consistent transport/pedestal/current profile modeling shows that 3MW top launch can drive as much ECCD at $\rho$~0.65 as 6+ MW outside launch in the "high qmin" AT regime owing to the near doubling in ECCD efficiency, allowing access to the highest stable $\beta_N$ (~4.5) with a non-inductive current fraction of 1. These results suggest that the combination of 3 MW top launch and 3 MW outside launch is a reasonable optimum to achieve the high-beta, steady-state goal of the DIII-D AT program.
Top launch ECCD is a promising off-axis current drive technique for future fusion reactors. Top launch ECCD shares the same reactor-relevant features of conventional outside launch ECCD, such as easy coupling to the plasma, no near-plasma antenna, and small port requirements, along with long experience in gyrotron development. Modeling for FNSF-AT shows >50% higher off-axis current drive efficiency for top launch ECCD compared to outside launch$^3$, similar to the predictions for DEMO$^4$. Greater than 35% improvement in ECCD at $\rho$~0.5 has already been found in modeling the CFETR baseline$^5$, reaching a current drive figure of merit of $\gamma$~0.16x10$^{20}$ A/m$^2$W for 14.5keV; or a dimensionless current drive efficiency of $\xi$~0.37. The experimental demonstration of doubling off-axis ECCD on DIII-D and the great enhancement found in simulations of FNSF-AT, DEMO and CFETR strongly support top launch ECCD as an exciting reactor-relevant and efficient off-axis current drive technique.
This work was supported in part by the US Department of Energy under DE-FC02-04ER54698, DE-FG02-97ER54415, DE-AC52-07NA27344, DE-AC05-00OR22725, and DE-SC0019352.
$^1$F. Najmabadi, et al, FED 38 (1997) 3; $^2$F. Najmabadi, et al, FED 80 (2006) 3; $^3$R. Prater, et al, APS-DPP (2012); $^4$E. Poli, et al, NF 53 (2013) 013011; $^5$Xi Chen, et al., EPJ Web of Conferences, 203* (2019) 01004Speaker: Xi Chen (General Atomics) -
Future DT operation in ITER and DEMO will face a significant number of challenges. From the physics point of view, the change from DD to DT plasmas is poorly understood. There are indications that the core confinement, the ELM behavior, the pedestal confinement and the Scrape of Layer behavior can significantly change from DD to DT. From the operational point of view, scenarios with high enough alpha power production are essential to demonstrate that efficient electrical power generation is possible but this also requires techniques for power exhaust, ELM and impurity sources and accumulation control in the presence of enough additional input heating power and a metallic wall environment.
In order to address the relatively scarce knowledge of DT plasmas and in support of safe future DT operations, JET has developed a scientific program in DD, TT and DT with the aim of supporting and minimizing the risks of the transition from DD to DT plasma operation in ITER 1. Such program helps to understand and document the physics characteristics of DT plasmas but also to provide scenario solutions integrating several key aspects such as core transport, impurity accumulation avoidance, power exhaust, enough fusion power generation to reveal alpha particle effects and potential difficulties for operation.
To this end, JET has gone through an upgrade of diagnostics and heating systems, in particular, Neutral Beam Injection (NBI) input power, which has delivered the record of 32MW. With such power, high performance scenarios have been developed following two main routes, i.e., the baseline scenario with $q_{95}$~3, β$_N$~2, H$_{98}$(y,2) ~1 at high current and magnetic field and the hybrid scenario with $q_{95}$~4, β$_N$~2.4, H$_{98}$(y,2) ~1.3 at reduced current and low central magnetic shear.
Compared to previous campaigns [2], a new baseline JET DD neutron rate record in the ITER Like Wall (ILW), R$_{NT}$=4.1x10$^{16}s^{-1}$, has been attained as shown in figure 1. This high performance was possible at current Ip=3.5MA and magnetic field B$_T$=3.35T through a complex non-linear interaction between edge and core plasma regions. From the pedestal side, high frequency ELM’s help to flush impurities. This is obtained by combining 50%/50% particle source with pacing pellets and neutral gas puff as it allows simultaneously good confinement and impurity flushing. The pedestal density is relatively low for such a high current, ~4x10$^{19}m^{-3}$, but the strong peaking, which is significantly high in these conditions of low core NBI fueling, leads to a core density ~1x10$^{20}m^{-3}$. Rotation and its shear are high both for the core and the pedestal with a Mach number reaching 0.6 at mid-radius. The ExB shearing of such plasmas, which usually are in the ITG regime, have been shown to be important for the decoupling of ions and electrons [3], with Ti/Te~1.7 in the best performing discharge at H$_{98}$(y,2)=1.0, and for increasing the density peaking through an increase of the inward pinch [4]. The radiation power is mostly localized in the low field side pedestal for the whole discharge and never penetrates in the inner core, which allows long and stable flat-top pulses up to 5s.
This new baseline scenario at high Ip and with relatively low density and high temperature at the pedestal offers an attractive alternative to hybrid scenarios as no specific q profile tailoring or core improved confinement is required in order to obtain high neutron rate. Although the Greenwald fraction is moderate, G$_{fr}$=0.65, a key point is the density peaking, which provides core densities compatible with high fusion power. Furthermore, the ion pedestal temperature, 1.3keV, provides a route for avoiding W accumulation through neoclassical screening. The relevance of this scenario for ITER is strong for the long pulse operation program [5] as it can soften some of the constraints for the hybrid scenario in terms of off-axis current.
The hybrid route has also stablished a new neutron record generation compared to previous campaigns, R$_{NT}$=4.7x10$^{16}s^{-1}$. This has been obtained in conditions of reduced β$_N$=2.4 by working at higher magnetic field, i.e. 3.45T and $q_{95}$=4.5 rather than 2.8T used in previous campaigns with $q_{95}$=4.0. Actually, temperature and density profiles are essentially identical for both magnetic field showing that, at least in this regime, high confinement appears to be largely independent of the magnetic field. However, it has been also obtained that β$_p$≥1 seems to be necessary to get good confinement. Such hybrid plasmas are type I ELMy H-modes with Ti/Te ~1.5. They are prone to suffer from impurity accumulation when the low gas used in order to obtain high confinement leads short phases of too low ELM frequency. Further techniques, such as pacing pellets, will be used to maintain an adequate level of ELM frequency, and hence impurity flushing, but at reduced gas injection.
A key element that has significantly improved the robustness of both scenarios is the use of real time control techniques which have been applied for ELM, fueling and impurity accumulation control. With such techniques, the disruptivity has been significantly decreased to ~9% for the hybrid and to ~25% for the baseline when previously reached 60%.
An extensive exercise of ‘predict first’ modelling has been carried out with the aim of predicting the DT fusion performance expected in the future JET-DT campaign. As a first step, DT equivalent fusion power from recent plasmas at high power has been compared to past predictions from lower power discharges showing a reasonable agreement when using validated models, such as TGLF or QuaLiKiz, for core transport [6,7].
DT extrapolations to the maximum available power at JET show that 11-16MW of fusion power are possible both for the baseline and hybrid routes. In particular, favorable core isotope effects are found for the hybrid scenario in conditions of low turbulent transport, such as strong rotating and high core thermal and fast ion beta plasmas.
Finally, in baseline plasmas, electron heating by alpha power can be dominant when ICRH is temporary removed. Such characteristic can be used to demonstrate alpha heating in ITER relevant conditions for the first time. These predictions will be compared to TT and DT plasmas when they will be obtained in the upcoming campaigns at JET.
References
- E. Joffrin et al 2019 Nucl. Fusion 59 112021
- L. Garzotti et al 2019 Nucl. Fusion 59 076037
- Hyun-Tae Kim et al 2018 Nucl. Fusion 58 036020
- J Garcia et al 2019 Plasma Phys. Control. Fusion 61 104002
- A.R. Polevoi et al 2015 Nucl. Fusion 55 063019
- J. Garcia et al 2019 Nucl. Fusion 59 086047
- F. Casson et al., submitted to Nucl. Fus.
Speaker: Jeronimo Garcia (CEA IRFM) -
Experiments on DIII-D support a new approach, confirmed by transport modeling, to achieving Q=10 in ITER using a scenario with low plasma current (~ 8 MA), high $\beta_{\rm p}$, and line-averaged Greenwald fraction ($f_{\rm Gw}$) above 1. At 8 MA the disruption risk and ELM challenge are greatly reduced, with the possibility that uncontrolled ELMs may be acceptable$^1$. Due to the need of sufficient fusion power and low plasma current, this approach requires high density with $f_{\rm Gw}$>1.0 simultaneous with high confinement quality ($H_{\rm 98y2}$>1). Using impurity injection, the recent DIII-D experiments achieve and maintain these simultaneous conditions. Previously, high $\beta_{\rm p}$ plasmas with $f_{\rm Gw}$ up to ~1.0 and $H_{\rm 98y2}$>1 were obtained in JT-60U, albeit transiently and usually operated at low absolute density$^{2,3}$, which is not favorable to reactor plasma. For the first time in a tokamak, experiments demonstrate that a stationary ITB at large radius ($\rho$~0.7) is compatible with $H_{\rm 98y2}$>1, at reactor-level absolute density ($n_{\rm e0}>1.0×10^{20}\;m^{-3}$), $f_{\rm Gw}$>1, and reactor-relevant $q_{95}$ as well (fig. 1). Such ITB is a key feature of the ITER 8 MA Q=10 modeling. Comparison between the experimental DIII-D profiles and the predicted ITER profiles shows also a good match of ITB location and profile shape (fig. 2). The DIII-D experiments confirm that the high density ITB in ITER modeling is achievable at similar $q_{95}$ using the high $\beta_{\rm p}$ scenario.
For the first time, neon injection is observed to trigger the formation of a large radius ITB in the density channel at reactor level density on DIII-D, which provides an effective experimental approach to achieve line-averaged Greenwald fraction well above 1. Pedestal density feedback control is used in the experiment. Therefore, the pedestal density is kept below the Greenwald limit, e.g. $f_{\rm Gw,ped}$<0.7. The large radius ITB strongly elevates the plasma density inside the pedestal. The density profile is fairly flat inside the ITB with $n_{\rm e0}>1.0×10^{20}\;m^{-3}$ (fig. 2). These high $\beta_{\rm p}$ experiments on DIII-D also show that the ITB is sustained long after (>8×the energy confinement time, $\tau_{\rm E}$) the neon injection is turned off. Impurity transport modeling based on the experimental data shows that neon provides electron source at $\rho$~0.75, which is the location where the foot of the ITB emerges. One of the possible mechanisms for the trigger of ITB formation could be that the important electron source creates a seed of local density gradient. The locally increased density gradient strengthens the effect of $\alpha$-stabilization of turbulence by reducing the non-adiabaticity of electron response$^4$, and starts a positive feedback leading to ITB formation. The same technique may also work for ITER high $\beta_{\rm p}$ plasma, when higher Z impurity is employed.
Using this technique, new high $\beta_p$ experiments on DIII-D demonstrate the stationary large radius ITB at reactor level density and reactor relevant $q_{95}$. Fig. 1 shows that the stationary phases of $f_{\rm Gw}$~1.0-1.1 and $f_{\rm Gw}$~1.3 at $q_{95}$~8 are sustained for 21 and 8 $\tau_{\rm E}$’s, respectively. These plasmas have even higher density than ITER at its 9 MA Greenwald limit. With ITB, the energy confinement is well above standard H-mode in these plasmas, e.g. $H_{\rm 98y2}$ up to 1.4. The degradation of confinement in one of the discharges shown in fig. 1 ($f_{\rm Gw}$~1.3 case) is due to excessive neon injection causing very high core radiation. The averaged neon injection rates are $1.3×10^{20}$ /s ($f_{\rm Gw}$~1.3) and $4.8×10^{19}$ /s ($f_{\rm Gw}$~1.1). With further optimization of the impurity injection waveform, sustained high confinement is also expected in the $f_{\rm Gw}$~1.3 case. These high $\beta$ ($\beta_{\rm N}$≤3.5, $\beta_{\rm p}$≤2.7) plasmas have $q_{\rm min}$>2.0 and quiet MHD behavior. Meanwhile, a non-inductive current fraction up to 0.9 is achieved simultaneously in the high density phase.
1D Transport modeling suggests the goal of Q=10 on ITER can be achieved at low plasma current (~ 8 MA) using the high $\beta_{\rm p}$ scenario. The simulations are performed using the STEP module in the OMFIT framework, integrating sub-modules for transport, heating, current drive and equilibrium calculations. Compared to previous ITER simulation works, the innovative features in this work include: 1) Plasma density, temperature, current profiles and equilibrium are all evolved in the simulations, while pedestal heights for density and pressure are prescribed to values slightly below the Greenwald limit (for density) and EPED prediction (for pressure); 2) Use TGLF model for transport prediction; 3) In TGYRO, E×B shear effect on turbulence suppression is turned off. Simulations use ITER “Day One” heating and current drive power: NBI≤33 MW, EC≤20 MW. Calculations predict that the following parameters can be achieved: Q=9.5±2.5, $I{\rm p}$=7.85±0.35 MA, $q_{95}$=7.54±0.39, $H_{\rm 98y2}$=1.74±0.13, $f_{\rm Gw}$=1.48±0.13, $f_{\rm NI}$≥98% and $P_{\rm fus}$=350±50 MW. A large radius ITB is shown in both temperature and density channels for electron and ion species (fig. 2).
Density profiles experimentally achieved on DIII-D match the density profiles simulated for ITER Q=10 in shape (ITB radius) and absolute core density value (fig. 2), while the core confinement quality is maintained well above standard H-mode levels. Although the experimental electron temperature is low compared with ITER simulation results, its shape is still a very good match to the ITER simulation results, if a multiplier of 9 is applied to the DIII-D data. The results confirm that the required large radius ITB at $f_{\rm Gw}$ above 1 and reactor level density in ITER Q=10 modeling is achievable experimentally at similar q95 using the high $\beta_{\rm p}$ scenario. These new high $\beta_{\rm p}$ experiments on DIII-D strongly support the ITER Q=10 simulations and pave the avenue to a new low plasma current (~ 8 MA) approach for ITER’s Q=10 Goal.
This work was supported in part by the US Department of Energy under DE-FC02-04ER54698 and DE-SC0010685.
$^1$R. A. Pitts, et al., Nucl. Mater. Energy, 20 (2019) 100696
$^2$R. C. Wolf, Plasma Phys. Control. Fusion, 45 (2003) R1
$^3$N. Oyama and the JT-60 Team, Nucl. Fusion, 49 (2009) 104007
$^4$M. Kotschenreuther, et al., “Regimes of weak ITG/TEM modes for transport barriers without velocity shear”, UP10.00020, 61st APS-DPP, Oct 21-25, 2019, Fort Lauderdale, USSpeaker: Siye Ding (ORAU/ASIPP) -
Recent EAST experiment has successfully demonstrated long pulse steady-state high plasma performance scenario with core-edge integration since the last IAEA in 2018 $[1]$. A discharge with a duration over 60s with $\beta_P$ ~2.0, $\beta_N$ ~1.6, $H_{98y2}$~1.3 and internal transport barrier on electron temperature channel is obtained with multi-RF power heating and current drive, i.e. ~2.5 MW LHW and 0.9 MW ECH, where the plasma configuration is the upper single null with the strike points on the tungsten divertor (shown in figure 1). Loop voltage was well controlled to be zero which indicates the fully non-inductive current drive condition. Small ELMs (frequency ~100-200Hz) were obtained in this long pulse H-mode discharge. In the operation, the optimization of X-point, the outer gap and local gas puffing near LHW antenna were investigated to maintain RF power coupling and to avoid formation of hot spot on the 4.6 GHz LHW antenna. Global parameters of toroidal field $B_T$ and line averaged electron density <$n_e$> were optimized for high current drive efficiency of LHW and for on-axis deposition of ECH. The on-axis ECH was applied not only for the core electron heating, but also for the control of high $Z_{eff}$ impurities in the core plasmas.
Meanwhile, a higher $\beta_N$ ~1.8 with a duration of 20s is achieved by using the modulated neutral beam. Several normalized parameters of $\beta_P$ ($\beta_P$ ~2.0), $\beta_N$ ($\beta_N$ ~1.8), $H_{98y2}$ (~1.3), $n_e/n_{GW}$ (~ 0.75) are close or even higher than the phase III 1GW scenario of CFETR steady-state $[2]$. Other features such as metal wall (tungsten divertor), low torque injection ($\Gamma_{inj}$~1.0Nm), electron dominated heating ($T_e$>$T_i$), moderate bootstrap current fraction ($f_{bs}$~50%), broaden current density profile with the central q(0)>1.0 and good energy confinement, have also been demonstrated in this scenario. Note that high-Z impurity accumulation in the plasma core was well controlled in a low level by using the on-axis ECH and reducing the fast ion losses through beam energy optimization.
More recently, EAST has demonstrated a compatible core and edge integration in high $\beta_P$ scenarios: high confinement $H_{98y2}$>1.2 with high $\beta_P$ ~2.5/$\beta_N$~2.0, $f_{bs}$~50% is sustained with reduced heat flux by active divertor heat flux at high density $n_e/n_{GW}$ ~0.7 and moderate $q_{95}$~6.7 (shown in figure 2). The energy confinement quality was almost maintained with $H_{98y2}$>1.2 during the radiation feedback control. By active impurity seeding through radiative divertor feedback control via radiated power, the peak heat flux is reduced by ~30% on the ITER-like tungsten divertor, here a mixture of 50% neon and 50% $D_2$ is applied. Note that EAST has developed a number of heat flux control techniques to reduce heat load in separate experiments.
In summary, recent EAST experiments demonstrated long pulse steady-state high plasma performance scenarios and heat flux feedback control. Detailed physics basis to investigate stability and particle transport will be presented for the understanding of fully integrated core-edge solutions on EAST. As a test bed for ITER and CFETR, the EAST upcoming experiments will further exploit additional heating power and demonstrate the core-edge integration of steady-state long pulse high performance scenarios with full metal walls.
This work was supported in part by National Natural Science Foundation of China under Grant No. 11975274,11975276, US Department of Energy under DE-SC0010685 and DE-FC02-04ER54698.
$[1]$ X. Gong et al 2019 Nucl. Fusion 59 086030
$[2]$ J. Huang et al 2020 Plasma Phys. Control. Fusion 62 014019Speaker: Xianzu Gong (Insititute of Plasma Physics, Chinese Academy Sciences)
-
-
-
Off-axis Neutral Beam Current Drive (NBCD) physics has been tested on DIII-D for Advanced Tokamak (AT) operation with increased off-axis injection power ($P_{OANB}\simeq7$ MW) by using the newly available, toroidally steerable co/counter off-axis neutral beam (CCOANB) injection capability. DIII-D experiments confirm that the new CCOANB drives current as predicted by the classical model NUBEAM for MHD quiescent plasmas (Fig. 1). Compared to on-axis injection, substantial broadening of the current and pressure profiles has been achieved with dominant OANB heating by injecting both the new CCOANB and the previous vertically steerable OANB. This is consistent with predictions of the theory-based IPS-FASTRAN integrated modeling that has guided the DIII-D beam system upgrade for the development of reactor relevant $\beta_N>4$ steady-state scenarios. Projecting to the Compact Advanced Tokamak (CAT) fusion pilot plant shows that off-axis NBCD aligns well with the high bootstrap current $f_{BS}>0.8$ operation, maintaining a broad current profile with $q_{min}>2$ and excellent off-axis NBCD efficiency.
The NBCD profile driven by the new CCOANB was measured in H-mode plasmas and compared with modeling. The NBCD measurement$^1$ is based on the local measurement of the magnetic field pitch angles from Motional Stark Effect (MSE) diagnostics. These pitch angles are converted to the flux-surface-average current density ($J_\parallel$) and parallel electric field ($E_\parallel$), either using kinetic EFIT equilibrium reconstruction or a more direct MSE analysis. This allows the beam driven current to be determined by $J_{NB} = J_\parallel – \sigma_{NEO}E_\parallel -J_{BS}$, where $\sigma_{NEO}$ and $J_{BS}$ are the neoclassical conductivity and bootstrap current density calculated by the Sauter model using the measured kinetic profiles as input. Differential NBCD measurement reduces model dependencies ($\sigma_{NEO}$ , $J_{BS}$) and systematic uncertainties of measurements. In this study, we compare two discharges, either i) on- and off-axis or ii) “Left” (more tangent) and “Right” (more perpendicular) off-axis NBCD in otherwise similar discharge conditions. The measured NBCD profiles driven by the new CCOANB in the co-current direction agree reasonably well with classical model of Monte Carlo beam ion slowing down calculation NUBEAM (Fig. 1). The NUBEAM modeling employs an accurate beam injection model validated against fast visible image and neutron measurement$^2$. The toroidal field direction (+BT) was chosen for better alignment of NBI to the local B, leading to good off-axis NBCD efficiency. The minimum value of q was maintained at $q_{min}>1$ without any significant core MHD activities. The measured $J_{NB}(\rho)$ in Fig. 1 shows a clear hollow NBCD profile with the peak at about half the minor radius $\rho\sim0.5$. The net driven current normalized to the total injection power is $I_{NB}/P_{NB} = 14.9$ kA/MW, which is as good as on-axis NBCD since the increased fraction of trapped electrons reduces the electron shielding in the outer radius region. The measured NBCD ($I_NB$) increases with the off-axis NB power. Higher NB power operation produced mild low-frequency MHD modes (n=2 and 3), resulting in the reduced NBCD compared with the classical model prediction. The measured NBCD profiles at the highest inject power matches the NUBEAM modeling with a modest anomalous beam ion diffusion $D_b = 0.3 ~{\rm m^2/s}$. However, even including the effect of finite $D_b$, the measured NBCD efficiency ($I_{NB}/P_{NB}$) does not decrease with $P_{NB}$.A low pressure peaking factor has been obtained with dominant off-axis NB heating. Figure 2 compares two discharges with different ratio of the OANB power to the total NB power ($f_{OANB} = P_{OANB}/P_{NB}$) at ~same injection power for elevated $q_{min} > {\sim1.5}$ discharges. The pressure peaking, $f_p = p_0/\langle p \rangle$ decreases substantially when the on-axis NB power (blue traces) is replaced by the new CCOANB (red traces), while maintaining $\beta_N\sim3$ with high total ($H_{89}\sim2.3$) and thermal ($H_{98}\sim1.25$) energy confinement. The lower pressure peaking also results in the increased low-n ideal $\beta_N$ stability limit$^3$. Both discharges inject full available power from the vertically steerable OANB at the maximum tilt angle. This profile broadening has been reproduced by the IPS-FASTRAN modeling$^4$ that integrates theory-based models of core transport (TGLF), edge pedestal (EPED1), equilibrium (EFIT), stability (DCON), heating and current drive (NUBEAM, TORAY) self-consistently to find steady-state ($d/dt = 0$) solutions. Figure 3 compares the pressure profile between the measurement (kinetic EFIT equilibrium reconstruction) and the IPS-FASTRAN modeling, where all transport channels ($n_e$, $T_e$, $T_i$, rotation, and current) are predicted without any significant free input parameters except the density and rotation values at the pedestal top.
The measured off-axis NBCD does not lose CD efficiency by going to a larger radius, which is beneficial for future AT reactors. The IPS-FASTRAN modeling predicts significant improvement of energy confinement time for broad current profile with flat or weak negative magnetic shear, compared with monotonic $q_0\sim1$, for the AT reactors. Projecting to the CAT $^5$ Fusion Pilot Plant with R = 4 m, B = 7 T shows an excellent off-axis NBCD efficiency for high Greenwald density fraction operation leading to high fusion performance. Off-axis NBCD aligns well with the high bootstrap current $f_{BS}>0.8$ operation maintaining broad current profile with $q_{min}$>2.
This material is based upon work supported by the U.S. Department of Energy, Office of Science, Office of Fusion Energy Sciences, using the DIII-D National Fusion Facility, a DOE Office of Science user facility, under Awards DE-AC05-00OR22725, DE-FC02-04ER54698, DE-AC02-09CH11466, DE-FG02-07ER54917, DE-SC0012656
$^1$ J.M. Park, et al., Phys. Plasmas 16, 092508 (2009).
$^2$ B.A. Grierson, et al., submitted to this conference.
$^3$ B.S. Victor, et al., submitted to this conference.
$^4$ J.M. Park, et al., Phys. Plasmas 25, 012506 (2018).
$^5$ R.J. Buttery, et.al., IAEA FEC, FIP-P3/26 (2018).Speaker: Jin Myung Park (Oak Ridge National Laboratory) -
The DIII-D tokamak has developed a new regime for high-beta hybrid plasmas where the broad current profile is achieved with strong off-axis electron cyclotron current drive (ECCD) rather than anomalous poloidal magnetic flux pumping. The high-beta hybrid regime with $q_{min}$ slightly above 1 and without sawteeth is a candidate for the $Q=5$ steady-state scenario on ITER$^{1-3}$, but the anomalous flux pumping mechanism that maintains $q_{min}>1$ despite strong central current drive is not yet understood$^4$. Experiments on DIII-D have found that high performance with $\beta_N=3.7$ and $H_{98y2}=1.6$ is maintained (Fig. 1) in high-density hybrids when 3.4 MW of ECCD is moved from $\rho<0.2$ to $\rho\sim 0.5$. The good agreement between the experimental $q_{min}$ evolution and TRANSP simulations (Fig. 2) differs from the usual hybrid situation where the simulation predicts $q_{min}<1$ but experimentally $q_{min}>1$, showing there is no evidence for anomalous flux pumping in this new hybrid regime with off-axis ECCD. Transport analysis finds higher density leads to weaker Alfven eigenmode (AE) activity (Fig. 3) and lowers the electron thermal diffusion, but there is little change to thermal transport in moving ECCD from on-axis to off-axis.
A reproducible high-beta hybrid regime has been developed on DIII-D$^{1-3}$ with stationary performance as high as $\beta_N=3.7$ and $H_{98y2}=1.6$ using 11.2 MW of NBI and 3.4 MW of co-ECCD (Fig. 1), with a total injected energy of up to 56 MJ. Stationary hybrid plasmas always exhibit a $m=3/n=2$ or higher order mode. While these discharges are fully non-inductive at moderate densities ($\sim4\times 10^{19}\mbox{ m}^{-3}$), at higher densities (up to $6.1\times 10^{19}\mbox{ m}^{-3}$) this long-duration regime is well suited for radiative divertor studies with the ECCD aimed at $\rho\sim 0.5$ to avoid the right hand density cutoff$^5$. Off-axis ECCD results in a broader current profile than on-axis ECCD, as measured by motional Stark effect (MSE) polarimetry. In these experiments the six gyrotrons use identical aiming, giving very localized ECCD profiles. Perhaps because of this, deleterious $m=2/n=1$ modes are destabilized when the ECCD location is between $\rho=0.25–0.40$, which corresponds to the region of the $q=2$ surface. Furthermore, for off-axis deposition, radial ECH injection is found to be more stable to $m=3/n=1$ mode onset than current drive injection.
While hybrids with strong off-axis ECCD have a broad current profile with $q_{min}>1$, this behavior is not anomalous and there is no evidence of poloidal magnetic flux pumping. It is well established that hybrid plasmas with on-axis current drive have anomalously broad current profiles with a measured $q_{min}$ well above simulations of the current profile evolution from transport codes like TRANSP$^{2-3}$. When the ECCD is moved off-axis to $\rho\sim 0.5$, the measured $q_{min}$ value increases (as expected for off-axis current drive) and TRANSP simulations of the expected $q_{min}$ evolution are in reasonable agreement with the minimum safety factor from MSE-constrained equilibrium reconstructions (Fig. 2). The measured and simulated loop voltage profiles are also in good agreement, with central values around 50 mV and edge values near zero; the peaked loop voltage profile shows that the current profile is continuing to broaden with time. (While the loop voltage profile rests at zero for hybrids with $n_e\sim 4\times 10^{19}\mbox{ m}^{-3}$ and on-axis ECCD, these hybrids with $n_e\sim 6\times 10^{19}\mbox{ m}^{-3}$ and lower off-axis ECCD efficiency retain positive loop voltage.) While steady-state hybrids with on-axis ECCD and $q_{min}\sim 1.1$ generate strong fishbones, the absence of the fishbone instability for hybrids with off-axis ECCD confirms the higher $q_{min}$ values ($\sim 1.5$).
An additional advantage of off-axis ECCD in hybrids is that higher density plasmas can be investigated without encountering the right hand density cutoff, which increases the confinement time and allows higher $\beta_N$ to be achieved due to lower electron thermal transport and reduced AE activity. Moving the ECCD deposition from on-axis to off-axis by itself has little measurable effect on local thermal transport, although the smaller electron temperature gradient inside the ECCD deposition zone does decrease the global confinement time. By raising the density, however, high confinement and high $\beta_N$ can be recovered for off-axis ECCD. About half of the confinement improvement at higher density is due to ~30% lower electron thermal transport. The remaining confinement improvement at higher density is due to reduced beam ion transport. In TRANSP simulations that adjusted an anomalous beam ion diffusion coefficient to match the experimental neutron rate, $D_{beam}$ decreased from $\sim 1.5\mbox{ m}^2/\mbox{s}$ to less than $1.0\mbox{ m}^2/\mbox{s}$ with higher density (Fig. 3). While high $D_{beam}$ does not directly affect thermal transport, it reduces the neutral beam heating effectiveness and thus lowers the heat flux and temperature gradients. The lower $D_{beam}$ at higher density is consistent with the weaker AE activity observed on the density interferometer. It is also found in these hybrid plasmas that particle transport, but not thermal transport, is strongly affected by the onset of a $m=3/n=1$ mode, such that higher and broader density profiles can be achieved only in the absence of the $m=3/n=1$ mode.
Since the mechanism of poloidal magnetic flux pumping in hybrid plasmas is still under investigation, there is interest in developing an ITER scenario with comparable performance that does not rely on anomalous flux pumping to maintain the target safety factor profile. The 'classical' behavior of the current profile with off-axis ECCD reported here demonstrates the existence of a new hybrid regime that does not rely on anomalous flux pumping to maintain $q_{min}>1$. It is interesting to note that this regime has similar characteristics to experiments in the "high $q_{min}$" regime with broad off-axis ECCD between $\rho\approx 0.3–0.6$ in which $q_{min}$ dropped to $\approx 1.4$, although the $m=3/n=2$ mode that seems important in hybrids was not present in these "high $q_{min}$" plasmas$^6$. This indicates that different paths (starting from low or high $q_{min}$) can be taken to access this regime.
This material is based upon work supported by the Department of Energy under Award Number(s) DE-FC02-04ER54698, DE-FG02-04ER54761, and DE-AC52-07NA27344.
$^1$F. Turco, et al., Phys. Plasmas 22 (2015) 056113.
$^2$C.C. Petty, et al., Nucl. Fusion 56 (2016) 016016.
$^3$C.C. Petty, et al., Nucl. Fusion 57 (2017) 116057.
$^4$C.C. Petty, et al., Phys. Rev. Lett. 102 (2009) 045005.
$^5$T.W. Petrie, et al., Nucl. Fusion 57 (2017) 086004.
$^6$T.C. Holcomb, et al., Nucl. Fusion 54 (2014) 093009.Speaker: C. Craig Petty (General Atomics) -
The radial width of heat flux flowing into the DIII-D divertor is found to expand beyond that of the established empirical scaling (1) for conditions of high input power and high plasma density. This expansion is consistent with a scrape-off-layer (SOL) radial pressure gradient limited by the MHD ballooning stability limit, but does not inherently result in a degradation of edge pedestal pressure or core confinement due to additional edge turbulence. This result has favorable implications for access to dissipative divertor regimes in future reactor-scale tokamaks.
At low heating power, ~3 MW, the DIII-D SOL heat flux width remains consistent with the empirical scaling law (1), dependent only on the midplane poloidal field. The low power midplane separatrix normalized pressure gradient, $\alpha _{MHD}$, increases with the higher density required for divertor detachment. At high heating power, ~ 13 MW, a higher separatrix density, $n_{e,sep}$, and resulting higher separatrix pressure, are required to achieve divertor detachment. For $n_{e,sep}$ approaching half of the Greenwald density limit, $n_{GW}$, the separatrix pressure gradient saturates, consistent with previous studies (2). Further increases in density or input power result in a broadening of the SOL and divertor temperature and density profiles, maintaining the pressure gradient near the MHD limit. The increase and saturation of the separatrix pressure gradient is summarized in Fig. 1, where the pressure gradient is normalized to the MHD ideal ballooning limit, $\alpha _{crit}$.
The saturation in $\alpha _{MHD}$ occurs at the same separatrix density even for attached divertor conditions at high power indicating the saturation is not due to divertor detachment.
The separate components of the midplane pressure profile are measured with Thomson scattering for the electron pressure and Charge-Exchange Recombination spectroscopy (CER) of the CVI impurity emission for the ion temperature and density contributions to the pressure profile. The separatrix normalized pressure stability limit, $\alpha _{crit}$, is evaluated with the ideal MHD code BALOO based upon magnetic equilibria across the data set at $\alpha _{MHD}\approx2.2-2.7$. As shown in Fig. 1, the measured pressure gradient, $\alpha _{MHD}$, saturates at about 50% above the MHD limit. The high pressure gradient is likely due to the high $T_{i}>>T_{e}$, at the separatrix taken from CVI CER measurements. Recent measurements from main ion CER indicate a separatrix $T_{i}$ much closer to $T_{e}$, resulting in a pressure gradient closer to the stability limit.
The saturation of the SOL pressure gradient results in an expansion of the SOL width as power and density are increased, as shown in Fig. 2. The SOL $T_{e}$ width, remains constant at ~1.8 times the ITPA $\lambda_{q}$ scaling for $n_{sep}∕n_{GwW}\leq0.3$, but then increases for higher density. The SOL $\lambda _{Te} $ is ~40% below that implied by the ITPA scaling given that $\lambda_{Te}\sim\frac{7}{2} \lambda_{q}$ at these collisionalities. For detached plasmas, shown by solid symbols in Fig. 2, as the input power is increased from 2 MW to 13 MW while the density required for detachment increases, the SOL $\lambda_{Te}$ and implied $λ_{q}$, increase $\geq$ 50%. Expansion of the SOL density width, $\lambda_{ne}$, is even stronger with a factor of 2.5 increase in width for the high-power case with divertor detachment. Implications for the SOL expansion at high power and density can be seen in the divertor plasma as well.
Shown in Fig. 3 are radial profiles at low and high power through the divertor leg halfway between the target and the X-point. These are both detached plasmas with near complete exhaust power dissipation. At high power the divertor plasma is ~3 times broader than the lower power case with both at $T_{e} \sim5-10\,eV$ at the same vertical location. The broader profile allows for greater total radiated power without a significant increase in the divertor density.
The increased turbulence and radial transport at high power and density might be expected to degrade the edge pedestal and resulting core confinement. However, no degradation of the pedestal is found with increased SOL width at high power and detachment onset. For low and high-power the density is increased with deuterium injection to achieve divertor detachment with the intrinsic carbon impurity radiation. While the SOL heat flux width increases by 50% for the high-power case compared to the low power case which remains at the empirical width scaling (Fig. 2a), the pedestal pressure is maintained at that expected from the EPED model. The normalized confinement at high power divertor detachment also remains similar to that for detachment at low power.
These results are encouraging for the compatibility of divertor heat flux control with core operational scenarios in future high-power density tokamaks. The expansion of the SOL width due to MHD stability can reduce parallel heat flux density allowing for divertor detachment at lower plasma and seeded impurity density than implied by simple scaling arguments (3). However, these results also imply that for study of high-power density, and resulting high divertor plasma density, in existing or future divertor test tokamaks, will require similar magnetic field values to those planned for reactor-scale tokamaks.
(1) T. Eich, et al., Nucl. Fusion 53 (2013) 093031.
(2) T. Eich, et al., Nucl. Fusion 58 (2018) 034001.
(3) M.L. Reinke, et al., Nucl. Fusion 57 (2017) 034004.
This material is based upon work supported by the Department of Energy under Award Number(s) DE-FC02-04ER54698.Speaker: Anthony W. Leonard (General Atomics) -
Detachment control tests at DIII-D and EAST have expanded to new sensors and integration with high confinement ($H_{98,Y2}$≈1.5, $\beta_N$=3) core scenarios (see $^1$ for details on core performance). Active detachment control protects the divertor target from extreme heat fluxes and temperatures which might otherwise cause melting and erosion while minimizing fuel or impurity seeding commands to what is required and thus mitigating core performance degradation. Using Langmuir probes (LPs), saturation current normalized to its value at rollover can be controlled in the range $J_{sat}/J_{roll}$ > 0.3, including tracking of target values that step or ramp. Additionally, triple probe tips at EAST measure Te, which is controllable for $T_e$ ≳ 3 eV, with the exception of dithering across the $T_e$ cliff.$^2$ Identification of the system by fitting the response of $J_{sat}$ or Te to gas puffing leads to two things: estimates for initial proportional-integral-derivative (PID) gains for the controller, and a discrepancy between the classic system models tested and the actual behavior. The inability of classic system models to adequately describe the response to gas puffing suggests that future work in detachment control should focus on a control-oriented reduced model and Model Predictive Control (MPC).
LPs have been demonstrated to be effective sensors for detachment control systems at DIII‑D and EAST, with a variety of puffing gas species as the control actuators. These results go beyond the original demonstration of LP control at JET$^3$ by using Te from triple LPs at EAST, testing the controller with D$_2$, CD$_4$, N$_2$, Ne, and Ar puffing, and integrating detachment control with high confinement core scenarios on both devices.$^1$ The $J_{sat}$ rollover detection logic at DIII-D also goes beyond previously reported results by allowing detection of roll-back-over or reattachment; the controller should always know which side of rollover it is on. This is important for coping with off-normal events because the system gain changes sign as $J_{sat}$ crosses the rollover point. An advantage of using LPs over Divertor Thomson Scattering (DTS) as the sensor as in previous DIII-D detachment control demonstrations$^2$ is that LPs typically have much greater spatial coverage. The $J_{sat}$ controller has demonstrated ability to track targets in the range $J_{sat}/J_{roll} \geq$ 0.3, including ramps, steps, and constant values (an example is shown in Figure 1). The EAST $T_e$ controller has been successfully operated at a variety of $T_e$ values (Figure 2) and has been used as a tool in detachment studies.
Detachment control is useful for supporting core-edge integration since it keeps gas puffing down to the minimum needed to reach the target. Detachment control experiments have achieved $H_{98,y2}$ > 1 at EAST and $\approx$1.5 at DIII-D. It is typically not challenging to avoid overshooting the target with this controller with appropriate tuning of the full PID system (PI-only can be effective as well) and a realistic target trajectory. That is, best results are found when abrupt steps in the target are replaced by transitions over a time period similar to the system dead time plus response timescale, or about 100-200 ms, as in Figure 1.
System identification of the $J_{sat}$ controller (which has been tested more aggressively) with a First Order Plus Dead Time (FOPDT) or second order (SOPDT) model is sensitive to the absolute value of $J_{sat}/J_{roll}$, meaning the optimal tuning for going from 0.8 to 0.5 is not the same as the optimal tuning for going from 0.5 to 0.3. Figure 1 shows a case where the target $J_{sat}/J_{roll}$ decreases in two steps and then is held at a constant value after each. When the whole history is fit to the SOPDT model (which extends FOPDT by adding damping), the response is underfit with reduced $\chi^2$=12. When the fit is restricted to a time window surrounding a single step, the model provides a much better description of the response with reduced $\chi^2$ of 1 or 2. The separate fits have different parameter values. The problem can be linearized locally over a small enough step, but a small enough step appears to be $\Delta J_{sat}/J_{roll} \approx$ 0.2, whereas the controller should be able to handle attachment down to deep detachment, or $J_{sat}/J_{roll} \approx$ 1. Additionally, classic formulae for calculating PID gains from FOPDT identification, such as the Ziegler-Nichols rule, consistently need fine tuning by hand for best performance. This is consistent with FOPDT and SOPDT being inadequate descriptions of the system. So, PID is not the optimal method for controlling this system if the PID gains need to be functions of the target value of $J_{sat}/J_{roll}$. Future work should implement MPC to produce a more effective and more general tool, but the difficulty is in obtaining a suitable reduced model for the system.
ITER will need an active detachment control system to protect its divertor plates. Detachment control tests on present-day devices aim to demonstrate ITER-relevant control schemes or identify critical research needs. Work so far on DIII-D and EAST shows that LPs, DTS, and radiated power measurements are adequate sensors for managing fuel or impurity puff commands in a narrow operating space, and that the algorithms must be improved in order to handle more general conditions. This work was supported in part by the US Department of Energy under DE-FC02-04ER54698.
$^1$Liang Wang, presentation at IAEA FEC 2020
$^2$D. Eldon, Nucl. Fusion 57, 066038 (2017); doi: 10.1088/1741-4326/aa6b16
$^3$C. Guillemaut, Plasma Phys. Control fusion 59, 045001 (2017)Speaker: Dr David Eldon (General Atomics) -
Recent tungsten (W) divertor experiments in the DIII-D tokamak have made significant progress elucidating key mechanisms responsible for high-Z erosion, re-deposition, leakage, and scrape-off-layer (SOL) transport. These results have important implications for ITER and other next-step fusion devices, including insight into W sourcing during mitigated edge localized modes (ELMs), diagnosing and understanding the net erosion of W in steady-state and transient phases, and developing a predictive capability for both local and global high-Z material migration.
The free-streaming plus recycling model (FSRM), a recently developed analytic model for intra-ELM, high-Z gross erosion$^1$, was validated in ITER-relevant, ELM-mitigated regimes achieved via pellet pacing and resonant magnetic perturbations (RMPs) in DIII-D (Fig. 1). ELM control via pellet pacing reduces W sputtering by suppressing impurity build-up in the pedestal (by increasing the ELM frequency), resulting in less physical sputtering of W caused by free-streaming, fully-stripped impurities expelled from the pedestal. Pellet pacing also decreases the divertor target electron temperature (due to increased edge neutral fueling), leading to less tungsten sputtering by low C charge states recycled from the divertor.
During the application of RMPs for ELM control, however, the peak W gross erosion rate actually increases. FSRM calculations indicate this is due to a decrease in the intra-ELM magnetic connection length, $L_\parallel$, relative to the length of the ELM filament, $L_{ELM}$. The ratio of these two quantities is the only free parameter in the FSRM. RMPs may cause the formation of a direct path between the strike-points and a reservoir of hot impurities deep into the plasma core$^1$, increasing the effective physical sputtering yield of tungsten from ELM impacts.
Leveraging newly calculated atomic rate coefficients and a recently developed ultraviolet spectroscopy system$^2$, we also present the first detailed, in-situ analysis of the net erosion of high-Z material in a fusion device. Tungsten gross and net erosion are measured spectroscopically via the ionizations per photon, or S/XB, method$^1$. W net erosion is nearly equal to gross erosion when the neutral ionization length, $\lambda_{iz}$, is large relative to the W$^{^+}$gyro-radius, $\rho_W$, and magnetic pre-sheath width, $\lambda_{sh}$, due to a lack of W prompt re-deposition. As $\lambda_{iz}$ decreases, the rate of W net erosion also decreases, with a strong inflection point near $\lambda_{iz}/\rho_W \approx \lambda_{iz}/\lambda_{sh}$ ~ 2-3. At the lowest ionization lengths achieved, $\lambda_{iz}/\rho_W \approx \lambda_{iz} /\lambda_{sh}$ ~ 0.4, net erosion drops to as little as ~20% of the gross erosion. Measurements are lower than calculations from analytic prompt re-deposition models$^{3,4}$ when $\lambda_{iz}$ is large. This is attributed to non-local W re-deposition, which is not included in the models. When $\lambda_{iz}$ is small relative to $\rho_W$ and $\lambda_{sh}$, more net erosion occurs than calculated by the models. ERO simulations indicate that W re-deposition from charge states W$^{^{2+}}$and higher (not included in the simple analytic models) becomes important in this regime.
Mixed-material, high-Z impurity migration models have also been validated by DIII-D tungsten divertor experiments. Simulations conducted using the coupled DIVIMP-WallDYN codes indicate $E$×$B$ drifts are the dominant driver of material migration in the divertor region in L-mode discharges with the ion $B$×$\nabla B$ drift away from the strike-points. Incorporating $E$×$B$ drifts, calculated via integrating Ohm's law along parallel field lines, is essential to match the W migration features observed experimentally$^5$. A peak in the W deposition profile appears many ionization lengths radially outboard of the ring of W-coated divertor tiles (Fig. 2). The WallDYN model agrees with experimental data within a factor of two over the entire radial extent of the W re-deposition pattern when $E$×$B$ drifts are adjusted to 60% of their calculated value (Fig. 2). This suggests (a) high-Z material migration involves multiple erosion/re-deposition events before reaching equilibrium, and (b) effects beyond Ohm's law, such as SOL currents, may play an important role setting the electric field in the divertor region.
Finally, the net force on high-Z impurities in the far scrape-off-layer is observed to change dramatically with ion $B$×$\nabla B$ drift direction in DIII-D, suggesting strong high-Z entrainment in SOL flows. Tungsten flow patterns in the low-field side, far-SOL (deduced via a midplane collector probe system$^7$) are observed to be primarily towards the outer target with the ion $B$×$\nabla B$ drift away from the outer strike point (OSP), and vice versa for $B$×$\nabla B$ drifts towards the OSP. An ad-hoc SOL flow model added to DIVIMP simulations (based on experimental flow profiles) produced reasonable agreement with measured high-Z flow patterns. This suggests that background plasma flow patterns may be instrumental in causing the long-hypothesized near-SOL, high-Z impurity accumulation, which sources impurities into the far-SOL via cross-field diffusion.
Because the atomic physics and surface physics of carbon and beryllium are quite similar, experimentally validated PMI models on DIII-D significantly advance the field towards a predictive capability for the high-Z divertor erosion, re-deposition, and leakage in the mixed-material environment of ITER. Such models are essential to develop/optimize mitigation strategies for minimizing high-Z core contamination and maximizing fusion gain.
This work is supported by the U.S. Department of Energy under DE-FC02-04ER54698.
- T. Abrams et al., Phys. Plasmas 26 (2019) 062504
- C.A. Johnson et al., Plasma Phys. Control. Fusion 61 (2019) 095006
- D. Naujoks et al., Nucl. Fusion 36 (1996) 671
- J. Guterl et al., Contrib. Plasma Phys. (2019) 1-7
- W.R. Wampler et al., Phys. Scr. T170 (2017) 014041
- E.A. Unterberg et al., Nucl. Fusion 60 (2020) 016028
- D.C. Donovan et al., Rev. Sci. Instrum. 89 (2018) 10I115
Speaker: Tyler Abrams (General Atomics) -
We report the achievement of a world unique capability of high power co/counter steerable off-axis neutral beam injection on a major tokamak, which widens the broad pressure and current profile parameter space for high beta steady-state advanced tokamak (AT) scenarios on DIII-D, while retaining the ability to balance the injected torque for low rotation studies. The unique steering capability of co/counter off-axis neutral beam (CCOANB) is being used to validate physics-based energetic particle and thermal transport models that are utilized in designing next-step facilities based on the steady-state AT approach. Prior to meaningful validation, however, a careful assessment of the transmitted power and energetic ion population produced by this novel heating and current drive system is critical. In this work, the off-axis beam injection is assessed through visible imaging (Fig. 1), neutron measurements and rotation profile measurements at balanced torque (Fig. 2), and used to broaden the pressure profile in steady-state advanced tokamak scenarios.
DIII-D has undergone a major upgrade and successfully injected high power off-axis neutral beam power (~4 MW) using the CCOANB in both co-current and counter-current directions. The total off-axis neutral beam power is approximately 7.3 MW at nominal operating voltage of 75 kV. Achieving an off-axis and co/counter steering capability necessitated significant modifications to the ion sources, internal neutral beam components and adaptor seal at the tokamak vessel connection. DIII-D is equipped with four neutral beam lines injecting through midplane ports at toroidal angles of 30, 150, 210 and 330 degrees. Each beamline houses two ion sources individually capable of ~ 2 MW injected power at ~ 80 kV, labeled left (LT) and right (RT) as viewed from behind facing the torus. For the CCOANB upgrade to the 210 degree beamline, each ion source has been modified to achieve a stronger vertical focus by modifying the accelerator grid modules to aim towards the beam centerline in a manner similar to Ref. 1, but the width of the ion source plates have not been reduced to retain high power. Tilting of the ion source plates is required for the beam to pass through the smaller effective aperture when the beam is off-axis. Minimal losses of neutral beam power have been achieved by optimizing the strong focus ion sources and optimization of the gradient grid voltage, enabling maximum power transmission by minimizing the amount of power “scraping off” on internal beamline components. Tilting of the ion source has been guided by fast visible imaging in Ref. 2, as shown in Fig. 1 (a,b), and resulted in neutral beam injection along the design centerline, as shown in Fig. 1 (c,d), with empirical characterization of each beam’s divergence derived from the beam vertical profiles, as shown in Fig. 1 (e,f). These measured beam characteristics are used in the parameterization of the beam in the NUBEAM Monte-Carlo heating and current drive package.
Through exclusive power injection of each source into MHD quiescent plasmas across a range of neutral beam voltage, perveance and plasma current in the same manner as Ref. 3, we conclude that a modest reduction of transmitted power (compared to on-axis, standard focus) has occurred. Prior to the beamline modification the 210LT and 210RT sources operating at 75, 81 kV and 2.60, 2.56 $μ$-perv respectively, each produced 2.0 MW. After commissioning, optimizing the neutral beam aiming, and implementing the off-axis injection geometry in NUBEAM, an approximate 15% lower than expected neutron yield has been observed. We attribute this reduced neutron yield to interaction with internal beamline components and reionization in the drift duct, as indicated by thermocouple measurements and photoemission detected by photodiodes in the drift duct.
Good ability to balance the neutral beam torque has been demonstrated by injecting the new (210 degree) off-axis counter injected beam against the existing (150 degree) off-axis co-injected beam in 2.0 T, 1.0 MA, MHD quiescent L-mode plasmas. L-mode at 1.0 MA has been chosen to minimize the edge intrinsic torque, retain good beam ion confinement with off-axis injection and minimize MHD. Verifying the ability to operate with balanced injection is critical for achieving “low torque” and “low rotation” operation for physics studies and in ITER demonstration discharges. This torque balancing exercise has been performed in a matrix using both tangential and perpendicular injection (2x2) and a three point scan in counter-injected voltage, as shown in Fig. 2. In these conditions, the intrinsic rotation profile has a positive edge offset with a slightly hollow core rotation, and therefore we do not expect to achieve a flat zero velocity profile at 0.0 Nm. Nevertheless, at zero injected torque the observed velocity shear is very weak and much smaller than a single co-injected off-axis neutral beam injecting +0.62 Nm of torque, confirming that low toroidal rotation shear can be achieved by balancing torque and controlling the beneficial effects of $E \times B$ shear on confinement. This submission supports DIII-D papers by C.S. Collins, J.M. Park and B.S. Victor.
This material is based upon work supported by the U.S. Department of Energy, Office of Science, Office of Fusion Energy Sciences, using the DIII-D National Fusion Facility, a DOE Office of Science user facility, under Awards DE-FC02-04ER54698, DE-AC02-09CH11466, DE-AC05-00OR22725, DE-FG02-07ER54917, DE-SC0020337
(1) C.J. Murphy, et. al., lEEE/NPSS 24th Symposium on Fus. Eng. SP3-21 (2011)
(2) M.A. Van Zeeland, et. al., Plasma Phys. Control. Fusion 52, 045006 (2010)
(3) W.W. Heidbrink, et. al., Nucl. Fusion 52, 094005 (2012)Speaker: B.A. Grierson (Princeton Plasma Physics Laboratory) -
Introduction: Disruptions due to tearing mode locking are one of few potential obstacles remaining for successful ITER fusion reactor and beyond. Here, we report the experiments of locking avoidance, but also contributing to the H-mode recovery and sustainment by slowly rotating edge-localized tearing mode (TM) layers by applying 3D external field. The process is expected as a non-linear response to applied external 3D field, static error field and/or their combinations. Preliminary NIMROD simulation are qualitatively consistent with experiment.
Localized edge tearing layer response: It has been discovered that the H-mode recovery with a slowly rotating external 3D field is accompanied by an edge localized tearing layer synchronized with the 3D field. The amplitude of the tearing layer is a few hundred Gauss, suggesting that the perturbed current density must be comparable to the equilibrium current density. In contrast, at q~2 and 3 the response to 3D external field is minimal. One possible hypothesis is that the 3D field facilitates the H-mode recovery and the stable edge tearing layers assist the sustainment of H mode by reducing the influence of error field (EF) and other 3D fields around core including the applied external 3D field. This is consistent with the concept of “screening out” EF and other 3D fields, proposed by refs.$^{1,2}$. These results are with similar dependence as previously reported$^{3}$. We also discuss preliminary results of NIMROD simulation based on experimental conditions.
Figure 1 shows the H-mode sustainment by the NTM locking avoidance with rotating external 3D field and also the successful H-mode recovery after the preprogrammed reduction of NBI torque input terminated the H-mode phase ( the total NBI power input level was kept same). In the recovery phase, the ratio of the average density (Fig.1(f)) to the edge density (Fig.1(g)) is near unity, re-establishing the good edge confinement property. The beta_n recovered to the level comparable in the initial H-mode sustainment period (Fig.1(c)). However, The further gradual increase of the 3D field by upper/lower I-coils current ~3.2 kA (Fig.1(d)) reduced the density gradient at the edge together the beta_n decrease (Fig.1(e,g)), indicating that the fine optimization of the external 3D fields pattern is the key for sustaining high plasma confinement performance.
Reconstructed plasma response: Figure 2 shows the time-evolution of the localized TM layer response amplitude by the perturbed vertical field component dBz measured with Motional Stark Effect (MSE) in the H-mode recovery period (3300-3520ms). Two cycles of oscillation were decomposed by taking the covariance with external magnetic sensor signals (n=1, 2, 3). Although the radial coverage by MSE measurements are limited, the localized tearing layer expanded from edge to q=3 domain with I-coil current increase(Fig.2(b)). The magnetic response around q=2 (Fig. 2(c,d)) is minimal with lower 3D field just after H-mode recovery up to beta_n maximum time period. With increasing 3D field, the response below q=2 became finite and is synchronized from the core to the edge (Fig.2(e)). Thus, the higher 3D fields impacts on the structure from edge to core rather than functioning as the screening-out. (Similar dependence was observed on the ion rotation and ion temperature perturbation, not shown).
NIMROD simulation: Figure 3 shows two results of slow and fast (Ω = 3×10$^{4}$, 9×10$^{4}$ rad/s) rotation of the external 3D field by plotting the n=1 perturbed dBz field along the outboard mid plane. The calculation was performed using experimental equilibrium profile parameters (at 3445 ms), including the near-zero plasma rotation due to the 3D field viscous damping. The external 3D field pattern includes multi-poloidal components, but was simplified to a mix of 2/1, 3/1, 4/1 and 3/2. The EF was not included in this calculation. The plasma resistive timescales are roughly 100 times faster than in the experiment, with S = 3.5×10$^{6}$ the Lundquist number (ratio of the resistive to Alfven time). The radial derivative of dBz is the indicator of the appearance of a perturbed toroidal current at the rational surface. For slow 3D field rotation, the radial derivative of dBz increases around q=2 (rho~0.67). On the other hand, for fast 3D field rotation, the variation of radial derivative of dBz is minimal. This result is similar with the expectation from the screening-out m=2 component of 3D field coupled to the plasma response near the edge, but without error field, which can be explained by taking into account of the relative velocity of the rotating external 3D field and the static error field EF on the mode-moving frame (1,4). The relative frequency can be large enough to induce the screening effect simultaneously for both fields. The detail analysis has just begun.
Summary: The slowly-rotating 3D external magnetic field, by inducing the edge localized TM layers, can not only avoid tearing mode locking, but also screen both the static error field EF and external rotating 3D field. The present operational frequency range is several tens-Hz. If the frequency limit is related to the resistive-wall frequency (this seems to be the case in the DIII-D experiment), in the ITER and fusion reactors the effective resistive-wall frequency with blanket is expected lower than one-Hz. Thus, this slowly-rotating 3D field application seems to become very practical. Furthermore, this slowly rotating 3D field approach can serve as a tool to control actively the plasma flow velocity at various rational surfaces utilizing multi-3D field patterns. This could assist the plasma performance optimization from the plasma core to the edge.
(1) Inoue, S., et al., Nucl. Fusion 57 116020 (2017). (2) Inoue, S., et al., Plasma. Phy. Control. Fusion 60 025003 (2018). (3) Okabayashi, M., et al., Nucl. Fusion 59 (2019) 126015. (4) Finn, J., et al., Phys. Plasmas 22 (2015) 120701.
This material is based upon work supported by the U.S. Department of Energy, Office of Science, Office of Fusion Energy Sciences, using the DIII-D National Fusion Facility, a DOE Office of Science user facility, under Awards DE-AC02-09CH11466, DE-FC02-04ER54698, DE-AC05-06OR23100, DE-FG02-04ER54761, DE-AC52-07NA27344 and DE-SC-0017992.Speaker: michio Okabayashi (Princeton Plasma physics Laboratory) -
Initial experimental evidence shows that the L-H transition power threshold P_th can be reduced via Neoclassical Toroidal Viscosity (NTV) associated with applied n=3 non-resonant magnetic fields (NRMF) [Figs. 1 and 2], and, independently, via a fast reduction or reversal of intrinsic edge toroidal (co)rotation [Fig. 3]. It is also demonstrated that a small/moderate increase in lower triangularity reduces P_th [Fig. 4]. These actuators are shown to enhance local E×B flow and E×B shear just inside the LCFS. Rotation reversals reduce the L-H transition power threshold with unfavorable grad-B drift direction to values similar to P_th in favorable configuration.
Techniques for reducing P_th, or mitigating its increase with applied magnetic perturbations are crucially important for ITER, particularly in view of marginal auxiliary power in hydrogen plasmas during the PFPO-1 campaign. Several approaches are demonstrated here that increase the L-mode E×B shear preceding the L-H transition, and potentially open up a path to reduce or mitigate P_th in burning plasma experiments. Control of L-mode edge shear flow is accomplished using either externally applied NRMF torque, or by changing the edge rotation boundary condition. In addition, P_th has been reduced via small plasma triangularity changes (within ITER-relevant margins).
Initial evidence shows that a large NRMF torque applied via the C-coil (1) can reduce P_th at low ion edge collisionality [ν_i*(ρ=0.95)~0.15] [Fig.1]. It is obvious from Fig. 1 that the NBI power required to access H-mode is reduced with applied NRMF. The power threshold (expressed approximately as P_th ~ P_NBI+P_OH-dW/dt where the Ohmic power P_OH and the time derivative of the stored energy W have been taken into account) is reduced by ~25% (from ~4.3 MW without NRMF to ~3.15 MW with applied NRMF/NTV). The core radiated power is not separately measured but is estimated to be smaller than 80 kW. This observation of a reduction in power threshold with NRMF differs fundamentally from the power threshold increase observed with applied Resonant Magnetic Perturbations (RMP) due to edge stochasticity and increased edge toroidal co-rotation (2). In contrast, CER measurements here show a substantial reduction of the L-mode edge toroidal (Carbon) co-rotation with applied NRMF/NTV just before the L-H transition, qualitatively consistent with the direction of the expected NTV counter-torque [Fig.2]. As a consequence, E×B flow shear near the LCFS is found to increase via increased radial shear in the toroidal rotation term in the radial ion momentum balance. The (partial) ITER 3-D coil set available during the PFPO-1 campaign can be used to generate large NTV in the edge plasma layer, favored by the low collisionality expected in the ITER L-mode edge.
We have also investigated in detail the effect of magnetic geometry on P_th, and attribute the increased L-H power threshold with unfavorable grad-B drift direction clearly to reduced E×B flow shear due to higher edge toroidal co-rotation [Fig. 3(a)]. The L-H power threshold is shown to decrease monotonically with increasing E×B shear in the outer L-mode shear layer near the LCFS (shown via NBI torque scan, Fig. 3c), and does not depend on shear in the inner shear layer.
However, it has been observed that spontaneous local rotation transitions increase E×B edge shear flow preceding the L-H transition in balanced/low torque ITER-similar shape (ISS) plasmas in DIII-D. Fig. 3(b) shows radial profiles of toroidal edge rotation approaching the L-H transition. The rotation reduction/ reversal commences locally near ρ=0.95; the layer of reduced rotation then expands radially over several ms, providing sufficient E×B flow shear to induce the L-H transition. As a result, P_th in ISS plasmas can be reduced to values similar to those observed with favorable grad-B drift direction [Fig. 3(a)]. The threshold reduction occurs over a range in NBI torque but is most effective for balanced torque or with low counter-Ip torque. Reversal/ reduction of intrinsic edge co-rotation appears to be triggered by transient increases in edge power flow (via sawteeth crashes or radial transport avalanches).
In ISS plasmas, increasing the lower triangularity by ~12% results in a clear reduction in L-H power threshold (by ~15%, shown here for hydrogen ISS plasmas (Fig. 4). A smaller data set in deuterium plasmas also demonstrates a ~10-12% reduction of P_LH at increased triangularity at low torque (≤ ±0.2 Nm). In the higher triangularity shape the outer divertor strike point was moved slightly further away from the lower divertor cryopump, which could also have affected divertor neutral pressure.This work was supported by the US Department of Energy under DE-FG02-08ER549841,
DE-AC05-00OR227252,DE-FG02-08ER 549993, DE-FC02-04ER546985,DE-FG02-07ER549176,
and DE-AC02-09CH114667.
(1) K.H. Burrell et al., Nucl. Fusion 53 (2013) 073038.
(2) L. Schmitz et al., Nucl. Fusion 58 (2019) 126010.
Speaker: L. Schmitz (University of California Los Angeles) -
A novel control architecture for simultaneous regulation of several plasma scalar variables, such as the thermal stored energy ($W$), the bulk toroidal rotation ($\Omega_\phi$), and the safety factor at various spatial locations (e.g., $q_{95}$, $q_0$), and for active suppression of Neoclassical Tearing Modes (NTMs) by means of ECCD {1} is shown to improve plasma performance even in the presence of NTMs on DIII-D. The Off-Normal Fault-Response (ONFR) system is employed as a supervisor to monitor the NTM development and assign the gyrotron authority to the competing control tasks (profile control vs NTM suppression). Instead of following the common previous approach based on attempting to regulate whole profiles {2}, a different control solution is utilized that consists in regulating such profiles only at specific spatial locations and/or by related volume-average magnitudes. Such point-wise/volume-average approach for regulating profiles in a tokamak may be more realistic in some cases due to the under-actuated nature of these devices. The integrated-control scheme has been tested in one-dimensional simulations using the Control Oriented Transport SIMulator (COTSIM), as well as in experiments in the DIII-D tokamak. The results obtained suggest that integrated control techniques may help to improve the plasma performance even in the presence of NTMs (see Fig. 1). In addition, good qualitative agreement is found between simulation and experimental results, showing that COTSIM can be a very powerful tool to advance the design of integrated control architectures {3}.
Successful regulation of $W$ has been demonstrated in DIII-D scenarios targeting $\beta_N \approx 3$ and $q_{min} \approx 1.4$ in both simulations (Fig. 1.A) and experiments (Fig. 1.B), despite the existence of a 2/1 NTM within the plasma. The IPB98(y,2)-scaling confinement factor, $H_{98(y,2)}$, is substantially improved when feedback control is employed (Fig. 1.C), possibly due to a close tracking of the $W$ target during the ramp-up and early flat-top phases of the discharge (which delays the appearance of the NTM) and during the flat-top phase (by increasing the injected power once the NTM develops). In addition, tight regulation of $W$ simultaneously with $q_{95}$ (see Fig. 2), which is equivalent to regulating $\beta_N$, and the use of preemptive NTM stabilization (see Fig. 3), also help to delay the NTM appearance and its effect on the plasma confinement, and allow for systematically achieving relatively high $\beta_N$ values ($\approx 3$) which could not be sustained without feedback. Simultaneously with $W$ regulation, control of $q_{95}$ and $q_0$ (see Fig. 2.A), as well as $\Omega_\phi$ (not shown in this synopsis), has been achieved in one-dimensional simulations for the same steady-state high-$q_{min}$ scenario using COTSIM. During experiments in DIII-D, $q_{95}$ was successfully controlled (see Fig. 2.B) under feedback by varying $I_p$ (Fig. 2.C), but the unavailability of counter-$I_p$ neutral beam injectors did not allow for $\Omega_\phi$ regulation. Moreover, the controller's performance to regulate $q_0$ was significantly worsened in experiments (see Fig. 2.B) as a result of having substantially limited off-axis power (the co-$I_p$, off-axis 210 beamline could only be employed for 2 s). When the 210 beamline was turned on at 3.25 s (see Fig. 2.C), $q_0$ gets close to its target for a short period of time. Later development of the 2/1 NTM caused a slight decrease in $W$ (with the associated reduction in the electron temperature and increase in the plasma resistivity), which relaxed the $q$ profile and made it difficult for the controller to raise $q_0$ in order to reach the desired target.
Active NTM suppression techniques by means of ECCD deposition at the applicable rational surface were also tested at the same time that regulation of the aforementioned individual scalars was carried out. Simulation results obtained with COTSIM estimated that the 2/1 NTM could be totally suppressed with about 3 MW of EC power in these steady-state high-$q_{min}$ scenarios (see Fig. 3) for initial seed islands of width 10 cm. During experiments, lack of EC power (1.5 MW maximum using the gyrotrons “Luke”, “Leia” and “Tinman”) and defective poloidal-mirror steering did not allow for total NTM suppression, although a delay in its development and a lower MHD amplitude were achieved.
These initial results suggest that feedback-control techniques that integrate individual-scalar-variable regulation algorithms and NTM suppression may substantially improve the plasma performance in steady-state high-$q_{min}$ scenarios in DIII-D. However, actuator availability is critical to achieve successful regulation of the different scalars. Further efforts will focus on developing control algorithms for scalars (possibly adding more scalars such as $l_i$, $\beta_p$ or $q_{min}$) and MHD stability that have greater levels of integration (e.g., with proximity controllers), as well as actuator management schemes. The final goals are to increase the effectiveness of the NTM suppression mechanisms, improve the controller’s performance, and achieve higher $\beta_N$ values.
Work supported in part by the U.S. Department of Energy under DE-SC0010661 and DE-FC02-04E54698.
{1} LA HAYE, R. J., GUNTER, S., HUMPHREYS, D. A., et al., Physics of Plasmas 9 (2002) 2051.
{2} WEHNER, W., BARTON, J., LAURET, M., et al., Fusion Engineering and Design 123 (2017) 513 , Proceedings of the 29th Symposium on Fusion Technology (SOFT-29) Prague, Czech Republic, September 5-9, 2016.
{3} HUMPHREYS, D., AMBROSINO, G., DE VRIES, P., et al., Physics of Plasmas 22 (2015) 021806.Speaker: Dr Andres Pajares (Lehigh University) -
Impurity seeding studies were performed for the first time in the slot divertor at DIII-D, showing that with suitable use of radiators, full detachment is possible without degradation of core confinement. First ever multi species SOLPS-ITER simulations including full cross-field drifts and neutral-neutral collisions in DIII-D demonstrate the importance of target shaping and plasma drifts on reducing impurity leakage. In addition to advancing our understanding of SOL impurity transport, these results show that neutral and impurity distributions in the divertor can be controlled through variations in strike point locations in a fixed baffle structure. This leads to enhanced divertor dissipation and improved core-edge compatibility, which is essential for ITER and for the design of any future fusion reactor.
This work demonstrates the impact of both target shaping and impurity species on the detachment onset, impurity leakage and pedestal characteristics. Experiments comparing nitrogen and neon seeding highlight that Ne dissipates further upstream than N, consistent with their difference in ionization potential and with neon leading to a higher pedestal pressure gradient. The results from these experiments are also crucial to show that the detachment front is not a purely local phenomenon, but it extends through the entire slot divertor. This study has been enabled by unprecedented diagnostic coverage in the closed divertor, which enabled multiple independent observation of plasma cooling evolution. The experimental data show strong dependence of detachment and impurity leakage on strike point location. For a database which includes power and density scans, the detachment onset consistently requires half the nitrogen amount when the outer strike point (OSP) is on the slanted inner surface, compared to the outer corner of the slot. Relative nitrogen contamination levels are reduced by 15-20% in the core under these conditions, as measured by the Charge Exchange Recombination Spectroscopy and independently confirmed by core SPRED measurements of N-IV, when the OSP is on the slanted inner surface (fig. 1a and 1b).
Insight on these experiments has been obtained by employing for the first time in DIII-D SOLPS-ITER simulations with D+C+N, drifts and n-n collisions activated. The modeling results highlight the impact of target shaping on the ionization source distribution. When the OSP is at the inboard surface, the resulting wetted target is different than with the OSP at the outer corner, thus resulting in different distributions of the ionization source.The inclusion of the drifts in the simulations enabled to study the behavior of these flows in a highly closed divertor showing the relevant role of convection on divertor asymmetry and divertor detachment. With the ion BxB directed into the divertor as in this work, drifts shift the recycling source radially towards the inner target. As such the recycling source is affected by the superposition of the closure effect and plasma drifts. This results in a redistribution of plasma flow in the SOL and divertor plasma, altering plasma profiles and shifting the parallel pressure balance between upstream and downstream. Flow reversal is found for both main ions and impurities affecting the SOL impurity transport and explaining the dependence on strike point location of the detachment onset and impurity leakage found in the experiments (fig. 1c).In addition to target shaping, the effect of different radiative species on power dissipation has been evaluated experimentally by replacing nitrogen with neon. The results show that Ne dissipates further upstream than N and thus removes the capability of the divertor to dissipate as confirmed by calculations using the 2-point model. The crucial implication is that for a fixed Te at the target, less SOL density is required with Ne which could be attractive from a core-edge integration point of view.
The two routes for dissipation we identified in the work (using N through divertor radiation and with Ne radiating mantle upstream) lead to different pedestal responses. While N does not significantly affect the profiles, Ne injection leads to an increased pedestal gradient (fig. 2).
These results are linked to the interesting observation that while neon readily enters the pedestal, nitrogen remains compressed in the divertor. This different leakage behavior between the two impurities is consistent with the higher ionization potential for Ne (21.6 eV) compared to N (14.5 eV), resulting in ionization further upstream from the target. Neon injection leads to an improved peeling-ballooning stability due to increased diamagnetic stabilization. Moreover, a self-enhancing mechanism of Ne build up has been identified as due to the increased pedestal stability and the radiative mantle. The understanding of SOL and pedestal changes associated with Ne injection as shown here is crucial to evaluate the use of neon as radiative impurity in future reactors. The findings of this work demonstrate that mitigated divertor heat flux with impurity seeding balancing core contamination, can be obtained by choosing appropriate radiative species for pedestal conditions, as well as by optimizing divertor geometry and tailoring drifts for particle entrainment.
Work funded by the US DOE under Awards DE-FC02-04ER54698 (DIII-D), DE-AC52-07NA27344 (LLNL), DE-AC02-09CH11466 (PPPL), and DE-AC05-00OR22725 (ORNL), DE-NA0003525 (SNL) and LDRD project 17-ERD-020.Speaker: Dr Livia Casali (General Atomics) -
Internal magnetic fluctuation measurements identify magnetic turbulence in the DIII-D ELMy H-mode pedestal as micro-tearing modes (MTM) and mode growth accompanied by degraded plasma confinement is observed. This work provides the first direct measurement of internal magnetic fluctuations supporting the prediction of gyro-kinetic simulations$^1$ that MTM exist in the H-mode pedestal. Using a Faraday-effect polarimeter, magnetic turbulence is observed in the edge of ELMy H-mode DIII-D plasmas. The turbulence amplitude correlates with confinement degradation in ELMy H-mode plasmas during a slow density ramp. Line-averaged fluctuation amplitude indicates the turbulence originates from electromagnetic instability. Frequency, poloidal wavenumber and propagation direction of the magnetic turbulence all agree with expectations for MTM. Magnetic turbulence amplitude non-monotonically correlates with collision frequency, peaks off mid-plane and correlates with temperature gradient evolution between ELMs, consistent with MTM features identified from theory and simulation. These internal measurements provide unique constraints towards validating models and developing physics understanding of H-mode pedestal in future devices.
Using a newly-developed high-speed Faraday-effect polarimeter, magnetic turbulence is observed to correlate with confinement degradation (Fig. 1). The polarimeter has been verified capable of directly measuring the absolute amplitude of line-averaged radial magnetic field fluctuations with wave number $k_\theta<1/cm$ and frequency up to 1 MHz. In an ELMy H-mode plasma with slow density ramp, the measured internal magnetic turbulence grows and saturates and is correlated with a ~20% drop in pedestal-top pressure and global plasma confinement. The line-averaged magnetic fluctuation amplitude after confinement degradation (>3.5 s) is ~20 Gauss over the full bandwidth, which could lead to the onset of a stochastic magnetic field and account for the observed confinement degradation, as expected for MTM.
Magnetic turbulence is observed in the edge of a wide range of ELMy H-mode DIII-D plasmas by the Faraday-effect polarimeter. Typically, the magnetic turbulence ranges from 100-500 kHz, peaking at 250 kHz. Magnetic fluctuations first appear ~2 ms after the ELM crash, grow quickly and then saturate before the next ELM event, correlating most closely with the evolution of pedestal electron temperature gradient. Density fluctuations in the same frequency range are observed simultaneously with similar temporal evolution between ELMs. BES shows the density fluctuation peaks in the pedestal steep gradient region, has wave number $k_\theta\sim0.3/cm$ ($k_\theta \rho_s\sim0.06$) and propagates in electron diamagnetic direction in plasma frame. Magnetic and density fluctuations measured by polarimeter at the same location show strong coherence up to 0.45 from 100-500 kHz, indicating they arise from the same instability.
Observations indicate the magnetic turbulence originates from electromagnetic instability. In a typical discharge, the line-averaged amplitude (averaged over the total chord length) of magnetic turbulence is 0.8 Gauss at the peak frequency and 15 Gauss integrated over the entire bandwidth (150-500 kHz). Corresponding lower-bound $|\delta b/B|$ is $4×10^{-5}$ and $8×{10}^{-4}$, respectively, and lower-bound $|\delta b/B|/|\delta n/n|$ is 0.08 and 0.15, respectively. These estimated values are comparable to that of electromagnetic instability from gyro-kinetic simulations (e.g. ref. 2).
The magnetic turbulence has been characterized and identified as MTM (Fig. 2). Theory predicts MTM are electromagnetic, propagate in electron diamagnetic direction with $k_\theta \rho_s\ll1$, driven by temperature gradient and saturate as $|\delta b/B|\simρ_e/L_{T_e},$ i.e. $\sim 1×10^{-3}$, all consistent with observations. Linear GENE calculation under the same experimental conditions find unstable MTM peak in the steep gradient region of the pedestal, with wavenumber and frequency quantitatively agreeing with the observed magnetic turbulence. The GENE simulation also shows non-monotonic collision frequency dependence of MTM growth rate, qualitatively consistent with the collision frequency dependence of turbulence amplitude in multiple shots (Fig. 3). The magnetic turbulence is always observed to peak off mid-plane, similar to MTM results from global GENE simulation$^3$.
This work provides the first direct measurement of internal magnetic fluctuations supporting the prediction of gyro-kinetic simulations that MTM exist in the H-mode pedestal. In addition, these measurements give critical constraints needed for model validation of pedestal predictions for future fusion devices. Work supported by the US Department of Energy under DE-FG03-01ER54615, DE-FC02-04ER54698 and DE-SC0018287.
References
1. M. Kotschenreuther et al., Nucl. Fusion 59, 096001 (2019)
2. W. Guttenfelder et al., Phys. Rev. Lett. 106, 155004 (2011)
3. D.Hatch et al., Nucl. Fusion 56, 104003 (2016)Speaker: Dr Jie Chen (University of California Los Angeles) -
One of the major challenges in magnetic confinement thermonuclear fusion research concerns the confinement, inside the reaction chamber, of the energetic particles (EPs) produced by fusion reactions and/or by additional heating systems, as, e.g., electron and ion cyclotron resonant heating, and neutral beam injection. In such experiments, EPs, having their velocities of the order of the Alfven velocity, can resonantly interact with the shear Alfven waves. In order to predict and, eventually, minimize the Energetic Particle (EP) transport in the next generation fusion devices, several numerical models, based on different theoretical approaches, have been developed. In this respect, it is crucial to cross verify and validate the different numerical instruments available in the fusion community. For this purpose, in the frame of the Enabling Research project MET [1], a detailed benchmark activity has been undertaken among few of the state-of-the-art codes available to study the self-consistent interaction of an EP population with the shear Alfven waves, in real magnetic equilibria in regimes of interest for the forthcoming generation devices (e.g., ITER [2], JT-60SA [3], DTT [4]). The codes considered in this exercise are HYMAGYC [5], MEGA [6], and ORB5 [7, 8], the first two being hybrid MHD-Gyrokinetic codes (bulk plasma is represented by MHD equations, while the EP species is treated using the gyrokinetic formalism), the third being a global electromagnetic gyrokinetic code (both bulk and EP species are treated using the gyrokinetic formalism). The so-called NLED-AUG [9] reference case has been considered, both for the peaked off-axis and peaked on-axis EP density profile cases, using its shaped cross section version. This case poses an exceptional challenge to the codes due to its high EP pressure, the rich spectrum of experimentally observed instabilities and their non-linear interaction [10].
Particular care has been devoted to consider plasma and numerical parameters as close as possible among the three codes: the same input equilibrium file (EQDSK) has been considered, ion density profile has been obtained by imposing quasi-neutrality ($Z_{i} n_{i} + Z_{H} n_{H}=n_{e}$), as required by ORB5 (here $n_{i}, n_{e}, n_{H}$ are the bulk ions, electrons, and EP densities (both bulk ion and EPs are assumed to be Deuterons), respectively, and $Z_{i}, Z_{H}$ their electric charge numbers); finite resistivity and the adiabatic index, $\Gamma=5/3$, have been assumed for both the hybrid codes (this is the usual choice used in MEGA, where also some viscosity is considered to help numerical convergence; note that HYMAGYC do not include viscosity).
Only finite orbit width (FOW) effects has been retained for now, and an isotropic Maxwellian EP distribution function of Deuterons with $T_{H}$=93 keV, constant in radius, has been considered.
Perturbations with toroidal mode number n=1 will be considered; the results of simulations considering both off-axis and on-axis peaked EP density profiles will be presented. First, simulation results referring to the linear growth phase will be considered.
For the peaked off-axis EP density profile case the three codes give very similar results (note that for MEGA, two MHD models are available, the 'Standard MHD' model and the 'Hazeltine-Meiss MHD'[11] model). The dominant drive comes from the positive gradient portion of the EP density profile, $0. \leq s \leq 0.4$ ($s \propto \sqrt{\psi}$ is the normalized poloidal flux function). The radial profile of the poloidal components of the eigenfuncion (see Fig.1), as obtained by the three codes compare quite well, the most unstable mode being located radially close to the magnetic axis, around $s \approx 0.2$, and in frequency within the toroidal gap.
Moreover, HYMAGYC and MEGA (both models) show a very similar growth-rate dependence on the ratio of EPs to bulk ion densities, $n_{H}/n_{i}$, while ORB5 exhibits some stronger dependence. Also, the results of varying the EP temperature will be considered.
Similar analysis have been performed for the peaked on-axis EP density profile case. Frequencies of the most unstable mode found by all codes have opposite sign, w.r.t. the off-axis case; eigenfunctions for HYMAGYC, MEGA-'Standard MHD' model and ORB5 are quite similar (they correspond to a mode located at $s \approx 0.4$, slightly inside the radial position where the $q$-profile has its minimum, $s \approx 0.5$), whereas the one shown by MEGA-'Hazeltine-Meiss MHD' model differs somehow. Growth-rates of MEGA are typically lower than the ones found by HYMAGYC and ORB5 (which are in reasonable agreement among them), and a more detailed analysis to understand such less favorable results are required. Weakly driven modes (with lower $n_{H0}/n_{i0}$ w.r.t. the nominal value) are also observed by HYMAGYC and MEGA-'Standard MHD' model, located within the toroidal gap, where the throat corresponding to $q(s)=2.5$ occurs, as already observed by HYMAGYC for the peaked off-axis EP density profile case.
Results of runs extending to the non-linear, saturation regime will also be shown, in order to benchmark these codes also in regimes where the EP transport can become relevant.Acknowledgment: This work has been carried out within the framework of the EUROfusion Consortium and has received funding from the Euratom research and training programme 2014-2018 and 2019-2020 under grant agreement No 633053. The views and opinions expressed herein do not necessarily reflect those of the European Commission. The computing resources and the related technical support used for this work have been provided by EUROfusion and the EUROfusion High Performance Computer (Marconi-Fusion).
References
[1] MET Enabling Research Project, https://www.afs.enea.it/zonca/METproject/index.html
[2] Aymar R. et al. 1997, FEC 1996, IAEA, Vol. 1, p.3
[3] JT-60SA Research Plan: http://www.jt60sa.org/pdfs/JT-60SA Res Plan.pdf
[4] DTT Interim Design Report, https://www.dtt-project.enea.it/downloads/DTT IDR 2019 WEB.pdf
[5] G. Fogaccia, G. Vlad, S. Briguglio, Nucl. Fusion 56 (2016) 112004
[6] Todo Y. and Sato T. 1998 Phys. Plasmas 5 1321–7
[7] Jolliet S. et al., 2007 Comput. Phys. 177 409
[8] Bottino A. et al., 2011 Plasma Phys. Control. Fusion 53 124027
[9] Ph. Lauber,“The NLED reference case”, ASDEX Upgrade Ringberg Seminar (2016), (Ph. Lauber et al.,
NLED-AUG reference case, http://www2.ipp.mpg.de/ pwl/NLED AUG/data.html)
[10] Ph. Lauber et al, EX1/1 Proc. 27th IAEA FEC (2018)
[11] R. D. Hazeltine and J. D. Meiss, “Plasma Confinement” (Addison-Wesley, Redwood City) p.222 (1992).Speaker: Gregorio Vlad (Associazione Euratom-ENEA sulla Fusione) -
Using modern deep neural network architectures, accurate disruption predictions on the DIII-D tokamak are made possible with the raw data from a single, high temporal resolution diagnostic which contains multi-scale, multi-physics sequences of events. This work illuminates a path forward to meet the disruption prediction requirements of devices such as ITER by using raw diagnostic data containing rich physics information without having to hand-craft features from these complex datasets generated by multiple tokamak diagnostics.
Excellent time-slice disruption prediction performance of 94% accuracy, and an F1-score of 91% (a metric indicating low false positives and false negatives rates) [Churchill2019] is achieved using a neural network architecture for multi-scale time-series, the Temporal Convolutional Network (TCN) [Bai2018] (see Fig. 1). This builds on other work using machine learning for disruption prediction [Kates-Harbeck2019, Rea2018, Ferreira2018], but in our case we show accurate performance training exclusively on a diagnostic previously unused due to its high sampling rate: the Electron Cyclotron Emission imaging (ECEi) diagnostic [Tobias2010] (see Fig. 2). The TCN is trained on a cluster of 16 GPUs over 2 days, on a subset of the available data, suggesting the need to incorporate High Performance Computing (HPC) resources to achieve the most accurate results.
These neural network disruption predictions were designed to be sensitive to pre-disruption markers up to 600ms before a disruption, giving sufficient time on DIII-D to avoid, not just mitigate, the disruption. This can then be tremendously beneficial as an accurate predictor for use in control systems to actively modify plasma actuators for avoiding disruptions. The TCN architecture can be adjusted and expanded depending on the required time sequence sensitivity, to adapt to newer machines, or different ranges of timescales.
The ECEi diagnostic is particularly well suited for disruption predictions due to its sensitivity to multiple time- and spatial-scales. There are a number of different root causes for disruptions, including edge radiation, too high density, and MHD instabilities [Boozer2012]. ECEi systems acquire data at such a high sampling rate (1 MHz) and spatial resolution ($k_\theta$ up to 2.1 cm-1) under ideal conditions that well-behaved signals span spatiotemporal scales to reflect the dynamics of turbulent fluctuations, Alfvén eigenmodes, tearing modes, sawteeth, ELMs, and other potential pre-disruption markers. Non-ideal conditions also impact the signal in ways that can be difficult for a human diagnostician to interpret, but are rich in information. For example, a sudden loss of signal can be the result of density cutoff. Alternatively, a sudden spike in signal can be the result of a non-thermal electron distribution. A wide range of other conditions can impact the signal, producing fluctuations or other features that machine learning techniques might become sensitive to, even when the human data analyst finds them to be troublesome or ambiguous. By using raw, un-processed ECEi data, we make full use of all these signal features.
These deep learning techniques show promise to greatly enhance the current machine learning worldwide effort being done on disruption predictions, one of the most pressing issues for the success of the tokamak concept. They open up the possibility of expanding the number and types of diagnostics used in disruption predictions, accurately including many different physics processes in these predictions. In the future, the correlations between channels and diagnostics that these networks learn on one machine can potentially be transferred and fine-tuned for a newer machine, requiring much less data from the newer machine then needed to originally train the models.
This work was supported in part by the US Department of Energy under AC02-09CH11466, DE-FC02-04ER54698, and FG02-99ER54531
[Churchill2019] R.M. Churchill, the DIII-D team, 2nd Workshop on Machine Learning and the Physical Sciences (NeurIPS 2019), arXiv:1911.00149
[Bai2018] S. Bai, et al., arXiv:1803.01271
[Kates-Harbeck2019] J. Kates-Harbeck, A. Svyatkovskiy, W. Tang, Nature 568 (2019) 526–531.
[Rea2018] C. Rea, R.S. Granetz, Fusion Sci. Technol. (2018) 1–12.
[Ferreira2018] D.R. Ferreira, ArXiv E-Prints ArXiv:1811.00333 (2018).
[Tobias2010] B. Tobias, et al., Rev. Phys. Instrum., 81 (10) (2010) 10D928
[Boozer2012] A. Boozer, Phys. Plasmas, 19 (5) (2012) 058101.Speaker: Randy Churchill (Princeton Plasma Physics Laboratory) -
Recent analysis leveraging the broad array of measurable plasma parameters on the DIII-D tokamak has been used to elucidate the physics underlying detachment processes in the divertor and reveal the 2D nature of detachment important for design of detachment scenarios for next step devices. The dominant role of EUV/VUV radiation for radiative power exhaust has been established experimentally with accompanying spectroscopy leveraged alongside collisional radiative modeling to calculate the impurity density and charge-state distribution in the divertor. 2D measurement of critical plasma parameters for power exhaust studies ($n_e$, $T_e$, P$_\text{rad}$, $E_\text{VUV/EUV}$) reveal a greater radial emission extent compared to UEDGE fluid modeling simulations. This larger extent provides opportunity for greater dissipation volume, but also further demonstrates that fully-2D simulations including cross-field drifts are required for detachment studies working towards a predictive capability of divertor heat loads.
A combination of EUV/VUV-VIS-IR spectroscopy, ColRadPy collisional radiative modeling (1), and 2D $T_e$ and $n_e$ measurements from Divertor Thompson Scattering has been used to infer impurity densities in the divertor. This analysis primarily uses the EUV/VUV resonance lines that make up the vast majority of radiative emission ($\sim$>95% (2)) and are particularly well suited for determining ground-state densities. Inter-ELM intrinsic carbon impurity fraction was found to be $\sim$5% in attached H-mode conditions, falling to $\sim$0.5% in detached conditions while maintaining about the same total carbon density. UEDGE modelling with a full physics drift model similarly shows a reduction in impurity concentration in detachment but limited to a $2.8 \times$ drop. Using the same set of calibrated EUV/VUV spectroscopy measurements, the carbon population in unseeded discharges is inferred to be dominated by C$^{4+}$ in the divertor whereas detached plasmas show highly radiating narrow bands of C$^{2+}$ and C$^{3+}$ at the detachment front. Figure 1 shows these charge state distributions for a partially detached plasma with $\mathbf{B} \times \nabla B$ drift towards the primary X-point in DIII-D’s ‘shelf’ open divertor (4.5MW, 1.8T, 0.9MA) alongside the associated UEDGE predictions. In nitrogen seeded detached cases the additional available charge state results in a slightly increased range of radiating species (N$^{2+}$ to N$^{4+}$ with $\sim$2eV of additional $T_e$ range) in a regime dominated by N$^{3+}$ and N$^{4+}$ ions. The charge state distribution comparison with UEDGE modeling displays quantitatively similar 2D profiles to those experimentally inferred albeit with an additional charge-state mixing caused by the finite lifetime of ions and transportation via parallel flows that are not accounted for in the CR model. Quantitative 2D comparison between UEDGE-predicted and measured flows has recently been achieved using velocity imaging (3). An excellent agreement of He$^+$ velocities in a pure helium L-mode plasma is achieved near the divertor target where He is the main-ion species and electron physics dominates. Further upstream where ion-dominated physics plays a more important role, a factor of 2–3 underestimation of the velocity is observed indicating an underestimation of the role of ions in determining local plasma characteristics near the X-point that impacts our ability to predict impurity transport via parallel flows in the divertor, estimate convective power fraction in detached conditions, and establish total pressure dissipation.
The radial extent of the radiative volume in detached H-mode discharges has been shown to display much broader features in detached H-mode discharges compared to UEDGE fluid modeling (5.2MW, 0.9MA, 1.8T, $\mathbf{B} \times \nabla B$ drift towards the primary X-point) (4) with an increasing level of broadening observed at higher powers (5). This is observed in both charge-state resolved line emission (Figure 2) as well as total radiated power (bolometry), Divertor Thomson Scatting, and 2D visible imaging. UEDGE simulations with drifts and currents show that in these conditions the poloidal $\mathbf{E} \times \mathbf{B}$ drift can dominate the poloidal heat transport in the radiative front, expanding the poloidal extent of the radiation front as well as increasing the total radiative power. This indicates drift flows leading to a larger volume for dissipation and enhanced ability for divertor radiation than predicted by more commonly used 1D and 0D modeling approximations, or 2D modeling without drifts. This directly impacts our ability to predict detachment onset, detachment stability, the impurity fraction required to achieve detachment, and the heat flux mitigation that can be expected in planned divertors.
This work was supported in part by the US Department of Energy under DE-FC02-04ER54698, DE-AC52-07NA27344, and DE-NA0003525.
(1) Johnson et al., 2019 Nuclear Materials and Energy 20 100579
(2) Mclean A.M. et al., 2018 IAEA FEC 2018 EC/PC-15; Mclean A.M. et al., 2020 Plasma Surface Interactions Conference (upcoming)
(3) Samuell C.M. et al., Phys. Plasmas, 25 056110
(4) Jaervinen A.E. et al., 2019, Contrib. Plasma Phys; Jaervinen A.E. et al. 2020 NF (submitted)
(5) Leonard A.W. et al, 2020, IAEA 2020 (this meeting)Speaker: Adam McLean (Lawrence Livermore National Laboratory) -
In tokamak discharges there are often saturated Alfven modes (1). They are destabilized by gradients of the high energy particle population, so are to be expected in discharges with a significant alpha particle population, such as expected in ITER (2) or any fusion device. These modes may produce only a small local flattening of the particle distribution, or if the number and amplitude be large enough, cause particle loss when the resonances associated with the modes overlap sufficiently to cause chaotic pathways for redistribution (3).
There are two types of events in which the \alfven modes have rapid significant effect on the high energy particle distribution, chirping modes and avalanches. A chirp is a complex frequency modulation of a single mode, causing also local particle distribution modification (4,5). An avalanche is a combination of a few modes to produce a large scale cascade of particles, leading to particle loss and even to discharge termination (6). These phenomena have been simulated using the guiding center code ORBIT (7) using a $\delta f$ formalism (8,9,10), which is able of capturing details of large phase-space structures. While not able to follow the large scale modification of the particle distribution or nonlinear mode coupling dynamics involved in the final stages of these events, this formalism is adequate to investigate the conditions for the onset of these phenomena and the initial behavior, and hence provide information as to how they can be avoided. Simulations of an NSTX (11) chirp and an avalanche are shown in Figs. 1 and 2.
An energetic particle driven \alfven mode has energy transferred from the particle distribution to the mode through distribution flattening within a resonance, producing the linear growth rate $\gamma_L$, accomplished by the fine scale phase mixing in the resonance. Once the distribution has been flattened, without collisional replenishment of the density gradient, mode drive ceases. In addition, a mode can be strongly damped due to the continuum, trapped particle effects, Landau damping and radiation, with total damping $\gamma_d$. We find that chirping occurs when a mode experiences small collisions and strong damping, such as could be due to a small equilibrium change causing contact with the continuum. A chirp of an Alfven mode involves complex particle organization so as to extract energy from the particle density gradients at the edge of the resonance island, producing a particle clump and a hole and frequency sidebands departing from the original mode frequency.
Avalanches depend on the overlap of resonances located so as to provide a path for large scale particle redistribution. Thus the possibility of avalanche depends on the nature and magnitude of the resonances associated with the modes. If this path extends to the plasma edge, particle loss can occur. During a discharge, the drive of the modes can be increasing due to changes in the high energy particle distribution. In addition, we find that the size of the resonances increases with decreasing Alfven mode frequency, inversely proportional to the square root of the plasma density. Thus the increase in plasma density and the increase in high energy particle density can both contribute to the onset of an avalanche. We show that the location and nature of the existing resonances can be determined using the equilibrium and particle distribution data, and hence the possibility of avalanche and possible means of avoidance, without extensive numerical simulation.
Acknowledgement: This work was supported by the US Department
of Energy (DOE) under contract DE-AC02-09CH11466.References:
(1) C. Cheng, L. Chen, and M. Chance, Annals of Physics 161, 21 (1985).
(2) The iter project, http://www.cad.cea.fr/gb/activites/fusion/ITER.php.
(3) C.S.Collins,W.W.Heidbrink,andetal(DIII-Dteam),Phys. Rev. Lett. 116, 095001 (2016).
(4) E. Fredrickson, R. Bell, D. Darrow, and et al, Phys. of Plasmas 13, p056109 (2006).
(5) W. W. Heidbrink, Plasma Phys. Control. Fusion 37, 937 (1995).
(6) E. Fredrickson, N. Crocker, D. Darrow, and et al, Nuclear Fusion 53, 013006 (2013).
(7) R. B. White and M. S. Chance, Phys. Fluids 27, 2455 (1984).
(8) R. B. White, The Theory of Toroidally Confined Plasmas (Imperial College Press, 2014).
(9) R. B. White, V. N. Duarte, N. N. Gorelenkov, and et al, Phys. of Plasmas 26, 092103 (2019).
(10) R. B. White, V. N. Duarte, N. N. Gorelenkov, and et al., Phys. of Plasmas x, x (2020).
(11) S. M. Kaye, M. Bell, R. Bell, S. Bernabei, J. Bialek, and et al, Nucl. Fusion 45, S168 (2005).
(12) H. L. Berk, B. N. Breizman, and V. Petviashvili, Phys Lett A 6, 3102 (1999).Speaker: Roscoe White (Princeton University) -
A portable and interpretable data-driven algorithm for disruption prediction has been developed and installed in DIII-D and EAST Plasma Control Systems (PCS). The Disruption Prediction via Random Forest (DPRF) algorithm guarantees explainable predictions in real-time thanks to the feature contribution analysis [Rea2019], which is able to identify the main drivers of the disruptivity – the probability of an impending disruption – potentially informing the PCS on the proper actuators. DPRF was tested in closed-loop mitigation experiments on both DIII-D and EAST tokamaks. On DIII-D, indicators of plasma temperature, density and radiation based on profile peaking factors [Pau2018] were successfully implemented in real-time and used as input features in DPRF during recent experiments.
DPRF is based on the Random Forest machine learning algorithm, which estimates the probability of an impending disruption, i.e. disruptivity, by developing a large number (~ hundreds) of independent, de-correlated base learners (models), thus collecting a parallel set of predictions. The final prediction from the ensemble is aggregated by averaging this very large number of models’ predictions. A calibrated threshold on the disruptivity is then found to maximize the success rate and warning time [Montes2019]. While it is necessary to provide enough warning time to predict an impending disruption, it is crucial to also diagnose what type of event in the disruptive chain is causing the deviation from the safe operational space. This allows the PCS to respond adequately to different instabilities by directing it to trigger different actuators. 1D/2D profiles of plasma temperature, density and radiation contain highly predictive information that can be used for example to diagnose impurity accumulation events leading to radiative collapse and therefore disruptions. Reducing profiles to their peaking factors (following what documented in [Pau2018]) handles the curse of dimensionality and represents a manageable and real-time compatible way for a data-driven algorithm to augment predictive capabilities. In addition to these synthetic diagnostics, other 0D (where time is their only dependence) and mainly dimensionless input features were used to train the machine learning-based algorithm on DIII-D, consisting in data coming from equilibrium reconstruction codes or magnetic probes.
On DIII-D, DPRF was tested in ITER Baseline Scenario plasmas, where closed-loop performances were verified by successfully integrating the disruptivity in the Off Normal Fault Response algorithm [Eidietis2018] for emergency response. In different discharges, the Massive Gas Injection (MGI) was triggered in response to dangerous tearing mode activity, the Electron Cyclotron Heating (ECH) system was aimed at q=3/2 surface when such precursor emerged, and an early shutdown via fast ramp down was triggered also due to the presence of a tearing mode (see Fig. 1a). Additionally, the peaking factors were verified to represent a good metric for impurity accumulation events, when probing via Argon injection – see Fig. 1b. The successful integration of DPRF in DIII-D PCS is part of a broader approach to qualify advanced disruption prevention strategies to address ITER’s critical needs. More details about the “Disruption Free Protocol” can be found in the contribution by J. Barr et al IAEA-FEC 2020.
On EAST, DPRF was tailored to predict high-density scenarios and closed-loop performances were tested during recent mitigation experiments, where several disruptions were mitigated via MGI. In Fig. 2 we show one example of a real-time experiment in January 2020, where predictions and interpretations were provided in around 200 microseconds computing time. EAST discharge 94520 shows symptoms of a density-limit disruption (high contribution from the Greenwald density fraction) but eventually disrupts due to a radiative collapse caused by impurities in the plasma (dominance of loop voltage contribution at the end). The developed algorithm still presents some residual flaws as it shows to be prone to false alarms when H-L back-transitions occur during the discharge. More work will be required to also implement analogous peaking factors synthesized from EAST diagnostic systems, crucial to define a device-independent framework for properly diagnosing disruption precursors.
Current results represent a step forward in data-driven control for scenario optimization and disruption avoidance for ITER and next generation devices. This work establishes the importance of developing tools capable of identifying and informing in real-time the PCS on the dangerous plasma parameters deviations to the disruptive space to enable the proper actuators’ response. Effort is ongoing to map the feature contributions to the actions of the available actuators, but also to define a robust framework to transfer the knowledge gathered using data from existing experiments to unseen devices.
This work was supported in part by the US Department of Energy under DE-FC02-99ER54512, DE-SC0014264, DE-SC0010720, DE-SC0010492, DE-FC02-04ER54698 and by the National MCF Energy R&D Program of China, Grant No. 2018YFE0302100.
[Rea2019] C. Rea et al 2019 Nucl. Fusion 59 096016
[Pau2018] A. Pau et al 2018 IEEE TPS 46 7 2691
[Montes2019] K.J. Montes et al 2019 Nucl. Fusion 59 096015
[Eidietis2018] N.W. Eidietis et al 2018 Nucl. Fusion 58 056023Speaker: Cristina Rea (Massachusetts Institute of Technology) -
New studies identify the critical parameters and physics governing disruptive neoclassical tearing mode (NTM) onset. A m/n=2/1 mode in DIII-D begins to grow robustly only after a seeding event (ELM Fig. 1, or sawtooth precursor and crash Fig. 2) causes the mode rotation to drop close to that of the plasma’s Er=0 rest frame; this condition opens the stabilizing ion-polarization current “gate” and destabilizes an otherwise marginally stable NTM. Our new experimental and theoretical insights and novel toroidal theory-based modeling are benchmarked and scalable to ITER and other future experiments. The nominal ITER rotation at q=2 is found to be stabilizing (“gate closed”) except for MHD-induced transients that could “open the gate.” Extrapolating from the DIII-D ITER baseline scenario (IBS) discharges, MHD transients are much more likely to destabilize problematic 2/1 NTMs in ITER; this makes predictions of seeding and control of both ELMs and sawteeth imperative for more than just “simply” minimizing divertor pulsed-heat loading.
While nearly steady state in betaN, li and rotation, the classical tearing stability index Delta’r0 in DIII-D may evolve slowly. IBS discharges (ITER similar shape, H-mode with ELMs and sawteeth, betaN~2 and q95~3) were run with 1 msec faster resolution CER (standard 5 msec) of toroidal and poloidal rotations. Discharges can exhibit a rotating m/n=2/1 magnetic perturbation, in response to sequential ELMs and sawteeth, that evolves from marginal to robust growth into an eventual locked mode and disruption. Key conditions for algebraic linear temporal growth include an MHD-induced transient with a large enough magnetic perturbation Brms and mode rotation change. Figures 1 and 2 show analyses of NTM seeding events with the fast CER that have been examined in greatest detail, among the multiple discharge broader database.The key stabilizing factor depends on the relative rotation [parameterized by a gate function F(fm)≤1] between the mode rotation frequency fm and the Er=0 frame of the plasma fE being non-zero but not more than that of the (positive in DIII-D) ion diamagnetic mode frequency f*i ~ 1 kHz in Figs. 1 and 2. The mode island width growth rate dw/dt is modeled with a modified Rutherford equation (MRE) along with models for changes in both mode and plasma rotation:
$\begin{equation} \frac{dw}{dt}=D_{\eta}\left[\frac{d_{NTM}}{w}-\frac{w^2_{pol}}{w^3}F(f_m)\right],\;\;F(f_m) \equiv -4\frac{(f_m-f_E)(f_m-f_E-f_{*i})}{f^2_{*i}} \;\;\;\;\;\;\;\;\;\; \end{equation}$
Here, the classical tearing stability index is negligible, i.e., Delta’r0=-0.1. The critical island for growth is w0=wpol(F/dNTM)1/2 where dNTM is bootstrap drive and wpol is 2X the ion banana width wib. A toroidal adaptation of recent slab-model theory (Beidler 2018) predicts MHD transients abruptly induce a 2/1 tearing response, radially local torque δJ_∥ δB_x, radial electric field and flows that reduce the relative mode frequency. The mode rotation dynamics is described by
$\begin{equation} \frac{df_m}{dt}\cong-\frac{f_m-f_t}{\tau_\zeta}-\frac{f_m-f_E}{\tau_w}+\delta J_\parallel\delta B_x,\;\; \tau_\zeta\cong\frac{a^2_{eff}}{4D_\mu},\;\;\frac{1}{\tau_w}\cong C_w\frac{v_{Ti}}{Rq}\frac{w}{r_0} \;\;\;\;\;\;\;\;\;\; \end{equation}$
Fits to experimental data (for an ELM-induced NTM in Fig. 1 as an example) capture the experimental behavior (pink lines in Fig. 1) of evolution from marginal to robust 2/1 growth. The stabilizing gate factor F depends on the relative rotation and plunges during an ELM, either recovering or remaining down for large enough mode amplitude. In addition to theory, the NIMROD code is used to study evolution at the 2/1 surface in response to an ELM. The code uses an extended MHD model with heuristic closures to model the electron and ion neoclassical parallel stresses. NIMROD indicates the Fig. 1 DIII-D equilibrium is stable to classical tearing modes but a pulsed MHD perturbation at the computational boundary can kick off a 2/1 mode.
Applying a MHD transient torque to the situation in Fig. 3 can drive relative rotation down and open the gate, as in Fig. 4. The predicted island growth rate in ITER is slower due to its much smaller magnetic field diffusivity but shifted to smaller w0 and very much smaller (0.1X) relative size w0/r0. The IBS equilibrium in DIII-D is very similar to what is modeled (Polevoi 2006, 2019) for ITER. Similar j and q profiles imply comparable classical tearing stability Delta’r0=-0.1 in ITER which is neglected in the MRE of Eq. 1. The ratio of q=2 bootstrap to equilibrium current density (~d_NTM) is also similar. At critical island w0 for F≅1, DIII-D mode rotation fm is about 0.5Xfi=0.6 kHz above fE (Fig. 3); ITER is similar with fi of 0.165 kHz.
This new work gives experimental and theoretical insights, as well as novel benchmarked toroidal theory-based modeling, to a longstanding uncertainty in projecting how NTMs are triggered (Buttery 2007, Hender 2007) for scaling to ITER and beyond. It also provides the framework (e.g., for real-time monitoring) to develop criteria for transient-MHD-induced excitation and robust growth of 2/1 NTMs that can lead to problematic locked modes and disruptions in burning plasma tokamaks, and for which experimental data is limited.This work was supported in part by the US DOE under DE-FC02-04ER54698.
Speaker: Dr Robert La Haye (General Atomics) -
The dynamics of fast electrons driven inductively or by resonant interactions with radio- frequency waves is known to be highly sensitive to the presence of impurities in hot magnetized hydrogen plasmas. The possibility to use tungsten for the ITER divertor, thanks to its low tritium retention and high melting temperature, has raised the question of the impact of partially ionized high-Z atoms on current drive efficiency by enhancing pitch-angle scattering but also collisional slowing-down. Pioneering work on the impact of the partial screening effect in kinetic calculations was carried out primarily for the problem of runaway electron mitigation in very cold post-disruptive plasmas [1]. In this paper the approach is adapted and extended to regular plasma regimes, allowing to take into account any type of high-Z metallic impurity in the plasma core on the fast electron dynamics. In addition, the enhancement of non-thermal bremsstrahlung by partially ionized high-Z atoms in the plasma is calculated.
Effect of partial screening is investigated in the framework of the Born approximation by calculating the usual form factor to account for the spatial extent of the high-Z ion using atomic electron densities deduced from simplified models (Thomas-Fermi, Yukawa potential) or from the Density Functional Theory (DFT) describing accurately many-body exchange and correlation interactions. In order to reduce computational effort either for kinetic calculations or bremsstrahlung emission, analytical formulas of the form factor deduced from Thomas- Fermi and Yukawa potential atomic models are used, with effective ion sizes determined for each ionization state by a best fit of the form factor deduced numerically from quantum relativistic calculations using the GAUSSIAN chemistry software package [2]. While Thomas- Fermi model form factor gives better results when high-Z atoms are weakly ionized, Yukawa potential model turns out to be more appropriate when the screened ion charge is greater than Z/3, where Z is the atomic number, a condition encountered for tungsten in standard core tokamak plasma conditions where electron temperature reaches few keV.
From the spinless relativistic Rutherford elastic scattering cross-section, the modified pitch-angle collision operator is used in Fokker-Planck calculations, taking into account of the partial screening for each species and all corresponding ionization states. It is proportional to the factor (Z-N)2lnΛ(p)+g(Z-N,p), where lnΛ(p) is the usual momentum-dependent Coulomb logarithm, N, the number of bound electrons, and g an analytical function describing the enhanced pitch-angle scattering by inner populated atomic shells. When g is small as compared to the usual Coulomb logarithm term, screening effect on pitch-angle is small. Inelastic scattering resulting from the mean excitation of partially-ionized high-Z atoms is also considered from the Bethe formula for the electron relativistic stopping power.
For non-thermal bremsstrahlung, Yukawa potential model is usually preferred for describing screening effect of bounded electrons, since inner atomic shells contribute significantly, whatever the ionization state [3]. In this case, an original semi-analytical formula is derived for the doubly differential quantum relativistic cross-section in photon energy and in photon emission angle from the most general Bethe-Heitler bremsstrahlung cross-section [4], which greatly enhances calculation speed, while keeping a high numerical accuracy. It is shown that bremsstrahlung scales like Z^2, with an enhanced reduction factor as the ratio k/Ec decreases, where k is the photon energy and Ec the incoming fast
electron kinetic energy (Fig. 1). Screening effects tends to disappear progressively when the angle between the photon emission and the incoming electron velocity increases.
Consequences on the current drive efficiency have been investigated using the kinetic solver LUKE of the 3-D linearized relativistic bounce-averaged electron Fokker-Planck equation [5] and on the fast electron bremsstrahlung using the quantum relativistic radiation code R5-X2 [6]. Thermal ionization states for all species are determined by ADAS code [7]. A full simulation of the high-power WEST tokamak discharge #55539 is investigated, where most of the plasma current is driven by the Lower Hybrid wave, taking into account of the tungsten level deduced from radiative power losses using the METIS tokamak code [8]. With an estimated concentration of tungsten of 4x10-4, it is shown that the reduction of the LH driven current is about 4%, while conversely, bremsstrahlung is increased by a factor 3 approximately as compared to the fully screened ion case. From simulations of high-power WEST tokamak Lower Hybrid discharges [9], the general impact of partially ionized metallic impurities on RF current drive is discussed, as well as on fast electron bremsstrahlung diagnosis capability.
Acknowledgements. This work has been partially funded by National Science Centre, Poland (NCN) grant HARMONIA 10 no. 2018/30/M/ST2/00799. We thank the PLGrid project for computational resources on the Prometheus cluster.
References
1 L. Hesslow, et al., Phys. Rev. Letter 118 (2017) 255001
2 M. J. Frisch, et al., Gaussian 09, Revision E.01, Gaussian, Inc., Wallingford CT, 2016. 3 M. Lamoureux and N. Avdonina, Phys. Rev. E, 55 (1997) 912
4 H. W. Koch and J. W. Motz, Rev. Mod. Phys., 31 (1959) 920
5 Y. Peysson and J. Decker, Fusion Science and Technology, 65 (2014) 22
6 Y. Peysson and J. Decker, Phys. Plasmas, 15 (2008) 092509
7 H. P. Summers, the ADAS User Manual, version 2.6 http://www.adas.ac.uk (2004)
8 J.F. Artaud, et al., Nuc. Fusion 58 (2018) 105001
9 C. Bourdelle, et al., Nucl. Fusion 55 (2015) 063017Speaker: Yves Peysson (CEA) -
ELM control and mitigation efforts are of highest priorities for ITER and beyond. However, often such efforts with RMPs and pellets do not focus on the pedestal turbulence and transport and their effect on ELM characteristics. Detailed analysis of pedestal turbulence as a function of the heating mix may provide crucial insights as well as additional handles for ELM control. It is found that recovery of pedestal pressure gradient in the inter-ELM phase is delayed and magnetic and density fluctuations increase along with excitation of several quasi-coherent modes in the ECH dominated discharges in DIII-D. As a result, ELM frequency decreases by 40% when heating is changed from pure NBI to predominantly ECH. This would be consistent with our hypothesis that turbulence driven transport increases with these quasi-coherent modes and thus keeping the pedestal away from ELM threshold for a longer duration of time, as shown in Fig. 1(a). In this study, localization of these modes at the pedestal is confirmed. Growth of these modes has some correspondence with the steepening of
$\nabla T_{e}$at the pedestal with ECH injection. That these mode activities may lead to enhanced transport at the pedestal steep gradient region in the inter-ELM period is supported by the baseline of the D$_{\alpha}$signal, which indicates increased particle flux.
Fig. 1 (right) shows the spectrogram of density fluctuations from a Doppler Backscattering (DBS) diagnostics channel at the steep gradient region ($\rho$= 0.95) from a candidate ECH shot. It can be seen that there are two modes evolving in the inter-ELM period: following the ELM crash, a low frequency (400 kHz) quasi-coherent mode (LFQC; black oval) is observed in the$\nabla T_{e}$phases #1b and #4. It is apparent that the LFQC emerges and survives in a narrow range of values of$\nabla T_{e}$, roughly bounded by the two horizontal broken blue lines on$\nabla T_{e}$. On the other hand, high frequency (~2 MHz) broadband fluctuations (HFB; white oval) dominates in phases #2 and #5 leading towards the small Dα spike and the large type-I ELM respectively. Whenever, the LFQC is present (phases #1b and #4), the D$_{\alpha}$baseline is enhanced indicating increased particle flux. Based on this evolution of$\nabla T_{e}$the occurrence of these two modes appears consistent with affecting transport and thereby frequency and amplitude of the ELMs. In phase #2, the gradients have reached the threshold for ELMs to occur, while the pedestal heights, especially$n_{e}$and hence,$p_{e}$pedestal heights, are still low and evolving. Hence, one or two small spikes in D$_{\alpha}$are observed in between two consecutive large type-I ELMs.A similar correlation is observed between magnetic fluctuations and
$\nabla T_{e}$, as seen in the DBS data. A distinct group of three modes (13~116 kHz) is seen in magnetic fluctuations in Fig. 2. ELM-synced analysis shows that these modes briefly grow and die down in the first 2-7 ms of the inter-ELM period. After that the modes grow again from 13 ms onwards and saturate at ~25 ms. The initial growth bump at ~5 ms and the growth from 13 ms are in$\nabla T_{e}$phases #1b and #4, similar to the growth patterns of the LFQC in DBS, as shown in Fig. 1.
The heat and particle diffusivities are calculated from the measured profiles in TRANSP. The electron particle diffusivity ($D_{e}$) is much less than the electron heat diffusivity ($\chi_{e}$) in the steep gradient region. As per the recently proposed fingerprint analysis approach [M. Kotschenreuther et al., Nucl. Fusion 59, 096001 (2019)], this suggests that the transport is possibly of the nature of$\nabla T_{e}$driven, dominated by TEM and MTM. TGLF simulations show that the linear growth rate of the most dominant mode peaks at$k_{\theta}$ρs ~ 0.4 at the steep gradient region. Corresponding frequency is in the electron direction over the entire pedestal for$k_{\theta}$ρs < 1.5, indicating that TEM and/or MTM could be important. This study may provide vital inputs towards understanding inter-ELM pedestal recovery in varied transport regimes and predicting/optimizing pedestal performance and ELM behavior in future reactor grade plasmas. This work is supported by US DOE under DE-SC0019302, DE-FG02-08ER54999, DE-FG02-08ER54984, DE-AC02-09CH11466 and DE-FC02-04ER54698.Speaker: Dr Santanu Banerjee (College of William and Mary) -
We present novel techniques for fast-ion modelling that allow more extensive studies and support orbit-following modelling [1], and we apply those to study fast-ion transport in ITER.
While orbit-following Monte Carlo simulations are frequently used to make predictive estimates for fast-ion losses and wall loads, these simulations have the drawback that they are computationally expensive to perform. Furthermore, orbit-following tools are based on first principles and, as such, it is difficult to interpret which processes are responsible for the losses seen in simulations. This contribution addresses these issues.
It is shown that the collisionless transport of fast ions can be treated as an advection-diffusion process, where the transport coefficients can be evaluated with an orbit-following model. This results in a significant (a factor of 50 - 100) decrease in time that is required to estimate fast ion losses during the slowing-down process with an orbit-following model alone as the coefficients can be evaluated within just a few bounce times. With this approach it becomes possible to make an extensive study of fast-ion losses due to ITER ELM control coils using different current configurations. We vary the mode and poloidal phase of the coils and find the configurations where the losses have their minimum and maximum as shown in Fig.3. However, the study performed here is done in vacuum approximation while plasma response is needed for more accurate studies [2].
Another technique that is presented is the so-called loss maps where the birth position of particles, which are lost due to a 3D magnetic field, are mapped to a constants of motion phase space. We show that this mapping allows one to make the connection between different loss-mechanisms and the losses, thus increasing the confidence in one's results. Furthermore, we show that the particle birth position also predicts to which location on the wall it will be lost to. However, the main benefit of using the loss maps is that they can be constructed solely using the analytical formulas for different transport processes as exemplified in Fig.4. This provides a basically instant way of estimating losses, thus avoiding the need to resort to more time-consuming orbit-following simulations. The use of loss maps is demonstrated for ITER baseline and reduced field scenarios in the presence of various stationary magnetic field perturbations.
[1] Särkimäki, K. (2020). Efficient and rigorous evaluation of fast particle losses in non-axisymmetric tokamak plasmas. Nuclear Fusion, 60(3), 036002.
[2] Varje, J., et al (2016). Effect of plasma response on the fast ion losses due to ELM control coils in ITER. Nuclear Fusion, 56(4), 046014
Speaker: Konsta Särkimäki -
In this paper we report on experimental and modeling results concerning the energetic particle (EP) dynamics in plasma scenarios with off-axis neutral beam (NB) injection at ASDEX Upgrade (AUG). The tools validated in this processes are applied to selected scenarios at JT-60SA and ITER pre-fusion plasmas.
Off-axis NB injection is an important tool to control and optimise the current profile in both conventional and advanced tokamak scenarios. Via tailoring the safety factor profile, rational surfaces can be avoided, the local magnetic shear can be changed or reversed shear regions can be established. Whereas in present devices the beam energies are typically 10-20 times larger than the plasma thermal energies and smaller than the Alfvén velocity (vNBI /vA ∼ 0.3 − 0.4), in future devices such as JT-60SA and ITER these ratios will go up to 100 for vNBI/vthermal and to vNBI/vA ~ 1. Thus, it is expected that the related EP-driven instabilities and the relaxation of the spatial EP pressure gradients will be different (e.g. mode number spectrum, non-linear saturation) than in present-day experiments.The related EP transport is directed to deplete the gradients, i.e. inwards in the positive gradient region and outwards in the negative gradient region. Since the redistribution of the EP beam will affect the background plasma properties through various channels, it is of interest to analyse the EP redistribution and to test if stability predictions and related EP transport calculations are able to catch the experimental signatures and thus can be used with confidence in future comprehensive scenario simulations.
In 2017 a new scenario on ASDEX Upgrade has been established for the dedicated investigation of energetic particle (EP) physics [1] that is optimised to maximise βEP/βth and the ratio vNBI/vthermal. This scenario has been recently further developed into both an L-mode and an H-mode scenario with stable flat-top phases and with more complete diagnostic coverage. As in the previous discharges, we let impurities (mainly tungsten) accumulate in the core. Due to strong radiation losses the background temperatures and pressures of both ions and electrons stay low, despite 2.5 − 5 MW NB heating. In order to avoid transient q = 2 sawtooth-like crashes, as seen previously with a total plasma current of 800kA, the current has been reduced to 700kA. Under these conditions we reach an EP-β comparable to the background β and ratios of 100 and larger for vNBI/vthermal, whereas vNBI /vA ∼ 0.3 − 0.4. A rich spectrum of modes is destabilised: EP-driven geodesic acoustic modes (EGAMs), beta-induced Alfvén eigenmodes(BAEs), reversed shear Alfvén eigenmodes (RSAEs) and toroidal Alfvén eigenmodes (TAEs). In particular the EGAM onset is triggered by TAE bursts indicating a coupling of these modes via EP phase space transport. Bicoherence analysis also reveals non-linear coupling signatures between various frequency bands. Comparisons of FIDA-measured EP profiles and neoclassical calculations are shown, and indications for anomalous background ion heating via EP-driven instabilities are investigated. Using the linear gyro-kinetic code LIGKA [2], the onset conditions for various instabilities are analysed, as the experiment provides excellent data in this respect. Experimental mode properties are determined and compared to the LIGKA results. In particular, symmetry breaking signatures of non-perturbative mode structures [4] are investigated. Non-linear hybrid simulations (HAGIS/LIGKA [2, 3]) are carried out in order to quantify the difference between experimental measurements and this model. By comparing to other models such as non-linear hybrid kinetic MHD models and gyrokinetic codes, the importance of e.g. non-linear wave-wave coupling processes can be inferred. This analysis is linked [5] to a benchmark and validation exercise [6] including the codes HYMAGIC, MEGA and ORB5, that is based on this AUG case (see also refs in [6]).
The exploration of scenarios with off-axis NB deposition leading to non-inductive steady-state op- eration at high β is one of the main missions of the JT-60SA project starting operation in 2020 [7]. The high-energy negative ion sources (~500 keV) at JT-60SA deposit exclusively off-axis. An exhaustive kinetic-hybrid MHD analysis using the MEGA code has been performed in refs. [8]. Building on these results further gyro-kinetic analysis using the LIGKA/HAGIS tool is carried out (higher mode numbers, EGAM/BAE thresholds) in order to investigate the scaling of various parameters connected to linear onset and non-linear EP transport compared to the AUG case discussed above.
During the lifetime of ITER various scenarios with off-axis NB injection are forseen. Although a change of the beam geometry from on-axis to off-axis will be possible, it cannot be performed frequently since the cycles are limited due to the mechanical stress it induces on the various compo- nents connected to the beam source. For this reason a good understanding of the expected heating characteristics and deposition properties including a possible deposition broadening due to EP- driven instabilities can help to optimise the planning and operation of the experiments. Based on the fully IMAS [9] integrated heating and current drive workflow [10] the stability of pre-fusion H-plasmas is investigated for different beam deposition locations (off-axis, mixed on/off-axis). The analysis will also serve as a verification test of the recently IMAS-integrated LIGKA/HAGIS workflow (see fig. 1). The results will be related and compared to the findings on AUG and JT-60SA.Acknowledgements: This work has been carried out within the framework of the EUROfusion Consortium and has received funding from the Euratom research and training programme 2014- 2018 and 2019-2020 under grant agreement No 633053. The views and opinions expressed herein do not necessarily reflect those of the European Commission. ITER is the Nuclear Facility INB no. 174. The views and opinions expressed herein do not necessarily reflect those of the ITER Organisation.
References:
[1] Ph.Lauber et al, EX1/1 Proc. 27th IAEA FEC (2018)
[2] Ph.Lauber et al, J.Comp.Phys.,226/1 (2007)
[3] S.D.Pinches, Comp.Phys.Comm.,111 (1998)
[4] Z.Lu et al, PPCF 25 012512 (2018)
[5] Eurofusion Enabling Research Projects ’NAT’ and ’MET’ (2017,2019)
[6] G.Vlad et al, this conference
[7] JT-60SA Research Plan v4.0, September(2018)
[8] A.Bierwage et al, PPCF 59 125008 (2017), PPCF 61 014025 (2019)
[9] S.D.Pinches et al, TH/P6-7 Proc.27th IAEA FEC(2018)
[10] S.D.Pinches et al, this conference
[11] ARTAUD,J.F. et al., Nucl.Fusion 58 105001 (2018)Speaker: Philipp Lauber (IPP Garching) -
Controlled mitigation of heat load has been demonstrated for the first time by doping a closed divertor$^{1}$ plasma at DIII-D with low Z impurity powders. Injection of low-mid Z impurities is a technique under investigation to address the issue of power exhaust in ITER and next-step fusion reactors$^{2,3}$. The use of non-recycling impurities in powder form provides a new capability extending the limited number of impurity species and mixtures usable with conventional gas injection$^{4}$. At DIII-D, it has now been shown that local doping of the outer strike point (OSP) with lithium (Li), boron (B), and boron nitride (BN) powders can be used to enhances radiative cooling of the divertor plasma. In particular, using B as a radiator has the advantage of concentrating the radiation zone more rooted in the scrape-off layer (SOL), allowing for high core performance and high dissipation in the boundary and divertor at the same time$^{5}$. Moreover, Li and B have beneficial effects in terms of wall conditioning and ELM mitigation and suppression$^{6,7}$.
The experiments focused on ELMy H‑modes (I$_{p}$~1 MA, B$_{t}$=2 T, P$_{NB}$~6 MW, f$_{ELM}$~80 Hz, n$_{e}$~3.6-5.0$\cdot$10$^{19}$ m$^{-3}$), confined in upper-single-null geometry with the OSP positioned in the small angle slot (SAS)$^{1}$. The geometry of the SAS closed divertor, the location of the OSP at the SAS target, the position of the Langmuir probes, the injection location, and the magnetic equilibrium are shown in figure 1 a, b. In this configuration, powders were dropped with rates of 1-50 mg/s directly into the OSP region, allowing to study near-target power dissipation and divertor leakage. For all impurity species, injection caused substantial drops in downstream electron temperature (T$_{e}$), particle fluxes (J$_{sat}$), and heat fluxes q$_{||}$, as measured by a densely spaced array of Langmuir Probes in the SAS (figure 1 a). The time traces of these downstream parameters measured poloidally along the SAS wall, are shown in figure 1 c-f for a discharge with BN injection at t~3 s, causing a sustained strong drop of heat and particle fluxes at the OSP.
In this and similar cases, features characteristics of detachment, such as a strong increase of near-target neutral pressure, were also observed. In general, the high divertor radiation state could be achieved with a 2-15% degradation in H-mode energy confinement.To study the impact of the impurities on SAS dissipation and examine the effect of different impurities’ radiation efficiency and spatial distribution, these scenarios have been analyzed with the 3D plasma-fluid and kinetic neutral edge transport code EMC3-EIRENE. In a first step, transport simulations have been conducted in an axisymmetric geometry, for upstream densities of 1-3$\cdot$10$^{19}$ m$^{-3}$ with B, N, Li as main impurities separately to survey the distribution of the radiated power of each impurity individually. The 2D emissivity patterns for Li and B are shown in figure 2 for an upstream density of 2$\cdot$10$^{19}$ m$^{-3}$. The Li radiation is concentrated near the separatrix and at the strike line while the B peak emission front is concentrated in the far SOL and outer tail of the main recycling region. These characteristic features, qualitatively consistent with measurements of spectral divertor imaging, are a result of the different cooling efficiencies of these two impurities combined with transport effects.
The combined set of experimental results and modeling indicates that injection of low Z impurities in powder form can be a novel effective method to enhance and control of radiative dissipation in closed divertor configurations. This first-time assessment of the effects of different impurity powder species supports the development of new divertor solutions avoiding impurity leakage from the divertor and maintaining high performance.
(1) H.Y. Guo et al 2017 Nucl. Fusion 57 044001
(2) A. Kallenbach et al 2013 Plasma Phys. Control. Fusion 55 124041
(3) L. Casali et al. Contributions to Plasma Physics 2018; 58: 725– 731
(4) A. Nagy et al Rev. Sci. Instrum., 10K121 (2018) 89
(5) A. Yu. Pigarov 2017 Phys. Plasmas 24, 102521
(6) A. Bortolon et al., Nucl. Mater. and Energy, 384-389 (2019) 19
(7) R. Maingi et al., Nucl. Fusion, 024003 (2018) 58- This work is supported by the US Department of Energy under DE-AC02-09CH11466, DE-FC02-04ER54698, DE-AC05-00OR22725, DE-SC0020357, DE-SC0020284, DE-AC52-07NA27344 and DE-FG02-07ER54917.
Speaker: Dr Florian Effenberg (Princeton Plasma Physics Laboratory, Princeton, NJ, 08543 USA) -
Energetic particles (EP) represent the main source of heat and momentum for burning plasmas. However, EPs can drive instabilities that, in turn, can cause redistribution and loss of EPs. The reduced physics, energetic particle kick model for EP transport enables interpretive and predictive capabilities for time-dependent integrated tokamak simulations including the effects of EP transport by instabilities [Podestà 2017]. Over the last two years, the model has been extended include effects of low frequency instabilities in addition to Alfvénic modes, thus providing a common framework to simulate the possible synergy between different types of instabilities. The model is based on transport probability matrices describing the effects of instabilities in the Monte Carlo module NUBEAM of TRANSP. Matrices are computed for each scenario through particle following codes. The kick model has demonstrated the importance of phase-space resolved simulations of EP dynamics to unravel details of EP transport for detailed theory/experiment comparison [Podestà 2017][Heidbrink 2018] and for scenario planning, e.g. based on the optimization of NBI parameters [Podestà 2019]. Enhancements to TRANSP via the inclusion of the kick model enables scenario development and predictions including realistic treatment of EP transport by instabilities, which is required for reliable and quantitative projections of present results to ITER and next-step devices.
The kick model has been mostly used to simulate the effects of Alfvénic instabilities (AEs) in NSTX/NSTX-U and other tokamaks, enabling detailed studies of the mechanisms driving the instabilities and of the resulting EP transport [Podestà 2017][Podestà 2019]. Recently, the model has been extended to include perturbations such as Neoclassical Tearing Modes (NTMs) [Heidbrink 2018][Bardóczi 2019], sawteeth [Kim 2019] and kink/fishbones [Podestà 2019]. Work is also in progress to extend the model outside the separatrix to include externally imposed 3D perturbations and toroidal field ripple. These developments have resulted in a unique tool to study the synergistic effect of multiple instabilities from the plasma core out to the edge, which can greatly enhance EP transport $\&$ loss and degrade plasma performance.
Figure 1 shows a spectrum of magnetic fluctuations and simulation results from a Neutral Beam heated NSTX-U plasma. Multiple instabilities are destabilized, including toroidal AEs (TAEs), fishbones and kink modes [Podestà 2019]. In the simulations, TAE radial structure and damping rates are obtained from NOVA-K [Cheng 1992][Gorelenkov 1999]. Amplitude of TAEs is inferred by balancing the damping rate from NOVA-K with the drive from NUBEAM [Podestà 2017]. Simple analytic expressions are used for the kink/fishbone radial structure. Their amplitude is adjusted to match the measured neutron rate and stored energy.
‘Classical’ TRANSP simulations that do not include EP transport by instabilities over-estimate the measured neutron rate (Fig. 1b). When both TAEs and fishbones/kink are included, the simulation agrees well with the measurements. By enabling only specific types of modes, their relative role in causing EP transport and loss can be studied (Fig. 1c). Fishbones/kink modes appear to be the primary cause of EP losses. Notably, the loss rate for the case including all perturbations does not equal the sum of loss rates obtained with specific modes only. This indicates a synergy between different modes. The phase space resolution implemented in the kick model enables a detailed analysis of the mechanisms leading to enhanced EP losses. Figure 2 shows examples of root-mean-square energy kicks vs EP phase space for a $n=1$ kink and a $n=4$ TAE mode. Because of their spatial localization, the kink and TAE modes cause larger kicks near the axis and from mid-radius to the plasma edge, respectively. The combined effect of the two modes is therefore to open up an efficient loss channel that connects the plasma core to the edge.
In addition to the effects of AEs and kink/fishbones, the kick model has also been extended to NTMs [Heidbrink 2018][Bardóczi 2019][Yang 2018].
Figure 3a shows magnetic fluctuations for a NSTX discharge in which a 2/1 NTM is destabilized towards the end of the pulse. Analysis of 2D soft X-ray (SXR) emission provides the width of the NTM island (Fig. 3b). Mirnov coil data are then rescaled based on results from SXR analysis, which provides the time-dependent island width information for the simulations. The measured and simulated neutron rate (Fig. 3c) shows that the NTM causes a maximum decrease in neutron rate of $\sim 10\%$ with respect to ‘classical’ simulations. Significant deviation from the classical run is only observed for island width exceeding $\sim 8$ cm, indicating a possible threshold in mode amplitude for EP transport. Below that threshold, the1 mode causes a local EP redistribution with no adverse impact on the overall EP confinement [Yang 2018]. Future work will extend these results to cases with other instabilities in addition to NTMs.
This work is supported by the U.S. Department of Energy, Office of Science, Office of Fusion Energy Sciences under contract number DE-AC02-09CH11466.References:
[Podestà 2017] M. Podestà et al., Plasma Phys. Control. Fusion 59 (2017) 095008
[Heidbrink 2018] W. W. Heidbrink et al., Nucl. Fusion 58 (2018) 082027
[Podestà 2019] M. Podestà et al., Nucl. Fusion 59 (2019) 106013
[Bardóczi 2019] L. Bardóczi et al., Plasma Phys. Control. Fusion 61 (2019) 055012
[Kim 2019] D. Kim et al., Nucl. Fusion 59 (2019) 086007
[Cheng 1992] C. Z. Cheng, Phys. Rep. 1 (1992) 211
[Gorelenkov 1999] N. N. Gorelenkov et al., Phys. Plasmas 6 (1999) 2802
[Yang 2018] J. Yang et al., Characterization of Tearing Modes in NSTX, 61st APS-DPP Meeting (2019)Speaker: Mario Podesta (Princeton Plasma Physics Laboratory) -
The understanding and control of runaway electrons (RE) is a top priority of the nuclear fusion program because, if not avoided or mitigated, RE can severely damage the plasma facing components of a tokamak. Two key open problems are the generation and the impurity-based mitigation of RE. The first problem requires the computation of the production rate of RE. That is, given a plasma state, determine how many RE are produced during the thermal and current quench phases of a disruption. The second problem requires the simulation of the interaction of RE with partially ionized impurities in the presence of spatio-temporal evolving electric and magnetic fields. The study of these problems is motivated by the practical challenges of finding the optimal impurity injection protocol for the controlled shutdown of a plasma (avoiding or minimizing the generation of RE) and the design of impurity injection strategies for the effective dissipation of the RE beam once it is formed.
The main goal of this paper is to advance the current understanding of RE generation and mitigation by presenting a numerical study focusing on the role played by usually neglected, or highly approximated, spatio-temporal effects. Of particular interest is the dependence of the RE production rate on general dynamic scenarios exhibiting time dependent plasma temperature and electric fields as well as RE beam losses due to radial transport. The study of spatio-temporal effects in the mitigation of RE by impurity injection is also an important goal of this paper. Most of the results will focus on prescribed electric and magnetic fields on a given plasma state. However, preliminary results on self-consistent effects involving coupling of the kinetics of RE to temperature and electric field evolution models will also be presented.
The study of these problems has been a topic of significant interest in the fusion community and many important results have been obtained over the last decade, see for example the recent review paper [1] and references therein. What distinguishes the study presented here from previous work is the focus on spatial effects and the unique approach followed. This approach is based on the combination of a probabilistic Backward-Monte-Carlo method (BMC) [2] and kinetic simulations using the Kinetic Orbit Runaway electron Code (KORC) [3]. The BMC method is based on a direct numerical evaluation of the Feynman-Kac formula that establishes a link between the solution of the adjoint Fokker-Planck problem for the probability of runaway, PRE, and the stochastic differential equations describing the dynamics of RE in the presence of collisions. Computationally, the BMC is a deterministic algorithm that reduces the problem to the evaluation of Gaussian integrals that can be efficiently computed with high accuracy using Gauss-Hermite quadrature rules. Reference [4] discussed how the BMC can be used to efficiently compute the coupling between a fluid and a kinetic description of RE dynamics.
Figure 1 shows an example of a BMC computation of the PRE, incorporating spatial dependence. Going beyond our previous study [2], that limited attention to the computation of the PRE for a given momentum and pitch angle, we extended the BMC to account for radial transport, which in this example is modeled as a diffusive process. The numerical implementation of this 3D extension uses hierarchical sparse-grid interpolation methods with piece-wise polynomials to approximate the map from the phase space to the runaway probability. Also, to handle the sharp transition layer between the runaway and non-runaway regions we use adaptive refinement techniques [5]. Another important extension of the BMC that will be discussed in this paper is the computation of the PRE in time dependent scenarios incorporating models for the temperature and electric fields evolution during the thermal quench.
Figure 2 shows an example of a KORC simulation of a RE beam in time-dependent magnetic and electric fields in the presence of impurities. For these simulations, KORC has been extended by incorporating state-of-the-art collisional models with partially ionized impurities and spatio-temporal models of impurity transport [6]. As indicated by the vertical dashed line, in this simulation, the evolution of the RE beam can be divided in two stages. For t<0.015, as shown in panel (b), there are no RE lost to the wall. However, as shown in panel (a), the energy of the beam actually increases. Interestingly, this increase of the energy is accompanied by a decrease of the parallel (zero pitch angle direction) current due to strong pitch angle scattering or the RE with the impurity. For t>0.015, the fraction of RE lost to the wall significantly increases. This happens because, in this simulation, the time evolution of the magnetic field exhibits a strong horizontal displacement and the RE “peel off” at the high-field side. This loss of confinement leads to a decrease of the RE energy (shown in panel (a)) but this is mostly due to the deconfinement of high energy RE and not because impurity-driven dissipation which, as seen in the green curve in panel (b), is small. The main message of this simulation is that the assessment of the effectiveness of impurity-based RE mitigation is a complex problem involving the competition of different physics mechanisms (magnetic confinement, electric field evolution, coulomb drag, pitch angle scattering, and impurity transport and ionization) with different times scales. In particular, if as in the simulation shown in Fig.2, the time scale of the magnetic field evolution is faster than the time scale of the stopping power of the impurity, then the RE might hit the plasma facing components of the tokamak before they can be significantly slowed down. Simulations exploring this general scenario in the context of DIII-D experiments will be discussed in Ref.[6].
[1] B. Breizman, P. Aleinikov, E. Hollmann and M. Lehen, Nucl. Fusion 59 083001 (2019).
[2] G. Zhang and D. del-Castillo-Negrete, Phys. Plasmas 24, 092511 (2017).
[3] L. Carbajal et al., Phys. Plasmas 24, 042512 (2017).
[4] E. Hirvijoki, C. Liu, G. Zhang, D. del-Castillo-Negrete and D. Brennan, Phys. of Plasmas 25, 062507 (2018).
[5] M. Yang, G. Zhang, D. del-Castillo-Negrete, M. Stoyanov, and M. Beidler, https://arxiv.org/abs/2001.05800 (2020).
[6] M. Beidler et al., 28th IAEA Fusion Energy Conference (FEC 2020).*Research sponsored by the Office of Fusion Energy Sciences of the US Department of Energy at Oak Ridge National Laboratory, managed by UT-Battelle, LLC, for the U.S. Department of Energy under contract DE- AC05-00OR22725.
Speaker: Diego del-Castillo-Negrete (Fusion Energy Division. Oak Ridge National Laboratory) -
This work reports on a breakthrough on the way to a comprehensive modelling of burning fusion plasmas. For the first time, global electromagnetic gyrokinetic PIC simulations of Alfvénic modes have been successfully performed for a high beta ITER plasma.
This finally gives us the ability to model alpha particle driven modes self-consistently in the non-linear regime and to predict the related alpha particle transport with a high level of confidence.
The ITER 15 MA scenario [1], with significant alpha particle pressure, is a scenario in which a broad range of Alfvén eigenmodes may be present. More specifically, a large number of possible of toroidal Alfvén eigenmodes (TAEs) can exist in the plasma, and multiple of these may be driven by the alpha particle gradient. Whilst the induced transport is expected to be small at nominal parameters, it is of significant interest to consider the sensitivity and borders to enhanced transport regimes. Previous modelling of this discharge has been performed (see, for example reference [2] and references therein), most notably by perturbative hybrid (MHD or gyrokinetic eigenvalues and drift- or gyro-kinetic energetic particles), nonperturbative hybrid MHD-kinetic, or local gyrokinetic. Although we know that kinetic, nonperturbative, and global effects are important for the TAEs in ITER, due largely to the difficulties caused by the large size and high plasma β of ITER, it has previously not been possible to apply nonlinear models containing all of these effects consistently. In parallel to this, significant progress [3] has recently been made in global, electromagnetic gyrokinetics, which has yielded e.g. detailed nonlinear studies of energetic particle driven Alfvén eigenmodes in simplified conditions [4]; benchmarks of Alfvén eigenmodes in experimental conditions [5, 6]; and the nonlinear interaction between electromagnetic turbulence and Alfvén modes [7].
The simulations in this work are performed with the ORB5 code [8, 9], a global electromagnetic particle-in-cell (PIC) code using spectrally filtered finite elements for the representation of the fields, with all species considered kinetically, and using the “pullback scheme” [3] to mitigate the cancellation problem.In this contribution, we present the first application of a global nonlinear electromagnetic gyrokinetic code to address the issue of TAE stability in ITER, focussing first on the linear stability of the modes over a range of mode numbers and gap positions, before looking at the saturation observed when retaining the wave-particle nonlinearity. This is expanded by considering also multiple modes present simultaneously, observing a significant increase of the mode saturation levels by approximately one order of magnitude, and accordingly increased alpha particle redistribution, in a case with double the nominal alpha particle density, and when neglecting the gyro-average on the energetic particles. This is consistent with the finding of previous hybrid modelling [10].
In the linear regime, we observe a range of mode numbers for which there exist multiple eigenmodes, with global mode structures spanning multiple gaps, reflecting what is found with linear eigenvalue calculations. We show that these cases require a global domain to describe the linear mode properties. For more localized mode structures, with higher mode numbers, considering a reduced domain is justified, and we see close agreement, even for nonlinear saturation levels between full and reduced radial domains.
Figure 1: A snapshot of the absolute magnitude of the poloidal harmonics of the electrostatic potential of an n=12 TAE in the linear phase.In linear simulations without energetic particles present, where we allow an initial perturbation to decay until eigenmodes are observed, we find also different modes present, for example those from the higher frequency gaps (e.g. EAE and NAE), as well as odd-parity TAEs from the upper part of the gap, as these modes (although barely driven in the presence of energetic particles) are weakly damped. An interesting observation is that in the post-saturation phase of the nonlinear multi-mode TAE simulations, we also observe odd-parity TAEs present. This observation motivates a reconsideration of which modes to include for hybrid perturbative simulations of the same scenario, where previously the odd-parity modes were neglected a priori due to the weak drive.
Figure 2: Time evolution of the peak radial values of the different toroidal mode numbers of the electrostatic potential in a nonlinear simulation.To make simulations feasible, we have allowed several simplifications of the physics problem, but we have considered the impact of each, which we will discuss, in particular, we show the effect of reducing the ion/electron mass ratio.
We also present the application of the ORB5 code to an experimental case taken from an ASDEX Upgrade discharge with particularly interesting energetic particle physics [11], and this result is benchmarked against hybrid codes. The details of this benchmark will be presented in reference [12].
Acknowledgement: This work has been carried out within the framework of the EUROfusion
Consortium and has received funding from the Euratom research and training programme 2014-
2018 and 2019-2020 under grant agreement No 633053. The views and opinions expressed herein
do not necessarily reflect those of the European Commission.References
[1] A. Polevoi et al., JPFR-S, (2002)
[2] S. Pinches et al., Phys. Plasmas (2015)
[3] A. Mishchenko et al., Comp. Phys. Comm. (2019)
[4] M. Cole et al., Phys. Plasmas (2017)
[5] S. Taimourzadeh et al., Nuclear Fusion (2019)
[6] F. Vannini et al., Submitted to Physics of Plasmas
[7] A. Biancalani et al., this meeting
[8] S. Jolliet et al., Comp. Phys. Comm. (2007)
[9] E. Lanti et al., Comp. Phys. Comm. (2020)
[10] M. Schneller et al., Plasma Phys. Control. Fusion (2015)
[11] Ph. Lauber et al, EX1/1 Proc. 27th IAEA FEC (2018)
[12] G. Vlad et al, this meetingSpeaker: Thomas Hayward-Schneider (Max Planck Institute for Plasma Physics) -
Recent experiments on DIII-D have utilized the new off-axis neutral beam injection (NBI) power to achieve $\beta_N$ = 3.8 with n = 1 ideal stability limits up to $\beta_N$ = 6. The NBI upgrade adds two additional co-current, off-axis, beams giving a total of 8 MW of on- and 7 MW of off-axis NBI power for advanced tokamak (AT) scenario development in these experiments. In addition, 1.6 MW of electron cyclotron (EC) power is used as an additional off-axis heating and current drive source. These off-axis current drive sources broaden the current and pressure profiles to better couple to the vessel wall thereby raising the ideal-wall, low-n kink stability $\beta_N$ limits.
Despite higher ideal stability limits with the additional off-axis current drive capabilities, a majority of these high-q$_{min}$ discharges are limited by tearing modes. Past analysis indicates discharges with higher ideal stability limits have higher tearing mode stability limits$^1$. However, these recent experiments, with the additional off-axis beam power, have increased the ideal-wall limit without apparent improvement in tearing mode stability. The DCON stability code$^2$ is used to calculate the ideal-wall and no-wall stability limits. At the time of tearing mode onset, the ideal-wall $\beta$ limits range between $\beta_N$ = 4.2-5.7, no-wall $\beta$ limits between $\beta_N$ = 2.8-3.5, and achieved experimental $\beta_N$ = 2.5-3.8, Fig. 1. In three of the discharges, tearing modes form with an ideal-wall $\beta$ limit of $\beta_N$ > 5 and experimental $\beta_N$ < 3. Clearly, increasing ideal stability limits has not been sufficient for preventing the frequent appearance of tearing modes in these discharges. Furthermore, it is observed that tearing modes frequently form when $\beta_N$ is near the no-wall stability limit. None of the discharges exceed the no-wall $\beta$ limit by more than 10% despite ideal-wall stability limits that exceed the no-wall limits by 50%. With the additional beam power available, these discharges are stability, not power, limited. A better understanding of tearing mode onset physics and avoidance requirements is needed for this regime.
Large tearing modes in these plasmas are m/n = 3/1 and result in a confinement reduction of $\approx$50%. A majority of the tearing modes have a 5/2 tearing mode precursor, which causes a relatively minor reduction in confinement. Tearing modes occur in 10 of 12 discharges shown in Fig. 1 with the tearing mode onset indicated by a circle. After the tearing mode forms, the plasma confinement does not recover in a majority of the discharges.
q$_{min}$ > 2 operations eliminate the 2/1 rational surface from the plasma and avoid deleterious fast-ion modes. However, confinement reduction from 3/1 tearing modes have been significant enough to prevent higher $\beta_N$ operation in this regime. Timing and onset of tearing modes do not show a clear relationship to broader current (higher q$_{min}$) or pressure (lower pressure peaking factor) profiles for operations near q$_{min}$ = 2.
The highest $\beta_N$ with the new off-axis NBI capabilities compared to a similar discharge with all on-axis NBI power was achieved with q$_{min}$ = 1.1-1.5, Fig. 2. This discharge achieved ideal $\beta_N$ stability limits near 6, significantly higher than the reference discharge with only on-axis beam power. Feedback control with 3D fields was applied in both discharges to maintain optimal error field correction and resistive wall mode stabilization. In addition, this result was achieved despite a reduction in available EC power from 2.9 to 1.6 MW.
Predictive TRANSP simulations aided the development of these discharges with increased off-axis NBI power. TRANSP runs of past discharges with majority on-axis beam power were modified to inject power with the new off-axis beam geometry. TGLF was used to evolve the temperature, density, and current profiles with the increased off-axis NBI power. These predictive simulations showed that early application of EC power raises q$_{min}$ and increases the non-inductive current fraction of the plasma, which was observed in subsequent experiments. Broadening of the NBI current density profile with the new beam geometry was also accurately predicted using TGLF in TRANSP.
High fusion gain steady-state tokamaks are based on broad current and pressure profiles to achieve wall-stabilization of ideal-MHD kink modes at high $\beta_N$. The results discussed show that obtaining a high ideal-wall limit, while necessary, is not sufficient, as tearing modes still appear at lower $\beta_N$, usually around the no-wall limit. The relationship between the ideal- and no-wall stability limits and the current density and pressure profiles that determine tearing mode stability will be explored to better sustain higher $\beta_N$ plasmas.
This material is based upon work supported by the U.S. Department of Energy, Office of Science, Office of Fusion Energy Sciences, using the DIII-D National Fusion Facility, a DOE Office of Science user facility, under Awards DE-AC52-07NA27344, DE-FC02-04ER54698, and DE-FG02-04ER54761.
$^1$F. Turco, et. al, Physics of Plasmas 19, 122506 (2012).
$^2$A. H. Glasser, Physics of Plasmas 23, 072505 (2016).Speaker: Brian Victor (Lawrence Livermore National Laboratory) -
In this paper the integrated modelling for the steady-state regime of a fusion neutron source DEMO-FNS (R/a=3.2m/1m, B=5T, Ip=4-5MA) [1] is complemented by the helium balance in the divertor and core plasma. The model describes the power and particle balances consistently in the divertor and core plasmas according to approach [2] and finds the condition to fulfill the global requirements on: neutron source value Sneut >1019/s, maximal peaking heat load is limited by qpk <10MW/m2, keeping up-down symmetry and avoidance of deep detachment, μ<1, condition for plasma current overdrive Ipl ≤ IBS+ICD , when the plasma current Ipl is generated by the noninductive methods from neutral beam ICD and from the bootstrap current IBS and on a source by pellets Spel>fT/(1-fT)SNB, that control the tritium fraction fT =nT/(nT+nD) in a core, where SNB is the D source from neutral beam, Spel is the pellet source of D and T.
The energy and particle balances in the core plasma are modelled with transport equations in the ASTRA code [3] using ITER IPB(y, 2) [4] with prescribed H-factor, ratio of main ions and helium confinement times to the energy confinement time τp/τE, τНе/τE , tritium fraction, fT, electron averaged density <ne> (maintained by the pellet source Spel) and density at the pedestal (see Table 1). Model for the pedestal poloidal beta and the pedestal width is taken from [5,6].
The state of divertor plasma is determined by the values of the incoming heat flux PSOL, the neutral pressure in divertor volume pn, the concentration of the neon impurity CNe and the pumping speed Cpump to pump out the helium ash produced by the fusion reaction (see Table 2). Results of SOLPS4.3 [7] series runs are approximated by analytical dependencies (scalings) in order to install the boundary conditions and heat and particle fluxes through the separatrix into the core plasma.
Self-consistent modeling of central and divertor plasma for DEMO-FNS [8] allows to determine the acceptable window of the divertor parameters pn and CNe where the global requirements are fulfilled (see Figure). It should be noted that in comparison to the modelling without helium balance [1] the acceptable parameter window become significantly constricted. The neutron yield is weakly reduced but remained to be at the reasonable level Sneut= 1.1·1019/s. As one of the neutron reduction factor is the decrease of the tritium fraction in the core plasma in order to fulfill the requirement to tritium fraction control with pellet source (see Table 3). The up-down symmetry requirement (μ<1 red line in Figure) become a strong factor in the divertor neutral pressure pn limit. We need to make an accurate investigation of a single X-point configuration as it can be more efficient to mitigate the power load to divertor plates.
References
[1] A.Yu. Dnestrovskiy et al., Nuclear Fusion 2019 59 (9) 096053
[2] H. D. Pacher, et al., J. Nucl. Mater. 313 (2003) 657
[3] G.V. Pereverzev and P.N. Yushmanov, IPP-Report IPP 5/98 (2002),
[4] ITER Physics Expert Group on Confinement and Transport, 1999 Nucl. Fusion 39 2175
[5] P. B. Snyder et al. Nucl. Fusion, 2011, vol. 51, p. 103016
[6] S.Yu. Medvedev., et al., 2012 Probl. At. Sci. Technol. Ser. Thermonucl. Fusion 35 21 (in Russian)
[7] A.S. Kukushkin, et al., Fusion Eng. Des. 86 (2011) 2954
[8] Yu.S. Shpanskiy and DEMO-FNS Team 2019 Nucl. Fusion 59 076014Table 1 Set of core plasma parameters
τp/τE, τНе/τE 4
Н-factor 1.3
fT 0.5
Ipl, MA 4.5
nped/<ne> 0.7
<ne> 1019/m3 7Table 2 Set of divertor plasma parameters
PSOL, MW 37.5
pn, Pa 2
Cpump, m3/s 20
СNe=ΣNNe/ne 0.025
qpk, MW/m2 9.7
μ 0.87Table 3 Result global values
Sneut 1019/с 1.1
IBS, MA 1.7
ICD, MA 2.8
PDT МW 31
PDT_pp /PDT_bp 1.8
SpelT,1019/s 58
SpelD,1019/s 21Speaker: Alexey Dnestrovskiy (NRC Kurchatov Institute) -
Contrary to previous thinking, recent experiments on DIII-D suggest that the low-frequency instability known as the beta-induced Alfvén-acoustic eigenmode (BAAE)$^1$ will not degrade high-energy fast-ion confinement on future devices. On the other hand, another low-frequency instability, the beta-induced Alfvén eigenmode (BAE)$^2$, interacts strongly with the high-energy fast-ion population and remains a threat.
Deuterium neutral beam injection (NBI) drives BAAE, BAE, and reversed-shear Alfvén eigenmode (RSAE)$^3$ instabilities (Fig. 1a). When the NBI and electron cyclotron heating (ECH) patterns are held constant, the AE activity (and q profile evolution) is highly reproducible at the time of interest. Transient changes in heating isolate the driving gradients for the instabilities. When the NBI heating is altered, the BAE activity rapidly changes (Fig. 1b) in a time much shorter than the ~100 ms slowing-down time, indicating that high-energy ($E_{beam}\sim80$ keV) beam ions drive the BAEs unstable. In contrast, BAAE activity persists even in the absence of NBI. Even though the neutron rate decays to <5% of its previous value, the BAAEs remain unstable, indicating that fast ions with energies above ~30 keV are not driving the BAAEs. In other discharges, the angle of beam injection was changed from tangential to perpendicular at approximately constant NBI power. The frequency of BAE activity increased but BAAE activity was hardly affected. Calculations of the resonance condition $\omega=n\omega_\phi+p\omega_\theta$, where $\omega_\phi$ and $\omega_\theta$ are the toroidal and poloidal orbital frequencies of recently injected beam ions, show that the increase in frequency $\omega$ allows perpendicular beam ions to resonate with the BAE. (Here, n is the toroidal mode number and p is an integer.)
The intermittent appearance of RSAE, BAE, and BAAE activity in Fig. 1 occurs because all three modes are sensitive to the q profile, which continuously decreases during the illustrated time. Electron cyclotron emission (ECE) and beam emission spectroscopy (BES) diagnostics show that the radial eigenfunction for all three modes peaks near $q_{min}$. Unlike RSAEs and BAEs, the BAAEs are undetectable on Mirnov coils, suggesting less electromagnetic polarization for these modes.
In addition to a database of the 20 shots in the dedicated experiment of Fig. 1, a database of over 1000 DIII-D discharges with ~ 2500 entries has been compiled. For each entry, RSAEs, BAEs, and BAAEs (as well as TAEs and EAEs) are classified as stable, marginal, or unstable. In both databases, BAAE activity occurs most often in plasmas with large electron temperature $T_e$ but relatively modest beta (Fig. 2a). Since the BAAE is expected to have an acoustic component to its polarization, these dependencies may reflect an underlying dependence on ion Landau damping, which is enhanced at low $T_e$ and in plasmas with large ion pressure. As expected from the dedicated experiment, the dependence on NBI parameters in both databases is very weak for the BAAE.
In contrast, BAEs occur more frequently as the poloidal beta $\beta_p$ increases (Fig. 2b) and in plasmas with more beam power and parallel beam beta. For $\beta_p>1.5$, the occurrence of BAEs is comparable to the occurrence of TAEs, highlighting the importance of predicting BAE stability in Advanced Tokamak reactor scenarios.
Initial comparisons of gyrokinetic (GTC and LIGKA) and gyrofluid (FAR3D) codes against the data of Fig. 1 are encouraging. GTC finds a linearly unstable mode with frequency and mode structure similar to the experimental BAE; an unstable mode with BAAE frequency is only found in simulations without beam ions. Preliminary nonlinear simulations indicate that zonal flows cause the saturation of the unstable BAE. After benchmarking, these codes will predict BAAE and BAE stability in ITER.
Work supported by US DOE under DE-FC02-04ER54698 and DE-SC0020337.
$^1$N.N. Gorelenkov et al., Phys. Lett. A 370 (2007) 70.
$^2$W.W. Heidbrink et al., Phys. Rev. Lett. 71 (1993) 855.
$^3$S.E. Sharapov et al., Phys. Lett. A 289 (2001) 127.
Speaker: william heidbrink (UC Irvine) -
Unstable Alfvén eigenmodes (AEs) are a key issue in magnetically confined fusion, both for currently operating machines (JET, AUG, etc) and for next step devices such as JT60-SA and ITER, due to their potential to cause energetic-ion (heated by NBI/ICRH or fusion alphas) redistribution and losses [1,2]. Toroidicity-induced AEs (TAEs), resulting from the coupling of two shear-Alfvén waves, are one of the most extensively studied Alfvénic instabilities in tokamaks [1,2]. However, lower-frequency AEs (i.e., $\omega<\omega_{tae}$) have been observed in DIII-D high-beta plasmas under intense NBI, associated with large energetic-ion loss levels, much similar to those caused by TAEs [3,4].
In this contribution, we report and discuss Alfvénic activity observed at JET, at about half the TAE frequency (figure 1), in plasmas heated by NBI and ICRH [5]. AEs with frequencies lower than the TAEs have been previously explained by the beta-induced coupling of shear-Alfvén and acoustic waves (BAAEs), via the lowest-order harmonic of the field-line geodesic curvature [6,7]. However, their expected frequency $\omega_{baae} \sim 2 \beta^{1/2} \omega_{tae}$ is too low to explain the measured eigenfrequencies on JET. Here, we show that the experimental measurements can be explained by the previously unexplored gaps in the frequency of shear-Alfvén and acoustic continua. In the vicinity of these gaps, several acoustic waves couple to a single shear-Alfvén wave via higher-order harmonics of the geodesic curvature caused by the plasma shaping and, in particular, the elongation of JET plasmas [8]. In the limit of plasmas with circular magnetic surfaces, the proposed model reduces to the well-known lowest-order coupling. New frequency gaps are predicted at integer multiples of $\omega_{baae}$, but only under certain conditions imposed by the local shaping parameters and a limiting value of the safety factor [8].
The continuous spectra computed with the compressible ideal-MHD code CASTOR display the predicted frequency gaps, which open where acoustic branches cross with the up-shifted shear-Alfvén branch (figure 2). Although quite narrow, such gaps allow the existence of global AEs whose radial structure and frequency is also computed with CASTOR (figure 2). In this contribution, the computed AE radial location and frequency are shown to be in fair agreement with experimental data, the latter within a few percent of measured values and the former matching soft x-ray observations.
The proposed high-order geodesic acoustic eigenmodes (HOGAEs) were found to be driven unstable by energetic-ion populations with characteristics similar to those accelerated by ICRH (figure 3), with a temperature around 1 MeV and on-axis density about 1% of the value corresponding to the thermal ions [8]. In this contribution, their stability in JET plasmas will be evaluated with the hybrid MHD/drift-kinetic code CASTOR-K, taking into account energetic-ion populations heated by NBI and ICRH, their distributions functions being computed, respectively, by the codes ASCOT and PION. Wave-particle resonances, along with drive/damping mechanisms, will also be discussed. Amplitude saturation levels and mechanisms will be addressed with the non-linear code POLARIS-K. Overall, the presented results will allow a better understanding of HOGAEs role in energetic-ion redistribution and losses and their potential impact on the operation of next step devices like JT60-SA and ITER.
Acknowledgments:
This work has been carried out within the framework of the EUROfusion Consortium and has received funding from the Euratom research and training programme 2014- 2018 and 2019-2020 under grant agreement No 633053. One of the authors (FC) was supported by FuseNet from the Euratom research and training programme under Grant Agreement No. 633053. The views and opinions expressed herein do not necessarily reflect those of the European Commission. IPFN activities were also supported by “Fundação para a Ciência e Tecnologia” (FCT) via project UID/FIS/50010/2013.References:
1 A. Fasoli et al., Nucl. Fusion 47, S264 (2007).
2 N. Gorelenkov et al, Nucl. Fusion 54, 125001 (2014).
3 W. Heidbrink et al, Phys. Rev. Lett. 71, 855 (1993).
[4] W. Heidbrink et al, Phys. Plasmas 6, 1147 (1999).
[5] R. Dumont et al, Nucl. Fusion 58, 082005 (2018).
[6] B. Holst et al, Phys. Plasmas 7, 4208 (2000).
[7] N. Gorelenkov et al, Phys. Lett. A 370, 70 (2007).
[8] F. Cella and P. Rodrigues, “Shaping effects on the interaction of shear-Alfvén and slow sonic continua”, Proc. 46th EPS Conference on Plasma Physics 8-12 July 2019.Speaker: Paulo Rodrigues (Instituto de Plasmas e Fusao Nuclear) -
Energetic particle physics is a crucial issue in burning plasmas such as the International Thermonuclear Experimental Reactor (ITER). Instabilities driven by energetic particles, such as fishbones and various Alfvén eigenmodes, can induce the transport and loss of energetic particles, degrade fast particle confinement, and even lead to serious wall damage. A non-monotonic safety factor profile with a reversed magnetic shear configuration has been proposed as an advanced scenario for future ITER operation. For the consideration of the fishbone instability, there are two different conditions: the minimum value of safety factor qmin is less or a little larger than unity. There are few simulations to investigate the fishbone instabilities with reversed safety factor profile. As a result, in this work, linear stability and nonlinear dynamics of the fishbone instabilities with reversed safety factor profile have been investigated by the hybrid code M3D-K[1,2], including both the non-resonant type with qmin larger than unity and the type with dual q = 1 surfaces, which we will infer as non-resonant fishbone (NRF) and dual resonant fishbone (DRF) in the following.
Based on EAST-like parameters, the linear simulation results of the n = 1 mode with double q = 1 rational surfaces are firstly presented. With fixed total pressure, the linear growth rate and mode frequency as a function of beam ion pressure fraction Phot,0/Ptotal,0 are shown in figure 1, where Phot,0 is the central fast ion pressure, and Ptotal,0 is the central total pressure. At zero beam ion pressure with Phot,0/Ptotal,0 = 0, the ideal internal kink mode is unstable. The corresponding mode structure is shown in figure 1 (a). This mode has an up-down symmetric structure with zero mode frequency, and it exhibits splitting feature due to double q = 1 surfaces. The dominant mode number is n = m = 1, where n is the toroidal mode number and m is the poloidal mode number. When the value of Phot,0/Ptotal,0 is small and increases from 0 to 0.4, the mode is stabilized due to the kinetic effects of beam ions. However, when Phot,0/Ptotal,0 is larger than 0.4, the DRF is excited, which is an energetic particle mode driven by trapped beam ions. Figure 1 (b) shows the mode structure with Phot,0/Ptotal,0 = 0.3. Compared to the ideal internal kink mode, the mode structure shows a twisted feature with finite mode frequency. The mode structure of the DRF with Phot,0/Ptotal,0 = 0.6 is shown in figure 1 (c), and it becomes more twisted with much larger frequency. When qmin increases from below unity to above unity, the fishbone instability transits from the DRF to the NRF, and the mode frequency of the NRF is higher than the DRF as the NRF is resonant with fast ions with larger precessional frequency. The mode structure of the NRF is shown in figure 1 (d).
Nonlinear simulations show that the saturation of the DRF with Phot,0/Ptotal,0 = 0.6 is due to MHD nonlinearity with a large n = 0 component. Figure 2 shows the reason why the DRF mode cannot saturate just with the nonlinearity of energetic particles. Without MHD nonlinearity, as shown in figure 2 (a), it is found that the distribution of fast ions becomes flattened in the core region during the nonlinear phase. However, near the magnetic axis there still exists the steep radial gradient of the fast ion distribution, which could drive the instability. Correspondingly as shown in figure 2 (c), the inner m/n = 1/1 DRF mode structure shrinks in the central region at t = 700 τA, where τA is the Alfvén time. As a result, the DRF mode still tap the free energy associated with the fast ion radial gradient during the nonlinear phase, and the DRF mode does not saturate without MHD nonlinearity. However, the saturation of the NRF is mainly due to the nonlinearity of fast ions. Figure 3 (a) shows the time evolution of the n = 1 mode amplitude of the NRF. It is observed that the NRF amplitude firstly grows and then saturates from t ≈ 1000 τA. Correspondingly as shown in figure 3 (b), the NRF frequency starts to chirp down at t ≈ 1500 τA from ω ≈ 0.12 ωA to ω ≈ 0.04 ωA, where ωA is the Alfvén frequency. Finally, the redistribution of beam ions due to the DRF and NRF with MHD nonlinearity is discussed, and it is found that the distribution of the fast ions become flattened in the core region. By comparing the fast ion redistribution induced by the DRF, the redistribution level of the fast ions due to the NRF is weaker, and the flattening region of the beam ions is located more centrally in the radial direction, which is consistent with the resonant analysis. Because the equilibrium profiles and parameters are chosen based on EAST-like conditions, the simulation results will provide guidance for and can be verified by future EAST experiments.
References
1 W. Park, E. V. Belova, G. Y. Fu, X. Z. Tang, H. R. Strauss, and L. E. Sugiyama, Phys. Plasmas 6, 1796 (1999).
2 G. Y. Fu, W. Park, H. R. Strauss, J. Breslau, J. Chen, S. Jardin, and L. E. Sugiyama, Phys. Plasmas 13, 052517 (2006).Numerical computations were performed on the ShenMa High Performance Computing Cluster in Institute of Plasma Physics, Chinese Academy of Sciences. This work is supported by National Key R&D Program of China under Grant No. 2017YFE0300400, the National Natural Science Foundation of China under Grant Nos. 11605245, 11975270, 11705236, 11575249, 11805239 and 11505022.
Speaker: Wei Shen (Institute of Plasma Physics, Chinese Academy of Sciences) -
Hybrid simulations for energetic particles interacting with a magnetohydrodynamic (MHD) fluid were conducted using the MEGA code [A, B] to investigate the spatial and the velocity distributions of lost fast ions due to the Alfvén eigenmode (AE) bursts in the Large Helical Device (LHD) [C, D]. It is found that the spatial distribution of lost fast ions in the divertor region during the AE burst is helically symmetric and peaks along the divertor location. Affected by the direction of grad-B and curvature drifts, the distribution of the co-going (counter-going) lost fast ions has two peaks in the outboard (inboard) side depending on fast-ion energy and pitch angle. The numerical fast-ion loss detector “numerical FILD” was constructed in the MEGA code. The velocity distribution of lost fast ions detected by the numerical FILD during AE burst is in good agreement with the experimental FILD measurements. This demonstrates that the MEGA is a useful tool for the prediction and the understanding of the fast-ion transport and losses brought about by AEs.
The LHD is one of the largest helical devices with non-axisymmetric 3-dimensional magnetic configuration. In the LHD, the fast-ion confinement has been investigated by using three tangentially injected neutral beams (NBs) with energy 180 keV and/or two perpendicularly injected NBs with energy 40-80 keV. The recurrent AE bursts were observed during the tangentially injected NB [C, D]. The fast-ion driven instabilities enhance the fast-ion transport and losses. It is important to identify the instabilities and clarify the properties of the fast-ion transport due to the instabilities. A hybrid simulation code for nonlinear MHD and energetic-particle dynamics, MEGA, has been developed to simulate recurrent bursts of fast-ion driven instabilities including the energetic-particle source, collisions and losses [A]. Since the equilibrium magnetic field in the real coordinates are used in MEGA, fast ion can be traced even in the peripheral region including the divertor region. The multi-phase simulation, which is a combination of classical simulation and hybrid simulation for energetic particles interacting with an MHD fluid, was applied to the LHD experiments #47645 [C] and #90090 [D] in order to investigate the AE bursts with beam injection, collisions, losses, and transport due to the AEs [A, B]. In the classical simulation, fast-ion orbits are followed in the equilibrium magnetic field with NBs and collisions while the MHD perturbations are turned off. The fast-ion loss rate brought about by the AE burst is proportional to the square of AE amplitude, which is consistent with the quadratic dependence of fast-ion loss observed in the LHD experiment [B].
In this work, the spatial and the velocity distributions of lost fast ions due to the AE bursts are investigated and compared with the experimental measurements. The multi-phase simulation was conducted for the LHD experiment #90090 where the 2-dimensional velocity distribution of lost fast ions is measured by scintillator-based FILD [D]. The AE bursts occur recurrently and then the fast ions are significantly lost during the AE bursts. Figure 1 shows the spatial distribution of lost fast ions in the divertor region during the AE burst. We see in Fig. 1 that the spatial distribution of lost fast ions in the divertor region during the AE burst is helically symmetric and peaks along the divertor trace. The lost fast ions reach the divertor region following the divertor magnetic field. There is a helical symmetry for the lost fast ion location even during the AE burst while some peaks are present in the number of lost fast ions in the poloidal direction. The time evolution of the spatial distributions of lost fast ions along the divertor trace is shown in Fig. 2. Affected by the direction of grad-B and curvature drifts, the distribution of the co-going (counter-going) lost fast ions has two peaks in the outboard (inboard) side depending on fast-ion energy and pitch angle. For the comparison with the lost fast-ion velocity distribution measured with the FILD, we have constructed the “numerical FILD” in the MEGA code. In the MEGA simulations, the guiding-center orbit is followed for fast ions. In the numerical FILD, when a fast ion approaches the FILD, we split the guiding-center particle into 64 particles around the guiding center with corresponding Larmor radius and 64 particles with different gyration phase are followed with Newton-Lorentz equation. The aperture of the numerical FILD is a circle with radius 6 mm. Only the fast ions passing through the aperture are detected by the numerical FILD. Figure 3 compares the pitch angle and energy distribution of the lost fast ions in the MEGA simulation and the FILD measurements in the experiment. Before the AE burst, fast ions with energy close to the injection energy are mainly detected by the numerical FILD. During the AE burst, we see in Fig. 3(b) that fast ions of 100-150 keV and 35-50 degree are detected by the numerical FILD. The velocity space region of the lost fast ions due to the AE burst is in good agreement with that observed in the experiment shown in Fig. 3(c) [D], although the two peaks observed in the experiment are not well resolved in the numerical FILD. The numerical FILD measurement is consistent with the experiment for the lost fast ions with pith angle = 30-40 degree which increased during the AE burst.
![Spatial distribution of lost fast ions in the divertor region during the AE burst. Color represents the number of lost fast ions [arb. unit]. The promptly lost particles are not included. The star indicates the installation position of FILD in the LHD.](http://workshop.nifs.ac.jp/fec2020/image/71-SEKI-image-fig_1.jpg)
![Comparison of pitch angle and energy distribution of lost fast ions among (a) MEGA simulation before AE burst, (b) simulation during AE burst, and (c) FILD measurements in the LHD experiment [D].](http://workshop.nifs.ac.jp/fec2020/image/71-SEKI-image-fig_3.jpg)
[A] Y. Todo, et al., Phys. Plasmas 24, 081203 (2017).
[B] R. Seki et al., Nucl. Fusion 59, 096018(2019).
[C] M. Osakabe et al., Nucl. Fusion 46 S911 (2006).
[D] K. Ogawa et al., Nucl. Fusion 52, 094013 (2012).
Speaker: RYOSUKE SEKI (National Institute for Fusion Science) -
Experiments on DIII-D and C-Mod show that high neutral opacity is compatible with a steep density gradient at the plasma edge [1,2]. Future reactors, including ITER, will operate at high neutral opacity, which will strongly limit direct fueling of the pedestal structure inside the Last Closed Flux Surface (LCFS) through ionization from edge sources in comparison with current existing experimental devices. At the highest opacities, the electron density pedestal structure is stiff and is not degraded by increases in fueling and Scrape-Off Layer (SOL) density. These experiments directly question integrated modeling predictions for ITER that show high opacity is incompatible with a steep pedestal density gradient inside the LCFS when particle transport is considered purely diffusive [3].
High neutral opacity means that neutrals will ionize predominately inside the SOL, limiting the impact of edge fueling and peaked density profiles, under the assumption of purely diffusive particle transport. To test whether limited edge fueling in high neutral opacity regimes results in the collapse of the pedestal density profile, as predicted by simulations for ITER [3], we conduct a set of experiments on DIII-D and C-Mod in which we increase the neutral opacity. In our experiment we obtained opacity values that are only a factor 2 lower than those on ITER, while current day machines operate typically at opacity values that are about a factor 10 lower than those expected on ITER. Opacity can be approximated using values of the electron density and the minor machine radius, $n \times a$. The experiments on DIII-D and C-Mod ranging in opacity range from $1.5 \times 10^{19} \; m^{-2}$ to $5.5\times 10^{19} \; m^{-2}$ by operating at various plasma currents as well as adding an additional gas puff. Within this span of opacities, both DIII-D and C-Mod have robust density pedestal structures and increases in fueling and opacity did not affect the maximum density gradient on either machine. The max($1/L_{ne}$) on DIII-D remained close $\sim 8 \times 10^{21} \; m^{-4}$ in absolute value for a fixed IP when the opacity was raised using gas fueling. This is also reflected in the fact that $ne_{ped}/ne_{sep}$ remains constant for DIII-D as well as C-Mod over a wide range of opacities, see figure 1b). Thus, suggesting that changes in neutral penetration has little effect upon the pedestal density structure.
Using the filterscope measurements of $D_{\alpha}$ light at the mid plane inside the LCFS as well as in the SOL, we observe that with increasing opacity, there is a decrease in the neutral density especially inside the pedestal structure on DIII-D. Even though direct fueling through ionization of edge neutrals sharply decreases in the pedestal structure, the pedestal density still increases with increasing opacity, see figure 1a). This agrees with SOLPS-ITER modeling of these DIII-D discharges, which also shows a decrease in neutral density, not just at the midplane, when the neutral opacity increases in these experiments.
While opacity does not exclude a robust pedestal density structure, it does reduce our ability to fuel the core plasma using gas puffing. With increasing opacity, the increase in the pedestal density as a function of the additional gas fueling becomes smaller. This observation is again counter to the integrated predictive ITER H-mode simulations, which indicated that there was no saturation observed in the pedestal or separatrix density with increasing gas fueling. These models rely on fixed transport coefficients in the SOL and use a Bohm/GyroBohm approximation for the transport on closed flux surfaces. The increase in fueling in these simulations increases the SOL and separatrix density and the choice of the transport model for the core results in this increase propagating up to the pedestal density.
In DIII-D at the highest fueling levels, when the outer strike point detaches, we observe a saturation in the SOL density increase due to the formation of a density shoulder. A density shoulder is associated with a strong change in radial transport, which cannot be captured by these integrated models or SOLPS-ITER self-consistently [4]. The strong increase in radial outward transport associated with the formation of a density shoulder effectively impacts the ability to increase the separatrix density. As such, fueling an opaque plasma using a gas puff will be very inefficient or even completely ineffective. It does however not result in the collapse of the steep pedestal density.
To complicate transport models even further, for the first time, an up-down asymmetry in the electron density inside the LCFS was observed in DIII-D H-mode plasmas. The electron density was higher on the LFS close to the X-point, when compared to measurements taken on the LFS close to the crown of the plasma inside the LCFS. This asymmetry has been predicted by neoclassical models [5]. These results are different from prior research where a high electron density region in the HFS SOL region close to the X-point is observed in AUG H-mode plasmas [6].
This work was supported in part by the US Department of Energy under DE-FC02-04ER54698, DE-AC02-09CH11466, DE-SC0014264, DE-SC0019302, DE-SC0007880
[1] S. Mordijck et al. Nucl. Fus. (2020) submitted to Nuclear Fusion
[2] J.W. Hughes et al. Nucl. Fus. 58 (2018) 112003
[3] M. Romanelli et al. Nucl. Fus. 55 (2015) 093008
[4] D. Carralero et al. Nucl. Fus. 57 (2017) 056044
[5] R.M. Churchill et al. Nucl. Fus. 53 (2013) 122002.
[6] F. Reimold et al. NME 12 (2017)193–199
This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights. Reference herein to any specific commercial product, process, or service by trade name, trademark, manufacturer, or otherwise, does not necessarily constitute or imply its endorsement, recommendation, or favoring by the United States Government or any agency thereof. The views and opinions of authors expressed herein do not necessarily state or reflect those of the United States Government or any agency thereof.Speaker: Saskia Mordijck (College of William and Mary) -
Improving Energetic Particle Confinement in Stellarator Reactors
A. Bader$^1$, M. Drevlak$^2$, D.T. Anderson$^1$, C.C. Hegna$^1$, S.A. Henneberg$^2$,
T.G. Kruger$^1$, A. Ware$^3$
1: University of Wisconsin-Madison, WI, USA,
2: Max-Planck Institut fur Plasmaphysik, Greifswald, Germany,
3: University of Montana, MT, USAEnergetic particle confinement is a key issue for the scalability of stellarators to fusion power plants. Prompt losses of alpha particles born from fusion reactions can cause significant material damage. Analytically derived proxies for collisionless energetic particle confinement (1) have been used for the first time to optimize quasihelically symmetric stellarator equilibria (2). This paper will expand on recently published results, along with inclusion of analysis to account for collisional alpha particle transport with reactor relevant alpha sourcing profiles.
A proxy for energetic particle transport, $\gamma_c$, accounts for both the net bounce-averaged radial particle drifts, a quantity to be minimized, and poloidal drift, a quantity to be maximized. Minimization of 𝛾c corresponds to aligning contours of the second adiabatic invariant, $J_\parallel$, to flux surfaces. This metric has been included in the ROSE optimization code (3) and used to optimize equilibria for good energetic and neoclassical particle transport. Previous results indicate that two classes of stellarators, quasihelically-symmetric stellarators, and maximum-J stellarators should have better energetic particle transport than other configurations (4). This paper focuses on optimizations of quasihelically symmetric stellarators. Results indicate the existence of equilibria which nearly eliminate all collisionless losses within the plasma mid-radius (figure 1a) for an ARIES-CS scale reactor (450 m3, 5.8 T). These configurations are obtained by optimizing simultaneously for $\gamma_c$ and a metric for quasihelical symmetry. Interestingly, configurations with improved energetic particle transport did not correlate with improvements to neoclassical transport in the 1/𝜈 regime as measured by $\epsilon_\mathrm{eff}$.
It is well known that the ripple associated with realistic plasma coils can negatively affect alpha particle confinement. However, using coil optimization codes REGCOIL (5) and FOCUS (6), we show that it is possible to reproduce configurations with high enough fidelity that the alpha particle confinement is not significantly degraded (figure 1b). A key feature for the coil optimization is the realization of the equilibria with coils placed farther from the plasma, thus reducing high order harmonics associated with the toroidal mode numbers equivalent to the coil number. These high order modes have been previously found to be deleterious to energetic particle confinement on both stellarators and tokamaks.
New results will be presented regarding alpha particle transport that include collisions with the background plasma. Collisional calculations require additional assumptions about configuration parameters, most importantly the density and temperature profiles which govern the collisional equations and the alpha particle source distribution. We show that when including collisions, configurations exist at the ARIES-CS scale and with ARIES-CS parameters with total energy loss below 4% (figure 2a) with most of the losses occurring from particles born outside the midradius (> r/a = 0.55) (figure 2b). Results will be presented that show energetic particle transport under a variety of density and temperature profile assumptions.
In light of these results, the outlook for energetic particle optimization for quasihelical stellarators is bright. The results presented here represent only a first attempt and with improved optimization algorithms better configurations may yet be found. Additionally, further improvements in stellarator coil design can help gain confidence that such configurations are realizable.
References:
(1) V. V. Nemov et al. Physics of Plasmas 15 052501 (2008)
(2) A. Bader et al. Journal of Plasma Physics 85 5 (2019)
(3) M. Drevlak et al. Nuclear Fusion 59 016010 (2018)
(4) W. Lotz et al. Plasma Phys. Control. Fusion 34 1037 (1992)
(5) M. Landreman Nuclear Fusion 57 046003 (2017)
(6) C. Zhu et al. Nuclear Fusion 58 016008 (2018)Work supported by DE-FG02-93ER54222, DE-FG02-99ER54546 and UW2020 135AAD3116
Speaker: Aaron Bader (University of Wisconsin-Madison) -
New DIII-D results may explain why achieving ELM suppression by resonant magnetic fields (RMPs) remains elusive in double null (DN) diverted configurations: the lack of ELM suppression in DN correlates with a damped high-field side response of field-aligned structures that could be indicative of a missing resonant tearing needed to stop inward growth of pedestal. This is found despite favorable conditions for RMP suppression in lower single null (LSN): low $\Omega_{E\times{B}}$ aligned with a resonant surface at the pedestal top at low $n_{e,ped}$. The DN configuration is advantageous for future machine design as it allows improved divertor power handling and particle control, but still needs ELM handling solutions and may not be compatible traditional RMP ELM suppression driven from low-field side coils.
In experiments where the magnetic balance is varied from LSN toward DN, ELM suppression was obtained for $dR_{sep}<-1.7$ cm, where $dR_{sep}$ is defined as the separation between the separatrices from the lower null and upper null at the outboard midplane. In discrete steps of $dR_{sep}$, $q_{95}$ is scanned to find a window in ELM suppression under the model that aligning a resonant surface in a region of low $\Omega_{E\times{B}}$ results in resonant tearing inhibiting the inward growth of the pedestal otherwise leading to an ELM {1}. Results of these scans are shown in Figure 1a where the values of $\Omega_{E\times{B}}$ at resonant surfaces are within $\pm 2\%$ of $\psi_{ped,top}$ (to account for uncertainty in profile fitting). Figure 1b shows $n_{e,ped}$ where ELM suppression was achieved (with the largest value of $n_{e,ped}\sim2.4×10^{19} m^{-3}$). This shows for each ELM-suppressed discharge, a resonant surface is near the pedestal top with $\left|\Omega_{E\times B}\right|<20$ krad/s.
In near balanced DN ($dR_{sep}\sim-0.1$ cm), similar scans show that nominal ELM suppression conditions are demonstrated while still ELMing. ELM suppression is not achieved over a range of $q_{95}$ from 3.4 to 4.1 where it was observed in LSN. This is shown in Figure 1c where $\Omega_{E\times{B}}$ at resonant surfaces aligned within $\pm 2\%$ pedestal top are $\left|\Omega_{E\times B}\right|<10$ krad/s—a tighter range than in LSN. These discharges also achieve a lower value of $n_{e,ped}$ than the highest value suppressed in LSN. The pedestal temperature width is consistently wider in ELMing DN plasmas compared to the ELM suppressed LSN plasmas. This leads to a wider total pedestal pressure, consistent with lacking a mechanism inhibiting the pedestal inward growth.
The 3D plasma response to applied RMPs measured on the high-field side (HFS) drops in plasma shapes transitioning from LSN to DN and recovers in upper single null (USN) as shown in Figure 2. The plasma response on the low field side (LFS) remains relatively constant during the shape transition. The reduced HFS response is found similarly for $n=2,3$ over a range of $q_{95}$ from 3.4 to 5. This feature is found broadly across a range of $\left|dR_{sep}\lt0.1\right|$ cm indicating it is not restricted to exactly balanced DN or specific pedestal conditions.
Linearized single-fluid resistive MHD modeling with M3D-C1 shows relatively good agreement with plasma response measurements transitioning from LSN to DN for both the HFS and LFS indicative of a strong damped of perturbations on the HFS in DN. This is further illustrated in the modeled $T_e$ perturbations in Figure 2 where the perturbations are strongly damping on the HFS in DN. This can be partially understood using a simple geometric model assuming field-aligned resonant perturbations. Field-aligned modes driven from the LFS (as is the case with I-coils in DIII-D) are connected to the HFS through a region of low poloidal field in the presence of a secondary null. This can lead to interference of radially-separated resonant modes on the HFS. In balanced double null, this leads to the strongest interference on the HFS.
Results presented here are consistent with using HFS response as a proxy for local tearing drive responsible for ELM suppression by stopping inward growth of pedestal. This is consistent with previous results showing correlation of HFS response and tearing drive needed for ELM suppression {2}. The benefit of DN in power handling resides in a narrow region of $dR_{sep}$ where exact splitting of the heat flux depends on cross-field drifts and has been shown to balance at $dR_{sep}\sim0.25$ cm (near double null) with the ion $\nabla B$ drift directed to the lower divertor {3}. This region of $dR_{sep}$ lies within the damped HFS response and lack of demonstrable ELM suppression. If this model is correct, we can use it to optimize shape and coil positions to better attempt ELM suppression.
This work was supported in part by the US Department of Energy under DE-AC05-00OR22725, DE-FC02-04ER54698, and DE-AC02-09CH11466.
{1} Snyder P.B. et al 2012 Phys. Plasmas 19 056115; Nazikian R. et al 2015 Phys. Rev. Lett. 114 105002; Paz-Soldan C. et al 2019 Nucl. Fusion 59 056012.
{2} Paz-Soldan C. et al 2015 Phys. Rev. Lett. 114 10500
{3} Leonard A.W. et al 2005 Fusion Sci. Technol. 48 1083Speaker: Morgan Shafer (Oak Ridge National Laboratory) -
Improved understanding of the mechanisms that govern thermal transport in the pedestal region is crucial for determining the fundamental processes behind the L-H transition and pedestal structure, and providing a foundation for predicting and optimizing the pedestal and performance of future devices such as ITER. We report world first inferred ion and electron heat fluxes in the pedestal region of deuterium plasmas using direct measurements of the main-ion temperature for power balance across ion collisionalities between 0.1 and 1.2. The ion heat flux (Qi) in the narrow H-mode pedestal, evaluated in the last 10% of the ELM cycle on DIII-D H-modes is compared with neoclassical transport simulations from the NEO code (1), showing agreement within a factor of 2 at higher collisionality (v*~1.2), but increasing anomalous transport as the collisionality is decreased (Fig 1a). This discrepancy suggests that turbulent transport becomes increasingly important for ion thermal transport in the pedestal at lower collisionalities.
Accurate assessments of the ion heat flux on several devices has historically been impeded by anomalies in the impurity charge exchange recombination spectroscopy (CER) temperature measurements near the plasma edge and relying on the assumption that the main-ion temperature is equal to the impurity based measurements. This can lead to a large ion-electron collisional exchange term that produce physically suspicious negative ion heat fluxes near the plasma edge, with a corresponding attribution of excessive transport through the electron channel. The development of main-ion CER (MICER) in the pedestal region on DIII-D (2,3) has allowed direct measurements of the D+ temperature which is free from these anomalies and improves the interpretation of both ion and electron transport inferred from power balance calculations. The main-ion measurements show significant differences compared with the impurities in the H-mode transport barrier. In particular, there can be large (>2x) differences between the main-ion and impurity temperatures measured in the steep gradient region. These differences are contrary to expectations based on equilibration time. Many of the discrepancies at the pedestal top are resolved by including Zeeman and fine structure broadening of the impurity measurement. In the steep gradient region increasingly accurate calculations of the ion motion may be needed to understand the charge exchange signals, this is being accessed using the SPIRAL code. Use of direct main-ion measurements for the temperature profile in the steep gradient region can have large impacts on the ion heat flux when assuming the main-ion temperature is the same as the impurities (Fig 2a) as opposed to using the direct measurements (Fig 2b). Additionally, the use of the steeper main-ion temperature gradient has the potential to impact the stability of microturbulence calculated in gyrokinetic simulations.
A dedicated experiment was run to acquire these new measurements across a range of collisionalities, providing the inputs to interpretive TRANSP calculations of the ion (and electron) heat flux using conditionally averaged profiles from 90-100% of the ELM cycle. The inferred ion thermal transport in the steep gradient region of the H-mode pedestal rises above the neoclassical level as the collisionality is decreased (Fig 1a) suggesting that anomalous ion thermal transport becomes increasingly important at low collisionality conditions relevant to the range expected in ITER. BES measurements show increasing ion scale fluctuations in the pedestal region as the collisionality is reduced (see Fig 1b) suggesting increased ion scale turbulence may be responsible for the anomalous ion thermal transport.This material is based upon work supported by the U.S. Department of Energy, Office of Science, Office of Fusion Energy Sciences, using the DIII-D National Fusion Facility, a DOE Office of Science user facility, under Awards DE-FC02-04ER54698, DE-AC02-09CH11466, DE-SC0019352, DE-FG02-08ER54984,DE-FG02-08ER54999, DE-AC05-00OR22725, DE-FG02-07ER54917, DE-SC0020337.
(1) E. A. Belli, et. al., Plasma Phys. Control. Fusion, 50, 095010 (2008)
(2) B. Grierson, et. al., Rev. Sci. Instr. 83, 10D529 (2012)
(3) S. R. Haskey, et. al., Rev. Sci. Instr. 87, 11E553 (2016)Speaker: Shaun Haskey (Princeton Plasma Physics Laboratory) -
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 and project values for ITER that the construction tolerances and correction coils satisfy. This paper presents a multi-variable n=2 threshold regression across a wide range of densities, toroidal fields, and pressures in 3 machines (DIII-D, EAST, and COMPASS) using a common metric to quantify the EF in each device. It compares this new n=2 scaling to updated n=1 scalings using a larger 6 machine ITPA database. The results validate nonlinear single-fluid MHD simulation scalings, which are used to lend confidence to the projected scalings to ITER. These projections set the tolerances for non-axisymmetric components (like Test Blanket Modules) and the corresponding requirements for EF correction coil arrays in ITER.
Nonaxisymmetric fields four orders of magnitude smaller than the axisymmetric field can drive islands that cause disruptions in tokamaks; and the GPEC overlap metric, $\delta$, provides a way of identifying and quantifying the most dangerous of these asymmetries. The metric uses a resonant field spectrum determined from a combination of the applied field and the plasma response (amplification and/or shielding of various components)$^{1, 2}$, surpassing the robustness of the old “3-mode” vacuum model$^{3}$. Tolerances for the design and optimization of tokamak coils have been projected to ITER using fit scalings of this overlap metric with macroscopic 0D parameters that are easily identified both in current and for future devices.
The n=2 database, consisting of 3 devices and a parameter range shown in Figure 1, reveals tolerances of a similar order of magnitude to those for n=1 in current devices. Experimental thresholds are determined by ramping up artificial EFs using 3D field coil arrays until a core island penetration event is observed and recording the corresponding amplitude in EF overlap $\delta$. The predictions of a power law fit to this data using a kernel density weighted regression are compared to the true experimental thresholds in Figure 2. The full regression used has a density scaling exponent of 1.07±0.09, a toroidal field exponent of -1.52±0.2, a major radius exponent of 1.46±0.09 and a normalized pressure ($\beta_N/\ell_i$) exponent of 0.36±0.11. This fit projects very high n=2 thresholds for ITER due to the strong size scaling.
Although single parameter scans across regimes accessible by a single machine can reveal varied behaviors, the multi-variable, multi-machine scaling provides the most robust projection to new devices. Figure 3 shows individual density scans in DIII-D can exhibit a wide variety of trends depending on the initial target plasma. The second panel, however, shows how the seemingly discrepant DIII-D experiments are unified when normalizing by the general toroidal field and pressure scalings. Observations at even higher density show that ohmic confinement regime transitions in any given device can drastically alter the density scaling in that particular experiment. The ratio of non-resonant to resonant 3D field applied in a given experiment also alters the dynamics through neoclassical toroidal viscosity (NTV) braking. Recent advances in optimization of nonresonant fields for robust quasi-symmetry minimizing such secondary effects are beyond the scope of this dominant-order resonant field analysis. The local and secondary phenomena are purposely smoothed out in the multi-machine scaling presented here, in an intentionally analogous manner to the treatment of single-machine variations in the international confinement scalings that have proven so useful for the fusion community.
The nonlinear, single fluid MHD code TM1 has been used to model the experimental scalings, providing confidence in some scalings and insight into experimental needs. The model reproduces the experimentally observed toroidal field and $\beta_N/\ell_i$ scaling, and shows the scalings hold out to ITER values. The density scaling exponent calculated by TM1 falls below the experimental n=2 fit, closer to the better constrained n=1 empirical fit. This, and a large discrepancy between n=1 and 2 size scalings, identifies what n=2 data is needed to improve ITER projections. The code projects n=2 thresholds in ITER roughly 2-3 times that of the projected n=1 thresholds, consistent with observations in existing devices to date.
This combination of robust, cross-regime experimental scalings and tightly coupled modeling project EF tolerances of above a Gauss for both n=1 and 2 EFs in ITER, which are criteria the ITER construction tolerances and correction coils easily satisfy.
This work was supported by the U.S. Department of Energy, Office of Science, Office of Fusion Energy Sciences, using the DIII-D National Fusion Facility, a DOE Office of Science user facility, under Awards DE-AC02-09CH11466 and DE-FC02-04ER54698.
- J.-K. Park, A.H. Boozer, et. al, Physics of Plasmas 16, 056115 (2009).
- N.C. Logan, C. Paz-Soldan, et. al, Physics of Plasmas 23, 056110 (2016).
- J.-K. Park, N. C. Logan, et. al, ITPA MHD, ITER Report #ITER_D_UMLSUW (2017).
Speaker: Nikoas Logan (Princeton Plasma Physics Laboratory) -
Energetic particle (EP) instability models based on gyro-Landau closure techniques (1) have addressed important nonlinear simulation and linear stability survey challenges that will be critical for the understanding and control of burning plasmas in ITER and the next generation of fusion systems. The long-term intermittency and frequency spreading characteristics of saturated Alfvén instabilities (Fig. 1) are important features that will determine peak wall heat loads from escaping energetic ions as well as play a central role in regulating energetic particle anomalous transport levels. Reduced models using gyrofluid closures such as the TAEFL (2) and FAR3d (3) models, allow routine simulation of Alfvénic instabilities far into the nonlinear regime and demonstrate the dependence of the nonlinear dynamics on source-sink balances, zonal flow/current damping, and turbulence levels in the thermal plasma (modeled in Fig. 1 using diffusivities). The efficiency of this approach also facilitates linear instability surveys as profiles/parameters/plasma shapes are varied (Fig. 2); this capability is essential for simulating the dynamical changes that occur in realistic simulations of tokamak discharge evolution.
Reduced physics models for EP instabilities based on gyro-Landau closures (2,3) offer a computationally efficient tool for understanding effects of macroscopic parameters that control the saturated turbulent state of these instabilities. These models are based on an optimized set of closure coefficients that provide good fits to response functions derived from kinetic theory. Wave-particle resonances that exist in phase space are effectively mapped into real space while preserving growth/damping effects associated with EP energy distribution functions. These reduced models have been checked against more complete models (4). In the nonlinear regime, mode-coupling nonlinearities that drive zonal flows, zonal currents, localized flattening in the EP pressure profile, and couple to damped modes are included. The viability of simulating EP modes far into the nonlinear limit with these models is based on using a third order predictor-corrector time stepping algorithm, and the fact that there is no particle noise or small perturbation ordering, as typically limits simulation times for particle-based models. In regimes with moderate drive, balanced sources/sinks and where zonal flows/currents dominate, predator prey phenomena (5) are evident (black waveform in Fig. 1). As diffusivities are lowered or instability drive is increased, bursting phenomena and frequency chirping are observed (green waveform in Fig. 1). Saturation into a steady level can be achieved as EP drive is diminished by EP profile changes (Fig. 3). EP nonlinear transport fluxes (Fig. 4) and 3D convection cell phenomena can also be predicted.
The reversed shear/high bootstrap current tokamak regime offers the possibility of steady-state operation but is often unfavorable for EP-driven instabilities. Reversed shear/high bootstrap current discharge formation is a dynamical process and EP instability evaluation requires consideration of evolving profiles/parameters and plasma shapes. The gyrofluid model provides a unique, efficient multiple eigenmode solver approach that has been applied to these regimes. Low frequency Alfvén-acoustic (BAE/BAAE) instabilities are of particular concern since they can lead to larger radial transport levels than higher frequency AE modes (6). The survey capability of FAR3d has been applied to a DIII-D discharge with evolving q-profiles where multiple low frequency modes were observed (Fig. 2) and the results have shown similarities with the mode frequency evolution from the experiment.This work was supported by the U.S. Department of Energy under DE-AC05-00OR22725, DE-FC02-04ER54698, and the U.S. DOE SciDAC ISEP Center.
(1) G. W. Hammett, F. W. Perkins, Phys. Rev. Lett. 64, 3019 (1990).
(2) D. A. Spong, et al., Nuclear Fusion 53, 053008 (2013).
(3) J. Varela, et al., Nuclear Fusion 59, 046017 (2019).
(4) S. Taimoirzadeh, et al., Nuclear Fusion 59, 066006 (2019).
(5) D. A. Spong, Plasma and Fusion Research 9, 3403077 (2014).
(6) W. Heidbrink, et al., 16th IAEA Technical Meeting on Energetic Particles (EPPI 2019).Speaker: Dr Donald Spong (Oak Ridge National Laboratory) -
Energetic-particle (EP)-driven instabilities such as Toroidal Alfvén Eigenmodes (TAEs) can be responsible for the effective ion heating via collisionless EP energy channeling. Although the quantitative estimation of the EP transport by instabilities has been actively conducted [1–3], studies on the energy channeling have been limited [4,5]. It is important to estimate collisionless energy transfer from EP to ions for the development of burning plasma scenarios with effective nuclear reaction, and to compare this beneficial effect against EP transport due to TAEs.
For high-n TAEs which is expected to be in ITER, the modes strongly overlap with slightly different radial locations and eigenfrequencies. It was suggested that the interaction between the modes such as ion Compton scattering (ICS) [5–7] could be a dominant mechanism for the saturation of high-n TAEs, while the nonlinear wave-particle interaction by EP redistribution is dominant for low-n TAEs [8].
The wave-kinetic equations derived in [5,6] which describe the evolutions of the TAE mode energy lead to the predator-prey system between the linearly unstable modes in the TAE gap and the stable modes near the continuum. Figure 1 (a) shows the time-evolution of the TAE energy and the ion heating, and (b) shows the predator-prey like energy transfer by ICS in a single burst. Since the TAE gap structure depends on the radial profiles, the mode amplitude evolves in real space as well. Figure 1 (c) shows the mode energy propagates in a radial direction by the nonlinear mode transfer, which can lead to an avalanching EP transport.
EP transport by TAEs changes the linear growth rate, so that makes the TAEs eventually saturated. At the saturated phase over the transport time-scale when the nonlinear effect balances with the linear growth, we can estimate the net ion heating rate by TAEs from the linear growth rate.
In order to estimate the energy channeling as well as the EP transport by TAEs, we used TGLF to calculate the stability of TAEs [3 ]. Figure 2 (a) and (b) show the TGLF calculation of TAE stabilities. We see that there is a critical gradient of EPs to destabilize TAEs. Figure 2 (c) shows the EP profile transported by TAE which is calculated by critical gradient model (CGM) [3 ]. We also calculated the energy channeling through TAEs by using a property held during the wave-particle interaction, (dE/dt)/(dP_ϕ/dt)=ω/n which is valid for single-n TAE. We implemented an integrated simulation with ASTRA for ITER baseline scenario. We used TGLF calculation for EP transport, and BgB model for thermal plasma transport. In order to maximize EP effect, we set NBI-only external heating with P_NB=50 MW.
Figure 3 (a) and (b) show the thermal plasma profile and EP density profile for Q=10 ITER reference plasma without TAE effect. Figure 3 (c) shows the ion temperature profiles for 3 cases, (1) without considering TAE, (2) considering EP transport by TAE, and (3) considering EP transport and energy channeling. If we consider the TAE-induced EP transport, Q drops 18 %, and increases by 4 % if the energy channeling is considered. Figure 3 (d) shows the heating profile by EPs and the energy channeling. The ion heating by energy channeling is ~6 MW, about 10 % of the external heating, but the temperature difference is not much significant due to the profile stiffness.
References:
- L. Chen and F. Zonca, Rev. Mod. Phys. 88 015008 (2016)
- M. Podestà et al., Phys. Plasmas 23 056106 (2016)
- E.M Bass and R.E Waltz, Nucl. Fusion 60 016032 (2019)
- A. Bierwage, N. Aiba and K. Shinohara, Phys. Rev. Lett. 114 015002 (2015)
- T.S. Hahm, Plasma Sci. Technol. 17 534 (2015)
- T.S. Hahm and L. Chen, Phys. Rev. Lett. 74 266 (1995)
- Z. Qiu, L. Chen and F. Zonca, Nucl. Fusion 59 066024 (2019)
- H.L. Berk and B.N. Breizman, Phys. Fluids B 2 2246 (1990)
Speaker: Jaemin Seo (Seoul National University) -
In this work we show that nonlinear MHD plasma response simulations are essential for understanding and predicting accurate heat and particle flux striations in DIII-D during ELM suppression. Understanding the nature of heat and particle distributions on the divertor plates due to splitting of the separatrix by 3D magnetic perturbation in RMP ELM suppressed discharges is an important issue for preventing the premature degradation of divertor components in ITER. Our simulations now match the experimental observations in DIII-D. We further show that in cases with no observable heat flux splitting during ITER-like shaped DIII-D discharges with RMP ELM suppression the loss of heat flux splitting may be due to an increase in carbon radiation above the surface. Since carbon will not be present in ITER, these results raise a question concerning whether a similar effect will be observed in ITER during divertor detachment operations with medium Z impurities such as neon or nitrogen.
In DIII-D, strike point splitting is routinely observed in the divertor particle flux during operation with RMPs (1). We use small modulations of the in-vessel I-coil n=3 RMP current amplitudes or toroidal rotation of n=2 RMP fields in order to modify the poloidal spectrum of applied perturbation fields. These modulations affect the position and the size of the divertor footprints, and provide useful information to validate different numerical models for non-axisymmetric footprint lobes by comparing to high-precision visible imaging and IR tomography experimental measurement. The observed splitting is consistent with the toroidal mode number n of the applied perturbation but the measured separation of the divertor particle flux footprint lobes exceeds predictions of vacuum (TRIP3D) and linear plasma response (MARS) models by factors of 3-5 (2). The plasma response to the RMP in ITER-like conditions using linear, resistive MHD simulations (M3D-C1, NIMROD) with both single-fluid and two-fluid models is dominantly a screening response that reduces the divertor footprint splitting below the vacuum model predictions. The nonlinear MHD simulations using JOREK, a fluid model for the main plasma and the neutrals, show a much better match to the measured separation of the lobes (Figure 1). The time evolution of the neutrals in JOREK is described by a diffusion model combined with a boundary condition that reflects outgoing ions as incoming neutrals (3). In this work we also examine the possibility of near-SOL field lines affecting the formation of the outer-most lobes in the measured particle flux footprints.
At the same time, the heat flux to the divertor often does not show significant splitting in DIII-D ITER-like RMP ELM-suppressed discharge, as displayed in Figure 2, which is potentially good for ITER. One hypothesis suggested by observations is that the lack of splitting in heat flux may be related to the C-III volumetric radiation immediately above the surface that obscures the lobe structure. We performed fully 3D plasma-fluid and kinetic edge neutral transport Monte-Carlo EMC3-EIRENE simulations using vacuum and linear plasma response MHD solutions, and the results highlight the effect of different levels of carbon impurity radiation near the strike point on the divertor footprints. We also examined a possibility of heat flux lobe smearing partially due to the ion grad-B drifts as MAFOT simulations suggest. MAFOT results also indicate minor effects on heat flux footprints due to ExB fields and plasma sheath near the divertor surfaces.
(1) Moyer et al., RSI 89, 10E106 (2018)
(2) Orlov et al., IAEA FEC 2018 (Ahmedabad, India)
(3) Hujismans et al., 46th EPS, P1.1059 (2018)This work is supported by the US Department of Energy under DE-FG02-07ER54917, DE-FG02-05ER54809, DE-FC02-04ER54698, DE-SC0012706, DE-AC52-07NA27344, DE-NA0003525, DE-AC02-09CH11466 and DE-AC04-94AL85000.
Speaker: Dmitri Orlov (University of California San Diego) -
Novel internal measurements and analysis of ion cyclotron frequency range fast-ion driven modes in DIII-D are presented which advance understanding of the dynamics controlling mode stability and thereby the physics basis for prediction of fast-ion (e.g. alpha) transport in burning plasmas. Observations, including internal density fluctuation ($\tilde{n}$) measurements obtained via Doppler Backscattering (DBS) [1], are presented for two classes of instability: modes at $f \sim 0.6 f_{ci}$ (Fig. 1) and modes at low harmonics of $f_{ci}$. The former are identified as global Alfvén eigenmodes (GAE) (never before reported in a conventional tokamak) and the latter as coherent Ion Cyclotron Emission (ICE—so named because until now observations have always been via external RF antennae [2]). Both modes are excited by Doppler-shifted cyclotron resonance with fast-ions and observation of these modes in an experiment reveals the creation or movement of velocity-space inversion features in the distribution of fast-ions. Multiple aspects of the theory for this instability mechanism are validated. Highlights include: 1) a fast-ion density stability threshold is demonstrated for the GAEs [3] utilizing a unique DIII-D beam capability, consistent with a fundamental theoretical expectation that mode stability is controlled by competition between damping from various processes (e.g. Landau damping) and the drive from fast-ions; 2) recently developed analytic theory [4] predicting the frequency range for unstable GAEs is supported; 3) simulations [5] show unstable GAEs with frequencies and toroidal mode nubmers similar to experiment; 4) observations of ion-cyclotron harmonic $\tilde{n}$ in the core (Fig. 2) challenge a prevalent theory (see [6] and references therein) predicting the spatial localization of modes, and 5) simulations of edge ICE observations [7] from a DIII-D H-mode plasma agree well with experimental measurements from edge magnetics. Analysis of the ion-cyclotron harmonic $\tilde{n}$ observations is also presented leveraging the theory for the instability mechanism to advance understanding of ELM-induced fast-ion transport.
A fundamental element of stability theory for fast-ion driven modes is validated in an experiment demonstrating a fast-ion density threshold for fast-ion driven modes [3]. The modes, with frequencies $\sim 5.5 - 5.6 \text{ MHz} \sim 0.6 f_{ci}$ (Fig. 1a), are excited by off-axis nearly perpendicular beam injection. Using a unique DIII-D beam capability, beam current is slowly ramped down (Fig. 1b) at constant voltage in order to change the resonant beam ion density without significantly changing the resonance itself. The slow ramp is designed to create an approximately self-similar fast-ion distribution (neglecting initial transients at beam turn-on) changing only by a scale factor over time. The mode is abruptly stabilized as beam current crosses below a threshold of ~ 49 A (Fig. 1b), consistent with the expectation for a stability threshold set by competition between fast-ion drive and mode damping processes. (Note that a sawtooth briefly re-excites the mode after the initial stabilization. Note also that the mode frequencies sweep downward slowly as beam fueling increases density, and sawtweeth modulate the frequencies on a ~ 25 ms time-scale.)
The modes in Fig. 1 are identified as Doppler-shifted cyclotron resonant GAEs by analysis of the fast-ion distribution for destabilizing resonances, taking into account measurements of the toroidal mode number (via external magnetics) and internal mode structure. Fig. 1c shows a spectrum of core measurements (via DBS in reflectometry mode) with a narrow peak corresponding to the dominant GAE. Fig. 1d shows the time-dependent amplitude and radial mode structure, which is broad, obtained from a spatial array of simultaneous measurements. The identification is consistent with recently developed analytic theory [4] which, for these experimental conditions, predicts GAEs to be unstable in a narrow range of frequencies, $0.5 < \left. f \middle/ f_{ci} \right. < 0.7$. The identification is also consistent with simulations using the Hybrid MHD code (HYM, [5]) for the experimental conditions. The simulations found unstable, core-localized Doppler-shifted cyclotron resonant GAEs with mode numbers close to the measured value and $f ≳ 0.6f_{ci}$.
Unique observations of fast-ion driven modes with frequencies above $f_{ci}$ ($f \sim 10 - 50$ MHz $\sim 1 - 5 f_{ci}$) test numerous aspects of theory. Measurements of $\tilde{n}$ are obtained in a variety plasma conditions via DBS inside the plasma across a broad spatial range, giving mode amplitude and structure. Observed frequencies can match low harmonics of edge and core $f_{ci}$. Fig. 2a shows an example from an H-mode plasma which poses a challenge to a prevalent theory that predicts the modes to be radially localized close to where mode frequency matches a harmonic of $f_{ci}$ (see [6] and references therein). Fig. 2a shows a mode at $f \sim 15$ MHz, matching $f_{ci}$ on axis as expected, but it also shows modes at $f \sim 10$ and 20 MHz, unexpectedly matching the 1st and 2nd $f_{ci}$ harmonics at the edge. Simultaneous measurements at multiple radial locations show that the 10 MHz mode is actually core localized. Simulations with the EPOCH code test theoretical and computational analysis of ICRF modes, which indicates the central role of velocity-space population inversion for fast-ions for which $v_\perp \sim v_{Alfvén}$. Simulations of related edge ICE ($f ≥ f_{ci}$) observations from a DIII-D H-mode plasma agree well with experimental measurements from the magnetics, down to the level of the relative amplitudes of the first four ICE spectral peaks [7].
The observations in Fig. 2 also test other aspects of theory. The mode amplitudes strongly increase with total power from two beams injecting (nearly) tangentially in the co-current direction (Fig. 2b) at ~ 80 keV. However, the amplitudes are insensitive to power from beams with other injection geometries (e.g. co-current nearly perpendicular). This sensitivity to injection pitch and direction is consistent with expectation for the Doppler-shifted cyclotron resonance mechanism. The rapid switching on/off of the beams reveals a dynamic consistent with another aspect of the theory. In the ~ 10 ms period after each step-up in beam power, the mode amplitudes surge and then decay. This time scale is comparable to a fast-ion slowing down time, during which the high energy part of fast-ion distribution evolves from an initial bump-on-tail, with a positive energy gradient favorable to mode excitation, to a slowing down distribution, with a negative gradient. Fig. 2b also shows that ELMs transport fast-ions in the resonant region of phase space. In particular, when an ELM occurs during injection, the modes are transiently excited, consistent with the expectation that these modes can be excited by fast-ions ejected by MHD events [2].
In conclusion, these results strengthen the physics basis for predicting fast-ion driven mode activity and concomittant fast-ion tranport in future burning plasmas. Comparisons of experiment with analytic theory and simulation show areas of agreement, building confidence in the predictive power of these tools. Observations of ICE advance understanding of ELM-induced fast-ion transport.
This work was supported in part by the US Department of Energy under grant and contract numbers DE-FC02-04ER54698, DE-FG02-99ER54527, DE-SC0011810, DE-SC0019352, DE-SC0020337, and DE‐AC02‐09CH11466. It was supported in part from the RCUK Energy Programme grant no. EP/T012250/1, and carried out in part within the framework of the EUROfusion Consortium, receiving funding from the Euratom research and training programme 2014-2018 and 2019-2020 under grant agreement No 633053.
[1] WA Peebles, et al., RSI $\bf81$ (2010) l10D902.
[2] KG McClements, et al., NF $\bf55$ (2015) 043013.
[3] SX Tang, et. al., APS DPP Meeting 2019, BP10.43.
[4] JB Lestz, et al., PoP $\bf27$ (2020) 022513.
[5] EV Belova, et al., PoP $\bf26$ (2019) 092507.
[6] KE Thome et al. NF $\bf59$ (2019) 086011.
[7] A Zalzali, R O Dendy, K E Thome et al, in preparation.Speaker: Dr Neal Crocker (University of California Los Angeles) -
Runway electrons (REs) [1] in a tokamak is of great concern irrespective of the size of the machine. Such runway electrons carries significant amount of plasma energy of several MeV can severely damage the first wall and in-vessel components of the tokamaks [2] as well as can interfere with the complex plasma phenomena like plasma equilibrium, MHD instabilities and plasma disruption. Therefore the control of runway electrons are of great concern for ITER and DEMO like reactors. Similarly for reliable operation of smaller tokamaks such runway electrons should be suppressed or extracted without affecting the plasma operation. Several techniques exists viz., massive gas injection (DIII-D, TEXTOR), magnetic field perturbation (JT-60U, Versator I), resonant magnetic field perturbation (TEXTOR), additional gas-puff (JET, ASEDEX, ADITYA) [3] , ECH heating and LHCD (FTU) [4] for suppression/extraction the runaway electrons in tokamak. Another magnetic field perturbation technique similar to Versator I tokamak [5], for runway electron extraction has been applied in the ADITYA tokamak experiment. In ADITYA local vertical field coils (LVF coils) to generate local magnetic field perturbation has been successfully applied to extract runway electrons during early phase of discharge time (0.5 - 15 ms) which improved the plasma performance as reported in ref [6]. In this work, an attempt has been made to numerically model the runway electron extraction experiment using LVF coil in ADITYA. All coil systems to generate magnetic field configuration and the runway electron dynamics without radiation loss of the RE has been taken into consideration in the numerical simulation to explain the experimentally observed extraction of the RE in ADITYA tokamak. The magnetic field generated by all coil systems of ADITYA are numerically modeled and calculated by EFFI code [7]. The field calculated by the EFFI code then used by the PARTICLE3D code [8] to track the runway electrons following relativistic dynamics without radiation loss. PARTICLE3D stops following the electron when it hits the vessel wall so as to record that hit point which can be used to match the experimental observations in case of localized hard X-ray emission or physical damage of the wall.
The Ohmic coil system of ADITYA consists of one single air-core solenoid and 4 pair of vertical field coils. Toroidal coil systems of ADITYA has 20 TF coils to generate magnetic field of 0.75 - 1.5 T at the major radius 0.75 m. LVF coils are a pair of up-down symmetric coils placed on top and bottom TF coil I-Beam of ADITYA tokamak. The dimensions of the coils are ID $\approx$ 42.4 cm, OD $\approx$ 51 cm, $\Delta$r $\approx$ 4.3 cm, $\Delta$Z $\approx$ 8.5 cm, conductor dia. 0.85 cm, number of turns (n) $\approx$ 50 turns/coil and the distance from the equatorial plane (Plasma centre) to coil center (Z) is $\approx$ $\pm$ 80 cm. The coils are connected in series to produce LVF perturbation in the direction opposite to the actual equilibrium field at that location.
All coils of ADITYA tokamak as detailed above has been numerically modeled to generate the magnetic field configuration of the initial phase of the discharge. The toroidal field considered in this preliminary numerical simulation study is ~ 1.5 T at the major radius 0.75 m. Local vertical field used in the simulation corresponds to 4.2 kA of input current. A single runway electron of 3MeV energy has been traced using PARTICLE3D code with and without LVF coils. The simulated orbits of the 3 MeV runway electron for three cases (I) no LVF (II) +ve LVF and (III) -ve LVF are shown in figure 1(a).
It can be seen that under the influence of LVF coil the trajectories of the runway electron for three different cases differs significantly. While for + LVF coil the electron takes lesser time to be extracted than the without LVF coil, for _ve LVF the time taken is higher than with out LVF coil case. The time scale of the deconfinement is ~$10^{-8}$ sec for 3 MeV electrons which is very small as compared to the experimental time scale of $10^{-3}$ sec. The phenomenon can be easily understood from the plot of Z-position of the particle vs time as depicted in figure 1(b).
Therefore it can be concluded from the simulation results that under the influence of the local vertical field applied in the opposite direction to the equilibrium vertical field (at the mid plane of LVF coil location) the runway electrons gets deconfined. The simulation study shows the capability of the simulation methodology to model the RE deconfinement experiment and to explain the experimental observations. Though simulation of RE dynamics related to other phenomena in the tokamak have been reported earlier, however the simulation of RE deconfinement experiment using local magnetic field perturbation presented here is the first of its kind. Details of modelling and simulation results of the ADITYA RE deconfinement experiments with corresponding input variables would be presented. As RE mitigation is of great concern for larger tokamaks, the discussed simulation of RE de-confinement technique has the potential to be adopted for RE mitigation in such machines.Reference:
[1] H. Knoepfel, D.A Spong, “Runaway Electrons in Toroidal Plasma”, Nucl. Fusion, Vol. 19, No.6, (1979).
[2] C. Rasouli, D. Iraji et al, “Runaway electrons energy measurements using hard X-ray spectroscopy in Damavand tokamak, Rev. Sci. Instrum. 80 013503 (2009).
[3] Yu. K. Kuznetsov et al, “Runaway discharges in TCABR”, Nucl. Fusion 44 (2004), pg. 631 –644.
[4] S A Bozhenkov, M Lehnen et al, “Generation and suppression of runaway electrons in disruption mitigation experiments in TEXTOR”, Plasma Phys. Control. Fusion 50 (2008) 105007 (18pp).
[5] D.S Stone and B. Richards, “Magnetic extraction of runaway electrons from a tokamak plasma”, Phys. Fluids 21(1), January 1978.
[6] Tanna R.L. et al 2015 Novel approaches for mitigating runaway electrons and plasma disruptions in ADITYA tokamak Nucl. Fusion 55 063010
[7] Sackett, S. J., EFFI: A code for calculating the electromagnetic field, force, and inductance in coil systems of arbitrary geometry. 1978.
[8]. Dutta et al 2019 Plasma Sci. Tech. https://doi.org/10.1088/2058-6272/ab2947Speaker: Someswar Dutta (Institute For Plasma Research) -
Tokamak devices aim for magnetically confined burning plasmas in order to reach steady state operations and produce economically exploitable fusion energy. One of the main issues are the strong levels of transport due to the highly nonlinear turbulent plasma behaviour, which causes an increment of heat fluxes with respect to neoclassical theory. It is believed that microturbulence, characterized by small-scale and local resonant excited instabilities, is one of the principal drivers for the confinement degradation. Therefore, controlling and reducing the microturbulent transport is of paramount interest towards the exploitation of future devices.
A possible mechanism for such reduction could be identified in the fast ion population which is generated by neutral beam injection (NBI) and ion cyclotron resonance frequency heating systems in present day tokamaks. Recently, an enhancement of the thermal confinement with respect to the IPB98(y,2) scaling law has been experimentally detected in several devices, such as JET, ASDEX-Upgrade (AUG) and JT-60U, among others, in the presence of a significant amount of fast ions. Later, extensive dedicated gyrokinetic analyses demonstrated that fast ions beneficially impact on the ion-temperature gradient (ITG)-driven transport, reducing and partially suppressing the main ion heat fluxes [1,2]. The suprathermal species account for both stabilization of the linear growth rate and for an extra reduction of the nonlinear heat transport. Linearly, a wave-particle resonant effect - between the frequency of the instability and the magnetic drift frequency of the energetic species - has been established to be the major mechanism for the reduction of the ITG growth rate in JET L-mode scenario studies [3]. Subsequently, a complex multi-phase interaction between the ITG-scale turbulent transport and the zonal flow generation has been shown to affect beneficially the intensity of the heat transport [4]. The same nonlinear positive effect is found also in JET, DIII-D and AUG H-modes [5] and advanced scenarios, all of them being dominated by the ITG instability. In addition, a combination of NB-injected fast ions and α particles have been shown also in ITER predictive hybrid scenario to reduce the nonlinear saturated ion heat fluxes [6]. The latter results is significantly relevant since in ITER the E×B shearing turbulence reduction mechanism is expected to be weak; thus, fast ions could provide a valid alternative for such reduction.
Nevertheless, previous dedicated studies lack of generality, since multiple turbulence regimes, or even alternative ones, can usually be dominant in plasmas beyond the ITG paradigm. As a matter of fact, another relevant source of core microturbulence is represented by the trapped electron mode (TEM), often excited by the efficient electron heating systems or high density peaking, which can be subdominant to ITG modes.
In this paper, extensive gyrokinetic numerical studies have been performed with the local version of the GENE code [7] for a NBI-heated JT-60U hybrid scenario [8], previously identified with TEM dominated core turbulence by linear analyses [9], which share common characteristics with the already analysed JET hybrid scenario 2. The multi-species simulations also include collisions, magnetic fluctuations (both in the parallel and perpendicular directions), and plasma magnetic equilibrium computed by the CRONOS integrated modelling suite of codes [10].
Spectra from linear analyses reveal the destabilization of dominant fast-ion-driven beta alfvénic eigenmodes (FI-BAEs) at low binormal wavenumbers k_y, for the case labelled standard which is set with nominal input parameters computed by CRONOS. Subsequent nonlinear simulations for the same case show that significant electromagnetic fluctuations (identified through the value of the electron-β) drive even more unstable the FI-BAEs, leading to a drastic increase of the thermal and fast ion energy transport with respect to experimental range of values – as it is displayed in Figure 1 for the thermal Deuterium heat diffusivity. Hence, in order to evaluate the fast ion impact on dominant TEM-induced transport, FI-BAEs had to be stabilized. As a result, tuning both the thermal and suprathermal input physical parameters (principally the density and temperature gradients of the main and fast ion species, and also the electron-β), a stabilization of the fast-ion-driven mode has been achieved. Eventually, in this new configuration, the TEM has been found to be the dominant driver of the nonlinear saturated transport. Thus, it is shown that fast ions do not affect the TEM-induced turbulent transport [11], demonstrating that the turbulence suppression due to fast ion presence is not universal – in Figure 2, the heat flux time-traces are shown for comparison between the with and the without fast ions cases. This lack of impact is established up to the definitive excitation of fast-ion-driven modes that leads to a complex scenario in which strong electromagnetic effects, fast ion pressure gradient and thermal turbulent transport are intimately related. Therefore, in contrast to what occurs for ITG-dominated systems, the fast ions do not affect the TEM-induced heat fluxes in this JT-60U hybrid scenario.
A possible explanation for the different impact of fast ions on ITG and on TEM is related to the different saturation mechanism. Indeed, for the ITG instability, zonal flows (ZFs) are well-established to play a significant role in the saturation and also, as already stated, in the beneficial interplay among fast ions. On the other hand, dissipative TEMs, driven mainly by the strong electron temperature gradient, do not saturate through ZFs [12]. Deeper analyses performed for the same JT-60U discharge highlight that no energy exchange occurs between the TEM wavenumber and the zonal component of the flux, even when fast ions are introduced as an active species in the simulations. Furthermore, a local conservation relation for the particle toroidal momentum is derived and then applied to the GENE code in order to study the main physical parameters underlying the ZF excitation in the presence of fast ions, for which a competition among neoclassical damping and turbulent Reynolds-Maxwell stress effects is undergoing.
The results achieved opens the way to a more detailed physical view about the fast-ion effect on the microturbulence saturation process. In this sense, the universal physical mechanism governing the interaction among fast ion pressure gradient, electromagnetic fluctuations and microinstability-induced transport could be unveiled, leading to a better tailoring of the fast-ion species through the control of the external heating systems.
This work has been carried out within the framework of the EUROfusion Consortium and has received funding from the Euratom research and training programme 2014-2018 and 2019-2020 under grant agreement No 633053. The views and opinions express herein do not necessarily reflect those of the European Commission.
![Deuterium heat diffusivity for different fast ion pressure gradient cases scanned over the electron-β parameter. ‘Standard’ input parameters case is displayed here. The dashed vertical line represents the nominal value of β [11].](https://nucleus.iaea.org/sites/fusionportal/Shared%20Documents/FEC%202020%20Images/807/1.png)
![Timetrace comparison between the electrostatic heat fluxes with and without fast ions of electron and thermal deuterium species for the ‘modified’ case [11].](https://nucleus.iaea.org/sites/fusionportal/Shared%20Documents/FEC%202020%20Images/807/2.png)
References
- J. Citrin et al, Phys. Rev. Lett. 111, 155001 (2013)
- J. Garcia et al, Nucl. Fusion 55, 053007 (2015)
- A. Di Siena et al, Nucl. Fusion 58, 054002 (2018)
- A. Di Siena et al, Nucl. Fusion 59, 124001 (2019)
- H. Doerk et al, Nucl. Fusion 58, 016044 (2018)
- J. Garcia et al, Phys. Plasmas 25, 055902 (2018)
- F. Jenko et al, Phys. Plasmas 7(5), 1904-1910 (2000)
- N. Oyama et al, Nucl. Fusion 49, 065026 (2009)
- J. Garcia et al, Nucl. Fusion 59, 093010 (2014)
- J.F. Artaud et al, Nucl. Fusion 50, 043001 (2010)
- S. Mazzi et al, Nucl. Fusion, doi.org/10.1088/1741-4326/ab74a1 (2020)
- D.R. Ernst et al, Phys. Plasmas 23, 056112 (2016)
Speaker: Samuele Mazzi (Aix-Marseille Université, CNRS PIIM, UMR 7345 Marseille, France) -
Introduction
A powerful method for diagnosing runaway electrons in tokamak experiments is to measure the synchrotron radiation that the relativistic electrons emit. The radiation intensity depends on the electron energy, pitch angle as well as the background magnetic field, and using a spatially resolved diagnostic, such as a camera, allows the spatial distribution of electrons to be studied. Synchrotron radiation is therefore sensitive to the details of certain parts of the electron distribution function, making it ideal for validating kinetic models for runaway electrons. The detailed predictions obtained from kinetic runaway models can otherwise be difficult to test experimentally, as one must then usually rely on signals which are averages of the full distribution function.Principles
Synchrotron radiation is the ultra-relativistic extension of cyclotron radiation, and is as such emitted almost exactly along the particle velocity vector due to relativistic beaming effects. This means that while usual cyclotron radiation would give rise to a familiar torus of radiation on a camera image, a camera observing synchrotron radiation would only see one or a few distinct ``patches'' of radiation from one side of the torus, corresponding to the locations in the device where the electrons' velocity vectors are pointing at the synchrotron camera. This complicated dependence of the synchrotron patches on the background magnetic field means that a simulation tool taking the magnetic field geometry into account must be used, and for this reason we have developed the synthetic synchrotron diagnostic SOFT [1]. With SOFT, it is possible to simulate all types of synchrotron diagnostics used in present-day experiments, including visible- and IR cameras [2], spectrometers [3] and MSE polarimeters [4].Dominant particles
Generally, electrons with larger pitch angles and higher energies emit more synchrotron radiation. Conversely, most runaway electrons tend to have very small pitch angles, and if the population has been sufficiently multiplied through large-angle collisions, the number of particles with energy $\gamma$ decreases exponentially with $\gamma$. As a result, only a small subset of the runaway electron population contributes significantly to any synchrotron radiation measurement, corresponding to the particles that dominate emission of synchrotron radiation. This small region in momentum space is therefore the only region which we can hope to explore in true detail, although with the help of state-of-the-art runaway models, this is often sufficient to gain some understanding for the kinetic processes affecting the majority of electrons in the population.A synchrotron image or spectrum usually shows clear similarities to the image or spectrum that would result if all electrons had the same energy and pitch angle as the particle emitting the most synchrotron radiation. Hence, one can often characterize a synchrotron measurement with the ``dominant particle'' parameters. While one should be careful not to conclude that all particles have the same parameters from this, determing the dominant particle parameters can greatly help in modelling the scenario appropriately.
Figure 1 Examples of SOFT simulations for a JET-like ($B_0 = 3.45\,\mathrm{T}$, $R_{\rm maj} = 2.96\,\mathrm{m}$, $a = 1.25\,\mathrm{m}$) circular plasma with a parabolic current profile and on-axis safety factor $q_0 = 1$. The figures show synchrotron (a) polarization fraction and (b) polarization angle measured by a single MSE polarimeter line-of-sight [5] as a function of the runaway electron momentum $p$ and pitch angle $\theta_{\rm p}$.
Polarized synchrotron emission
The light emitted by runaway electrons is linearly polarized, and the polarization degree and direction of this light can be measured using for example MSE polarimeters [4,5]. These polarimeters are usually equipped with a number of narrow lines-of-sight, viewing across the plasma and observing light near the Balmer-$\alpha$ spectral line ($\lambda\approx 656\,\mathrm{nm}$). This is beneficial for observing synchrotron radiation, since the continuous spectrum of synchrotron radiation well covers the polarimeter spectral line, and oftentimes dominates over background radiation.The direction of the synchrotron radiation polarization vector is determined by the local magnetic field direction as well as the direction of motion of the emitting electron. Most of the radiation is observed to have either vertical or horizontal polarization, with very little light consisting of a clear super-position of the two. Hence, in a fixed magnetic field, a radially varying threshold is found in pitch angle across which the observed polarization sharply transitions between the two directions, as is shown in Figure 1. The sudden transition of the polarization vector also gives rise to a significant drop in the polarization fraction around the threshold. These signatures can help in identifying the pitch angle of the dominant particles, which are key to understanding experimental observations.
References
[1] M. Hoppe et al, Nucl. Fusion 58 026032 (2018).
[2] R. A. Tinguely et al, Plasma Phys. Control. Fusion 60 124001 (2018).
[3] R. A. Tinguely et al, Nucl. Fusion 58 076019 (2018).
[4] R. A. Tinguely et al, Nucl. Fusion 59 096029 (2019).
[5] N. C. Hawkes et al, Rev. Sci. Instrum. 70 894 (1999).Speaker: Mathias Hoppe (Chalmers University of Technology) -
Measurements of pellet-triggered edge localized mode (ELM) heat fluxes are presented here from experiments in ITER-relevant low collisionality pedestals (normalized pedestal collisionality $\nu^*_{ped}$ < 1) on DIII-D. These measurements demonstrate a reduction of peak ELM energy fluence at the inner strike point as compared to natural ELMs by as much as $\sim$50%. The inner strike point is typically the limiting location for material heat flux limits due to ELMs, so reducing the ELM heat flux there enables a higher allowable pedestal height before an ELM would need to be triggered. This in turn reduces the pellet pacing frequency required to protect materials and increasing the time-integrated pedestal pressure.
Pellet ELM pacing will be investigated in ITER in its pre-fusion power operation phases while access to RMP ELM suppression and/or other ELM-free operational regimes are assessed. Although previous experiments have demonstrated a reduction in peak ELM heat flux using pellet pacing for higher collisionality pedestals {1}, the capabilities and limits of this actuator for mitigating ELM heat flux and core impurity accumulation in low collisionality conditions are poorly understood. At low collisionality, the edge bootstrap current is increased for a given pedestal pressure, leading to pedestal growth being limited by the current density (kink/peeling instability) rather than the pressure gradient (ballooning instability). Exploring pellet ELM triggering in low-collisionality regimes in DIII-D tests the applicability of this ELM control actuator in this new regime and enables model validation so that we may more confidently extrapolate to ITER and other future devices. This pellet ELM triggering can then be extended to ELM pacing at a rate several times that of the natural ELM frequency, but the foundational experiments presented here use only infrequent pellets.
By shifting the location of the outer strike point away from the pumping duct during pellet ELM triggering while simultaneously attempting to access low collisionality conditions, heat flux measurements were obtained during pellet ELM triggering experiments in DIII-D. Figure 1 gives the ELM energy fluence profiles (the heat flux profiles integrated over 2 ms in this case) from one of these discharges for two natural ELMs (in blue and green), an ELM triggered by a large pellet (in red), and the inter-ELM heat flux integrated for the equivalent time frame, for reference (in dashed black). All four profiles are within the same $\sim$150 ms of DIII-D discharge 178555, with normalized pedestal collisionality < 0.7. The x-axis in Figure 1 follows the lower divertor target from the inner wall to the floor to the far scrape-off layer, with the inner strike point (ISP) and outer strike point (OSP) indicated around 0.6 m and 0.9 m, respectively. An area near the OSP is obstructed from view of the infrared camera used here, resulting in the shadowed region of Figure 1 between 0.98 m and 1.1 m where no heat flux measurements are available.
As seen in Figure 1, for large pellet–triggered ELMs, the relative amplitude of the integrated heat flux at the inner strike point is reduced by about a factor of 2, while a smaller increase in heat flux is observed in the far SOL. Because the ISP is the location of largest energy fluence for low collisionality ELMs in this dataset, the reduction of heat flux at this location would be beneficial for minimizing material damage for a given ELM size. The far scrape-off layer (on the shelf, where distance > 1.1 m) is also shown to have an additional heat flux lobe when large pellets trigger ELMs. This feature appears to be a three-dimensional structure that propagates in time as the ELM evolves, and also appears in the D-alpha fast camera images.
One goal of a pellet ELM triggering or pacing scheme would be to keep the peak ELM energy fluence values below the melting limit of the divertor material. The peak ELM energy fluences are plotted in Figure 2 for the ISP (star symbols) and OSP (circle symbols) for a series of natural and triggered ELMs from DIII-D discharge 178555. For natural ELMs, plotted as small unfilled symbols, the ELM energy fluence is almost always larger at the ISP than the OSP. This finding that the limiting value of ELM energy fluence occurs at the ISP is consistent with previous studies for naturally ELMing discharges with low collisionality pedestals in DIII-D {2}. $\nu^*_{ped}$ is $\sim$0.5 before pellets are injected and increases throughout the discharge, to $\sim$1.0 at 5.0 s and $\sim$1.5 at 5.2 s.
When large pellets (1.8 mm) are injected from the low field side of the device and trigger ELMs, as shown with large red symbols in Figure 2, the peak ELM fluence at the ISP is observed to be greatly mitigated. For ELMs triggered by smaller (1.3 mm) pellets plotted in blue, however, the heat flux distribution is comparatively unmodified with respect to the natural ELMs. Assuming the allowable ELM size determines the allowable pedestal height, these results suggest that reducing the peak ELM energy fluence at the ISP would increase the allowable pedestal height before an ELM triggering event is necessary. This could enable lower frequency pellet pacing and higher peak and time-averaged pedestal heights.
This work was supported in part by the US Department of Energy under award numbers DE-FC02-04ER54698, DE-AC05-00OR22725, DE-AC02-09CH11466, DE-FG02-07ER54917 and DE-AC52-07NA27344.
{1} L.R. Baylor et al., Phys. Rev. Lett. 110, (2013) 245001; Phys. Plasmas 20, (2013) 082513.
{2} M. Knolker et al., Nucl. Fusion 58, (2018) 096023.Speaker: Robert Wilcox (Oak Ridge National Laboratory) -
Reliable whole device modeling (WDM) of present and future burning plasmas critically depends on correct treatment of the auxiliary heating by energetic particles (EP) introduced into the plasma externally or by the fusion alpha particles. These energetic, or super-thermal, ions can effectively resonate with Alfvénic plasma oscillations, lead to EP losses and modify the profiles of the current drive. For burning plasma experiments, such as in ITER, it is important to not only predict the plasma operational regimes when those instabilities occur, but to learn how to operate in the regimes when the effects of the Alfvénic mode (AM) instabilities are benign. This is possible when all the essence of EP distribution function are accurately evaluated, which is envisioned in recently developed resonance broadened quasi-linear (RBQ) code [1].
We report on the RBQ code development with two dimensional (2D) computational capabilities to compute the distribution function evolution in time. This is an essential element of WDM simulations for self-consistent evaluation of EP distribution function. The RBQ code employs both the ideal MHD computed mode structures and their kinetic growth and damping rates. The velocity space diffusion coefficients in the constant of motion space are the result of RBQ simulations. They are transferred to the whole-device modeling code TRANSP.
RBQ utilizes a self-consistent quasi-linear (QL) theory in the presence of multiple unstable Alfvénic modes in the near-threshold regimes when the resonances between the EP and AM are broadened by a specific value prescribed by the analytic theory [2]. The formulation includes the discrete-resonance collisional functions to broaden the resonance regions for both Krook and Fokker-Planck scattering collisions. These functions are shown in Fig.1(b), which replace a simple resonance delta function that appears in the diffusion coefficient for the case of no broadening. For considered DIIID plasma the resonances are broadened by of the plasma minor radius independent on the mode amplitude. These functions remove a major arbitrariness with respect to previous resonance broadening approaches, which consisted of tuning broadening parameters to match the expected saturation levels [2]. The resonance functions are essential components of the QL model of the code to describe the dynamics of EP distribution evolution, where particle diffusion occurs in both canonical toroidal momentum and particle energy. We verify with the help of the guiding center code ORBIT that the resonances are indeed broadened by the amount determined by the pitch angle scattering frequency. The broadening enhanced by scattering is shown in left Fig.1(a), which shows that the broadened region is insensitive to the resonance island and fully determined by the scattering frequency.
The RBQ simulations find constant of motion diffusion rates for subsequent computation by TRANSP’s NUBEAM package using the probabi-lity density functions. The RBQ code has the capability to efficiently evolve the mode amplitudes simultaneously (in regimes of both overlapping and isolated resonances) while self-consistently relaxing the fast ion distribution function in the presence of collisions in both interpretive and predictive regimes. RBQ employs realistic eigenstructures, damping rates and wave-particle interaction matrices pre-computed by the NOVA-K code. The code in its 1D version was applied to DIII-D critical gradient experiments in predictive mode and reproduced the observed hollow fast ion radial profiles within the experimental uncertainties [1]. Examples of EP distribution function with and without canonical momentum diffusion computed by RBQ1D are shown in Fig.2.
The EP diffusion dependence across the resonance region allows the RBQ to compute the amplitude evolution which can exhibit oscillating, intermittent and saturated state behavior depending on the pitch angle scattering rate. This can be clearly observed in experiments when only few Alfvénic modes are excited such as in DIIID [3] and TFTR [4].
Our work reports on time-efficient, realistic simulations of fast ion redistribution and losses in tokamaks in presence of Alfvénic instabilities and their implementation in WDM using TRANSP. The RBQ code is upgraded to include 2D constant of motion diffusion with the goal to include simulate and design the burning plasma experiments, such as ITER. An important element of such simulations is to understand how tokamak devices can operate with unstable but benign Alfvénic instabilities. Another goal is to predict the current drive by beam ions and fusion products.
References
[1] N.N. Gorelenkov, V. N. Duarte, C.S. Collins et al., Phys. Plasmas 26 (2019) 072507.
[2] V.N. Duarte, N.N. Gorelenkov, R.B. White, H.L Berk, Phys. Plasmas 26 (2019) 120701.
[3] M.V. Van Zeeland et al., “Modification of Alfvén Eigenmode Drive and Nonlinear Saturation Through Variation of Beam Modulation in DIII-D”, IAEA TCM 2019, Shizuoka, Japan.
[4] K.L. Wong, R. Majeski, M. Petrov et al., Phys. Plasmas 4 (1997) 393.
Speaker: Nikolai Gorelenkov (PPPL, Princeton University) -
The Kinetic Orbit Runaway electrons Code (KORC) {1} has been extended to model post-disruption runaway electron (RE) dissipation by impurity injection incorporating state-of-the-art collisional models for partially ionized impurities {2} and models of thermal electron and impurity spatiotemporal dynamics. We fit these models to data from the DIII-D and JET tokamaks, exploring the role of spatial effects as compared to previous studies (e.g. Ref. {3}). This work presents model validation through detailed comparisons of shattered pellet injection (SPI) and massive gas injection (MGI) dissipation of REs and a detailed study of different collisional dissipation models. We find that the evolution of the spatiotemporal electron and impurity density profiles due to impurity injection and interaction with REs plays a significant role in the dissipation process and is an outstanding and critical modeling need. Additionally, we find that the dissipation timescale is fundamentally sensitive to the initial RE energy distribution. Beyond validation, this work allows for comparing and optimizing different impurity-based mitigation scenarios and their scaling towards ITER.
KORC has been extended to serve as a general framework for simulating RE physics, including validation and verification of the theoretical models needed to understand RE dissipation by impurity injection. To make calculations of RE dissipation numerically feasible, the RE guiding center orbit equations of motion have been implemented and are evaluated from experimental fields and plasma profile information via interpolation. The magnetic field configuration is held fixed and taken from experimental reconstructions, with EFIT used for JET and JFIT for DIII-D, as seen in overlaid (black) contours in the rightmost panels of Fig. 1 for JET pulse 95128 and DIII-D shot 164409. The toroidal electric field uses a 1/R spatial model fit to multiple experimental loop voltages. Initial RE distributions are initialized using a flexible sampling algorithm to initialize desired multidimensional distributions. The electron and impurity density profiles are taken from ad hoc models with a variable impurity charge state ratio that are fitted to experimental data using a synthetic line-integrated electron density (LID) diagnostic at multiple locations. The solid traces in the left panels of Fig. 1 correspond to experimental data from JET pulse 95128 with Ar SPI (top) and DIII-D shot 164409 with Ne MGI (bottom), and dashed traces correspond to synthetic LID signals for the profile shown in the right panels of Fig. 1 that are used in KORC calculations. A linearized, relativistic, Coulomb collision operator has been implemented via a Monte Carlo approach, where the effects due to bound electrons of partially ionized impurities are implemented through various models {2}. Additionally, synchrotron and bremsstrahlung radiation are included through additions to the guiding center orbit equations of motion.
The flexibility of KORC allows for simulations employing a hierarchy of models. Fig. 2 shows experimental plasma current from DIII-D shot 164409 with Ne MGI (top panel) and corresponding modeling results from KORC (bottom panel). The canonical case (black trace) includes bound electron physics, spatiotemporal electron and impurity density profiles fitted according to Fig. 1, a mix of singly and doubly ionized Ne, and an initial 10 MeV monoenergetic RE distribution. The critical importance of including bound electron collisional effects is evident from the different timescales of the without bound electrons (dark blue trace) and canonical cases. This physics is not only important for capturing accurate dissipation rates, but also the difference in dissipation rates for different injected impurities (not shown) and impurity charge state ratios (green trace). The KORC simulations also show a significant importance on the inclusion of a spatiotemporal density profile, as indicated by the difference between the constant and uniform density (red trace) and canonical cases. The initial energy distribution also plays an important role in the simulated dissipation timescale, shown by the difference between the canonical case and a case using an energy distribution inferred from a separate DIII-D study {4} (cyan trace). Interestingly, in the present work we find that for RE dissipation, as opposed to RE generation, the toroidal electric field plays a smaller role even when scaled (not shown), albeit a role that increases as the dissipation timescale increases. We additionally find that synchrotron and bremsstrahlung radiation have a comparatively minor effect (not shown). These results indicate the importance of including bound electron physics and spatiotemporal density evolution for assessing the efficacy and optimization of RE dissipation strategies for ITER.
{1} Carbajal et al., Phys. Plasmas 24, 042512 (2017)
{2} Hesslow et al., Phys. Rev. Lett. 118, 255001 (2017)
{3} Martin-Solis et al., Nucl. Fusion 57, 066025 (2017)
{4} Hollmann et al., Phys. Plasmas 22, 056108 (2015)*This work was supported by the US DOE under contracts DE-AC05-00OR22725 and DE-FC02-04ER54698 and by the ITER Organization (TA C18TD38FU) and carried out within the framework of the EUROfusion Consortium, receiving funding from the Euratom research and training programme 2014-2018 and 2019-2020 under grant agreement No 633053. The views and opinions expressed herein do not necessarily reflect those of the European Commission or the ITER Organization. This research used resources of the National Energy Research Scientific Computing Center (NERSC), a U.S. Department of Energy Office of Science User Facility operated under Contract No. DE-AC02-05CH11231.
**See the author list of E. Joffrin et al. accepted for publication in Nuclear Fusion Special issue 2019,
https://doi.org/10.1088/1741-4326/ab2276Speaker: Matthew T. Beidler (Fusion Energy Divison; Oak Ridge National Laboratory) -
Analysis of “super H-mode” experiments on DIII-D shows that high rotation, not high pedestal, plays the essential role in achieving very high energy confinement quality ($H_{98y2}$ > 1.5) $^1$. While the stored energy increases as expected with higher pedestal, the energy confinement quality mainly depends on the toroidal rotation (figure 1). At moderate rotation, similar to levels expected in ITER, very high pedestal height only yields marginal confinement quality improvement relative to standard H-mode. The dependence of the thermal energy confinement quality on the toroidal rotation is similar to that observed in the DIII-D advanced inductive scenario (also called hybrid scenario), which is typically characterized by high normalized fusion performance at high rotation, but suffers strong confinement degradation at low rotation $^2$. Linear gyrofluid and nonlinear gyrokinetic transport modeling confirm that the effect of E×B turbulence stabilization is far larger than other mechanisms, including EM stabilization and so-called hot-ion stabilization ($T_i/T_e$), as well as the fast ions effect.
Good confinement is critical for an economical fusion reactor design. Understanding the mechanisms leading to good confinement quality in a particular experimental scenario is essential to the ability to extrapolate that confinement quality to a future reactor. Super H-mode plasmas exploit strong plasma shaping to achieve high edge pedestal pressure at high density $^3$, and often exhibit excellent confinement quality exceeding standard H-mode, i.e. $H_{98y2}$ >> 1.0. It was initially believed that the very high pedestal was responsible for the very high energy confinement quality $^4$. However, careful analysis of the experimental data and transport modeling shows that the reason for the very high H-factor values is the rapid DIII-D toroidal rotation, which does not extrapolate to a reactor.
Figure 1 shows trajectories of several super H-mode discharges illustrating how the confinement quality is linearly correlated with the toroidal rotation in the core, while the pedestal pressure often increases as both the rotation and the confinement quality decrease. Very high confinement is reached early on in the H-mode phase of these discharges, when the pedestal is still low, but after the toroidal rotation has built-up to very high levels in the core. As the discharge evolves, the confinement quality is linearly correlated with the core toroidal rotation, which varies according to different levels of injected neutral beam torque per particle. Note that, even beyond the set of discharges analyzed in this paper, there is no experimental observation of very high confinement achieved at low rotation in any DIII-D super H-mode experiments to date. Also note that, as shown in figure 1, for given rotation the same $H_{98y2}$ is obtained with significantly different pedestal heights (the electron pedestal pressure is used as proxy of total pedestal pressure, since the ratio of electron to ion pedestal pressure stays fairly constant throughout the H-mode phase). Consistent with what one might expect, the stored energy is indeed higher with higher pedestal (not shown in figure), but other parameters (like plasma current, auxiliary power, etc.) must change to yield a different pedestal height, resulting in a constant $H_{98y2}$ for given rotation.
Figure 2 shows that linear gyrofluid modeling achieves a good match of the experimental observations by including the toroidal rotation and related E×B shear. The temperature reduction that results from removing the E×B effect in the simulation is calculated to decrease $H_{98y2}$ from 1.7 to 1.3. Similar analysis was repeated for several time slices in the same discharge, and all the calculated $H_{98y2}$ with and without ExB effect are reported in Fig.1 (blue circles), showing a similar prediction of the energy confinement quality without rotation, for time slices with different experimental confinement values. Note that all cases without ExB shear are slightly higher than the empirical “low rotation” $H_{98y2}$ (the dashed line shown in Fig.1) and this over-prediction may be due to having a fixed boundary condition at the pedestal in the modeling. Figure 3 shows an example of nonlinear gyrokinetic modeling exploring both the effect of the ExB shear and of the $T_i/T_e$ ratio. A large change in the normalized ion energy flux is predicted when the E×B effect in the simulation is removed. Little effect is predicted when the $T_i/T_e$ ratio is increased from the experimental value during the stationary confinement phase, ~1.16, to the peak value achieved in the discharge, ~1.32.
In conclusion, super H-mode plasmas can achieve very high energy confinement quality when accessed following a large neutral beam torque injection that transiently drives very rapid plasma toroidal rotation during and shortly after the L- to H-mode transition. As the density increases during the H-mode and super H-mode phases, the rotation decreases. With rotation closer to reactor-relevant levels, most of the confinement improvement over standard H-mode is lost, despite sustained very high pedestal pressure. Further studies on how to improve the energy confinement at low rotation will be key to the development of super H-mode scenario for a high performance fusion reactor.This material is based upon work supported by the US DOE under contracts DE-FC02-04ER54698, DE-AC02-09CH11466, DE-SC0010685, DE-SC0018287, DE-FG02-94ER54235
$^1$ Ding S. et al, 2020 Nucl. Fusion 60 034001
$^2$ Solomon W.M. et al, 2013 Nucl. Fusion 53 093033
$^3$ Solomon W.M. et al, 2014 Phys. Rev. Lett. 113 135001
$^4$ Snyder P.B. et al, 2015 Nucl. Fusion 55 083026Speaker: Andrea M. Garofalo (General Atomics) -
The high-power helicon antenna has been extensively tested at low power with excellent results, and the full installation in the DIII-D vessel was successfully completed in February 2020, as shown in Fig. 1. High-power striplines feed the antenna from either end (not visible in Fig.1). Commissioning of the 1.2 MW 476 MHz klystron source and associated high voltage power supply and switching network is scheduled to be complete by the end of Spring 2020, enabling the first high-power helicon current-drive experiments in the Summer of 2020. Helicon current drive, or fast wave current drive in the lower hybrid range of frequencies, has long been predicted to be a promising current drive tool for reactor grade plasmas. The system at DIII-D will be the first test of high-power helicon current drive in reactor-relevant plasmas, where full single-pass absorption is expected [a]. Substantial off-axis driven current is predicted for DIII-D (~60 kA/MW).
The antenna is a traveling wave antenna of the comb-line type with 30 modules, fed from either side to allow both co and counter current drive. Power is transferred down the antenna through mutual coupling from one module to the next with low losses. All 30 antenna modules were tuned individually to the same resonant frequency to within ±0.5 MHz, and show low losses with a quality factor in excess of 1000. The linear electromagnetic characteristics of the unloaded module array has been extensively tested on the bench and on the DIII-D mock-up wall at instrumentation power levels, with sub-arrays of 10, 15, 20 and 30 modules on curved backplates. The six backplates are coupled capacitively to minimize disruption forces while providing a good RF ground plane. The placement of the modules on the backplates in a 2-module poloidal stagger pattern, which increases the opacity of the Faraday shield to field line penetration, resulted in a periodicity in the mutual coupling between modules that can drive large reflections (~20% power reflected). Adjustment of module placement compensates for this, resulting in a low 1-2% power reflection coefficient and 98.7% power transfer from module to module in air, as shown in Fig. 2 for the 30-module installation in the DIII-D vessel. Also shown in Fig. 2 is data obtained with a pick-up coil in front of the straps just outside the Faraday screen bars. The phase progression and probe amplitude are steady along the length of the array, as expected for a traveling wave antenna.A 1.2 MW klystron, obtained from SLAC, and the associated high-voltage power pulsing network is being installed and commissioned. The klystron with circulator and dummy load feeds power into a 4-port switching network. The 4-port switch allows feeding of a stripline on either end of the antenna, with the reflected power and the remaining transmitted power each dumped into the dummy load. To satisfy the vessel constraints, the stripline geometry was 3D printed out of Inconel and stainless steel, and copper-coated. The striplines act as a 180-degree phase splitter combined with two 3dB splitters. The four equal-power out-of-phase signals are fed to the coupling straps on the end-modules of the antenna. The antenna and striplines are instrumented with pick-up coils, arc detectors and thermocouples.
Several new and upgraded diagnostics are planned for the helicon program. An edge-density diagnostic will provide density profiles at the antenna. A laser-based technique known as Doppler-free saturation spectroscopy (DFSS) will be implemented to measure the helicon wave electric field vector over a 2D region of space in the edge plasma near the antenna. The wave electric field vector is obtained by fitting the Schrodinger equation to Doppler-free Dβ spectra. This technique results in an accuracy of <5 V/cm, two orders of magnitude greater than that of previous work. For far-field wave measurements, the existing Phase Contrast Imaging (PCI) diagnostic is being upgraded with a heterodyne detection system able to detect fluctuations at 476 MHz. Using values for the wave electric field estimated by full-wave modeling work, it is expected that the PCI will be able to measure the spatial structure of perturbed density with favorable signal-to-noise ratio.Helicon wave current drive will soon be investigated in high density, high electron beta plasmas where full single-pass absorption is expected – conditions similar to that in steady-state AT burning plasma devices.
This work was supported in part by the US Department of Energy under DE-FC0204ER54698, DE-AC05-00OR22725, DE-FG02-94ER54235, DE-SC0013911 and DE-AC02-09CH11466.
[a] R. I. Pinsker et al, Nuclear Fusion 58, 106007 (2018)Speaker: Bart Van Compernolle (General Atomics) -
Understanding impurity transport within fusion research plasmas is of critical importance as progress is made toward burning plasmas. High core impurity concentration can have a deleterious effect on plasma performance: fuel dilution (for fully stripped low Z impurities), triggering of MHD instabilities, or increased core radiated power (for high Z impurities), thus hindering the achievement of stable long-pulse, high energy confinement discharges. A full understanding of impurity transport conditions is especially crucial when observations from multiple devices display significantly different results to what is presumably a similar set of conditions. The effects of evaporated lithium on the NSTX lower divertor and the effects of lithium powder injection were primarily evidenced in the edge/pedestal region where changes in the electron pressure profiles resulted in ELM free operation1. This state however did not lead to an enhanced lithium presence in the core with concentrations rarely rising above 0.1%. Lithium aerosol injections into DIII-D2 discharges utilized 90% less material, but were also successful in generating 300 ms ELM free periods and enhanced pedestal pressures. However these discharges saw substantial penetration of the lithium into the core plasmas, with measured concentrations up to 15-20%. New experiments combining the introduction of lithium and boron powders counterposed with carbon and lithium granule injections have shown the ability to manipulate the core carbon concentration to levels both higher and lower than those observed in reference discharges allowing a comparison with neoclassical impurity transport theory.
According to neoclassical theory, carbon will act as the dominant impurity and screen out lower-z impurities above a threshold concentration. To determine if a carbon threshold effect can mitigate core lithium penetration, powders and granules were introduced into DIII-D ITER baseline discharges with IP = 1.3 MA, BT = -1.7 T, ne = 8x1019 m-3, PNBI = 9.5 MW and N = 2. Powders are gravitationally dropped into the upper edge of a lower single null H-mode discharge by means of an impurity powder dropper (IPD)3 and granules were horizontally injected into the discharge utilizing an impurity granule injector (IGI)[4]. A reference discharge shown in Figure 1 displays a rapid increase in core carbon concentration during the transition to H-mode which stabilizes to a near constant signal intensity that is maintained during the discharge.
To modulate core carbon concentration the two actuators were utilized separately. Core carbon levels were increased by injection of 400 micron vitreous C granules 2.5 – 4.5 s at mass injection rates of 6 and 4 mg/s, respectively. Measurements of core carbon impurities saw the signal levels start to rise at approximately t = 3 s with the higher injection rate reaching a saturation point after 300 ms. Saturation occurs at an intensity twice that of the baseline level as seen in panel 1 of Figure 2. These results are contrasted with those seen from Li powder injection. For these cases, the introduction of lithium powder caused a prompt reduction of the carbon signature by a factor of 4x which was maintained for the duration of the injection event before returning again to the baseline level. The introduction of Li powder also had a marked effect upon the ELM cycle, with the higher level of injection resulting in extended periods of ELM free behavior as seen in figure 3. These results are consistent with previous reports of Li powder injection2.
To determine if the lithium buildup within the DIII-D core plasma during Li dropper operation can be flushed by carbon microgranule injection a combination of the two actuators was used. As is shown in panel 2 of figure 2, Li powder is utilized to drop the core carbon concentration prior to the introduction of carbon granules. Carbon levels are seen to increase during the period of C injection from 3.5-4.5 s. However, once the granule injection is stopped the C signal returns to the Li suppressed baseline level rather than remaining at the standard carbon level, thus demonstrating a substantial pumping effect by the lithium leading to the observation that this carbon level does not provide an impurity screening effect. In contrast, panel 3 shows the results when the two actuators are temporally swapped. The long carbon granule pulse is active from 2.5-6 s while the short lithium powder pulse is active from 2.5-5 s. At these elevated carbon levels the Li aerosol does not seem to be able to have a substantial effect on the core carbon intensity, thus there is the possibility that a carbon transport barrier has been established. While increasing the Li powder rate may overcome this barrier, these discharges are already at a substantial impurity fraction and it was observed in subsequent injections with boron powder that larger impurity injections lead to a premature discharge termination event.
As can be seen from these results, there is a complex interplay between the impurities present within the plasma core dependent upon both the species of impurities introduced as well as the order in which they are deposited into the discharge. These measurements provide benchmarking data for neoclassical transport codes such as NEO and XGC to help determine if the corresponding variations in impurity transport can be explained by simulation and will inform favorable transport conditions in future tokamaks.This material is based upon work supported by the U.S. Department of Energy, Office of Science, Office of Fusion Energy Sciences, using the DIII-D National Fusion Facility, a DOE Office of Science user facility, under Awards DE-FC02-04ER54698 and DE-AC02-09CH11466
1 F. Scotti et al., Nucl. Fusion 53 (2013) 083001
2 T. H. Osborne et al., Nucl Fusion 55 (2015) 063018
3 A. Nagy et al., Rev. Sci. Instr. 89 (2018) 10K121
[4] M. Vorenkamp et al., Fusion Sci. & Tech. 72:3 (2017) 488Disclaimer: This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights. Reference herein to any specific commercial product, process, or service by trade name, trademark, manufacturer, or otherwise does not necessarily constitute or imply its endorsement, recommendation, or favoring by the United States Government or any agency thereof. The views and opinions of authors expressed herein do not necessarily state or reflect those of the United States Government or any agency thereof.
Speaker: Robert Lunsford (Princeton Plasma Physics Laboratory) -
We present an overview of theory and simulation of low-frequency drift Alfvén waves (DAW) in toroidal fusion plasmas based on the framework of the general fishbone like dispersion relation (GFLDR) [1, 2, 3]. In addition to recovering various limits of the kinetic MHD energy principle, this approach can also be applied to general e.m. fluctuations characterized by a broad range of spatial and temporal scales consistent with gyrokinetic descriptions of both core and supra-thermal plasma components. Therefore, the GFLDR provides a unified description of DAW excited by energetic particles (EPs) as either Alfvén eigenmodes (AEs) or energetic particle modes (EPMs). Furthermore, the GFLDR is given in the following generally variational integral functional form
$\begin{equation} 2 \pi^2 \int_0^a dr \left[ \frac{k_\vartheta^2 c^2 (d\psi/dr)}{{\cal J} B_0^2} \right]_{\vartheta = 0} \left| A(r)\right|^2 \left[ \delta \bar W_f (r) + \delta \bar W_k (r) - i \lambda (r) \right] = 0 \; . \;\;\;\;\; \end{equation}$
For radially localized fluctuations, the GFLDR can then be written as
$\begin{equation} i \lambda (r) = \delta \bar W_f (r) + \delta \bar W_k (r) \; , \;\;\;\;\; \end{equation}$
where $\Lambda$ represents a generalized inertia, while $\delta \bar W_f$ and $\delta \bar W_k$ describe the potential energy of the fluctuations accounting, respectively, for the fluid and kinetic plasma response [1, 2, 3]. From the structure of Eqs. (1) and (2), the present approach allows extracting spatial and temporal scales of the considered fluctuation spectrum as well as the underlying physics. For example, radial singular structures of the shear Alfvén wave (SAW) continuous spectrum are shown in FIG. 1 using the analytic as well as the numerical solution for the generalized inertia, Λ, in the MHD fluid limit [4, 5]. Meanwhile, analytic results on mode frequency, damping and finite orbit width effects can be verified numerically with very good agreement for both SAW continuum accumulation points [6, 7] as well as for geodesic acoustic mode oscillations [8]. Thus, the GFLDR as a unified theoretical framework can help understanding experimental observations as well as numerical simulation and analytic results with different levels of approximation.
The gyrokinetic analysis of low-frequency DAW is necessary for a proper description of short wavelengths and/or accounting for finite parallel electric field and wave damping as well as for EP excitations [3, 6, 7]. Here, we derive the corresponding GFLDR theoretical framework in general tokamak geometry as a joint effort of ongoing research projects [9, 10]. As first application of the general theory, we extend to gyrokinetic analysis the application to the Alfvén - acoustic fluctuation spectrum of Ref. [4] in the MHD fluid limit. This analysis, presented in detail in a dedicated contribution [11], confirms the general prediction that fluctuations with acoustic polarization are unfavored because of the stronger Landau damping and the weaker response to EPs [4], in agreement with recent observations in DIII-D [12].
As further application, motivated again by recent experimental observations [12] consistent with earlier theoretical predictions [4], we apply the GFLDR theoretical framework to simple (s, α)-equilibrium tokamak geometry. In particular, we use the analytic expression of the generalized inertia, Λ, derived earlier for well circulating/deeply trapped thermal ions [13, 14]. Our results, presented in detail in a separate work [15], show, again, that the acoustic branch is strongly suppressed by Landau damping and that EP excitation is generally weak.
Numerical results by the Drift Alfvén Energetic Particle Stability (DAEPS) code will also be shown and discussed [10]. DAEPS belongs to a long-term project launched to better relate experimental observations and theoretical/physical understandings of DAW (AEs/EPMs) interacting with thermal plasma as well as EPs (including alpha particles). DAEPS is a gyrokinetic eigenvalue solver, which can properly represent kinetic compression and wave-particle resonances of core and energetic plasma components. Compared to initial value codes, the DAEPS approach is a valuable method to analyze the whole stable/unstable spectrum besides the most unstable mode. As demonstration, the beta-induced AE (BAE) is used to illustrate the relevant aspects of the code.Acknowledgments
This work was supported by EUROfusion Consortium and ITER-CN grants.[1.] Zonca F and Chen L 2014 Phys. Plasmas 21 072120
[2] Zonca F and Chen L 2014 Phys. Plasmas 21 072121
[3] Chen L and Zonca F 2016 Rev. Mod. Phys. 88 015008
[4] Chen L and Zonca F 2017 Phys. Plasmas 24 072511
[5] Falessi M V, Carlevaro N, Fusco V, Vlad G and Zonca F 2019 Phys. Plasmas 26 082502
[6] Bierwage A and Lauber Ph 2017 Nucl. Fusion 57 116063
[7] Lauber Ph and Lu Z 2018 J. Phys. Conf. Ser. 1125 012015
[8] Biancalani A, Bottino A, Ehrlacher C et al. 2017 Phys. Plasmas 24 062512
[9] MET Enabling Research Project, https://www.afs.enea.it/zonca/METproject/
[10] Li Y et al. 2020 “Gyrokinetic analysis and simulation of Alfvén Eigenmodes” To be submitted
[11] Chen L et al. 2020 “On energetic particle excitations of low-frequency Alfvén eigenmodes in toroidal plasmas: gyrokinetic theory” This Conference
[12] Heidbrink W W 2019 Private communication
[13] Chavdarovski I and Zonca F 2009 Plasma Phys. Contr. Fusion 51 115001
[14] Chavdarovski I and Zonca F 2014 Phys. Plasmas 21 052506
[15] Chavdarovski I et al. 2020 “Effects of core plasma on the low frequency Alfvén and acoustic eigenmodes” This Conference.Speaker: Dr Fulvio Zonca (ENEA C.R. Frascati)
-
-
