First, an MHD equilibrium is calculated for the time slice of interest using the EFIT code. Effects such as further broadening of the current profile and electron heating have been demonstrated to close windows on energetic particle driven instabilities. The high li, high βN discharges have not yet been operated with a stationary current density profile as sufficient externally-driven current is not yet available (but planned upgrades are projected to address this—see section "Future Research Requirements and DIII-D’s Role"). Shaping is a key lever to raise pedestal pressure; for example with higher triangularity (a rise in triangularity to 0.9 is possible with removal of the upper inner cryopump in DIII-D, Fig. DIII-D research in recent years has taken substantive steps in validating and developing a model based understanding of this concept. about the last third of the minor radius) of a small number of DIII-D tokamak9 L-mode discharges failed to reproduce the experimental heat flux levels.10,11 One of the first and most detailed publications in this context is Ref. Also, the drive for AEs can be altered by varying the radial gradient in the fast ion pressure profile grad-βfast [129, 136] to keep it below a critical threshold [139] through variation of density and NBI power as well as NBI injection geometry. ECH will also enable precise perturbative tests of turbulence and pedestal behavior. This naturally occurs in high li and high qmin scenarios. This has enabled the injected beam ion population to evolve with plasma characteristics in order to navigate the AE space, to optimize heating and AE stability in real time, and thus discharge performance [66]. Van Zeeland, C.T. Key opportunities include assessment of long pulse control technologies and long time scale wall evolution [147]. Control. The result is the capability to create flat or potentially off-axis peaks in the fast ion pressure profile and thus completely stabilize the RSAEs (Fig. In, M. Okabayashi, S.A. Sabbagh, E.J. However, if these prove insufficient (for example, if energy confinement is too low) additional auxiliary heating will be required, Padditional heating, to balance the power lost from the plasma due to transport. While turbulent transport is known to generally degrade as the electron-to-ion temperature ratio increases from below 1 to near unity [97], recent experiments in steady-state scenario plasmas on DIII-D have shown that negative central magnetic shear (NCS) mitigates the transport degradation observed during electron cyclotron heating [98]. Achievement of an understanding of the advanced scenarios discussed in the previous section and the projection of scenarios to future reactors present new physics challenges. major radius R 0 ¼ 1:76m, minor radius defined by the hori-zontal distance from the axis to the separatrix a¼ 0:72m, toroidal magnetic field B 0 ¼2.03T, and deuterium to electron mass ratio m D=m e ¼3672. Soukhanovskii et al., Phys. On the other hand, the advantage of the SF-DN was tied to a much larger poloidal flux expansion at its lower outer divertor target, i.e., ≅ 28 for the SF-DN versus ≅ 9 for the DND; the result was a factor of two lower peak heat flux at the outer divertor target with the SF-DN shape, both before and during deuterium gas injection operation (Fig. The second key consideration is the role of \(q_{95}\). We leave technology development such as higher field superconductors to other studies, though their leverage on these issues make them important to pursue also. The DIII-D program achieved several milestones in fusion development, including the highest plasma β (ratio of plasma pressure to magnetic pressure) ever achieved at the time (early 1980s) and the highest neutron flux (fusion rate) ever achieved at the time (early 1990s). Conversely, since the plasma current is relatively low at high \(\beta_{P}\), maintaining a high fusion power density generally requires operation at high \(\beta_{N}\), and at confinement quality higher than standard H-mode. This requires a population of current carriers with velocity close to the wave phase velocity, and low collisionality as collisions scatter the current carriers [29]. A key mission of the DIII-D program is to develop the physics basis for fully noninductive steady-state operation at high normalized β. The challenge for a tokamak steady state reactor is to have sufficient fusion performance to generate net energy, after powering auxiliary systems to sustain the regime non-inductively. radius of the flux surface, i.e., πBφ,0ρ2 =Φ. Doyle, J.R. Ferron, A.M. Garofalo, J.C. Hillesheim, C.T. Campbell, Tokamaks, 3rd edn. Research in recent years has focused on characterizing this relationship in relevant electron heated and low rotation regimes to aid the development of predictive transport models ("Turbulent Transport in High β Scenarios" section). Lett. In addition to guiding the DIII-D experiment and H/CD upgrade ("Future Research Requirements and DIII-D’s Role" section), the same FASTRAN modeling has been used to extrapolate the steady-state scenarios developed on DIII-D to future reactors such as ITER [51] and C-AT DEMO [52; journal paper in preparation], thereby identifying gaps between the present-day experiments and future reactors, and further research needs on DIII-D. One such example for the ITER projection is shown in Fig. Both configurations were biased slightly toward the lower divertor. Petty, R. Nazikian, J.M. The capabilities discussed above will position DIII-D well to validate physics models and develop technical solutions for phenomena from the core to the edge at reactor relevant parameters for each region, developing a valuable projective physics understanding. Solomon, E.J. Control. Control. Consistent with the stability calculations, a global, pressure-limiting instability has not yet been clearly observed in the experiment. Active stabilization of the n = 1 RWM was applied to some of the high-qmin discharges in Fig. 97, 135001 (2006), C.Z. Many of the features of such a discharge have been demonstrated in DIII-D, and the highlights of the most recent experiments [48] are reviewed in this section. These are particularly important in DIII-D plasmas, since the 80 keV beam energies are not able to fulfill the fundamental wave-particle resonance condition at normal operation fields. Plasmas 14, 056102 (2007), R. Nazikian et al., Phys. Time evolution of parameters in two high li discharges with high βN. Plasmas 24, 056114 (2017), J. McClenaghan et al., Nucl. Plasmas 23, 062511 (2016), J. Qian et al., Nucl. Fluids B 3, 1865 (1991), A.D. Turnbull et al., Phys. Penetration depths for neutral ionization, ΔCX, scale predominantly with density (ΔCX = 1.91E17 T0.425ped Notably, its performance in maintaining stability above the no-wall limit using more reactor relevant external control coils matched that of a proportional gain controller using internal coils—a significant development [112]. And a parallel program will develop improved divertor concepts with increased closure. 41, 42) will test this physics, providing the first detailed spectral optimization studies for ELM and rotation control with n = 3 or 4 fields. Van Zeeland, Plasma Phys. Austin, J. Lohr, Rev. This work demonstrated that certain shapes strongly suppressed a variety of instabilities in the plasma, which led to much higher plasma pressure and performance. Energy 12, 1141 (2017). In 2018, DIII-D neutral beam systems will be re-oriented to double off-axis current drive power. Discharge performance was close to the estimated requirement for the ITER steady-state mission with G = βN H89/q295 Fusion 45, 30 (2005), M. Podesta, M. Gorelenkova, E.D. 58, 613 (2010), H. Zohm et al., in Proceedings of 43rd EPS Conference on Plasma Physics (2016), K. Tobita et al., Nucl. Plasmas 15, 056107 (2008), W.W. Heidbrink et al., Plasma Phys. Vertical dashed line marks time when the Ohmic heating coil current is fixed. This enhanced 3-D coil set coupled with upgraded 3-D diagnostics will also be able to probe the plasma response to applied field in order to measure RWM dissipation physics at n up to 6, and develop advanced active control techniques at βN approaching the ideal-wall limit (Fig. In section "Future Research Requirements and DIII-D’s Role", we discuss plans for the facility as it goes into an exciting series of upgrades to increase current drive and heating capabilities to explore reactor like advanced scenarios with βN up to ~ 5, and develop compatibility with and techniques for a detached divertor solution. A new saturation model in TGLF that incorporated this effect improved accuracy of temperature predictions, with the electromagnetic effects reducing temperature gradients, to achieve a better match to experiment as shown in Fig. Fusion 57, 036018 (2017), A.M. Garofalo et al., Plasma Phys. Reaching high bootstrap fraction fully non-inductive plasmas, requires operation at high poloidal beta, βP, the pressure relative to poloidal field. Collisionality and rotation can further play important roles in pedestal stability and height. Like the hybrid, bootstrap current fraction is typically ~ 50%. Linear simulations using the GKV electromagnetic gyrokinetic Vlasov code find that the growth rates of low and higher-k modes increase less with electron heating for NCS plasmas than for PS plasmas, and the dominant mode switches from ITG to TEM in the PS region as Te/Ti is increased. More reactor relevant DND plasmas that produce significant power outflow will very likely have peak divertor heat flux levels well above 10 MW/m2 and thus require reliable methods of reducing excessive heating to the divertor structures. High density will also lead to increased electron–ion collisionality to explore more equilibrated transport dynamics, and thus capture turbulent transport interactions in reactor-like electron–ion coupled regimes. The proximity of these two profiles both (1) improves confidence that FASTRAN contains most of the important physics, and (2) confirms that the experimental discharge was approaching a stationary state. Research has also developed major advances in physics understanding, validating concepts of kinetic damping of ideal MHD instabilities that enable high β operation, identifying how current profile and β influence plasma turbulence in order to validate and improve turbulent transport models, and understanding the physics of energetic particle redistribution due to Alfvénic and other instabilities. The scaling paper [1] describes in more detail the logic for picking out this single region in minor radius to establish this velocity given the variety in a typical H-mode intrinsic rotation Rev. values would be obtained at double the present accessible density. This suggests that with further work, it may be possible to translate the learning and physics advances being developed with internal coils, as discussed above, to practical reactor relevant external coils. 21, the low-k broadband turbulent fluctuations increase less in NCS plasmas compared to PS plasmas with increased Te/Ti [98]. Recent experiments on the DIII-D tokamak have focused on determining the effect of trapped particles on the electron cyclotron current drive (ECCD) efficiency. Gerhardt, J. Manickam, M. Podestà, Phys. Sci. Not least it is the approach for ITER’s steady state research toward fusion energy [13]. This has the potential to close the gap on reactor parameters in order to enable the development of integrated solutions for fusion energy and proceed with a generation of steady state fusion reactors. Fusion 54, 093009 (2014), R.E. The illustrated fast-ion profiles are from classical calculations performed prior to the experiment. It should be noted that other scenarios such as steady-state hybrid plasmas (with qmin~ 1–1.3) are also susceptible to EP transport, and, depending on qmin, they can have either AEs or Fishbones that cause significant transport and limit performance [51]. Plasma instabilities, including edge-localized modes (ELMs) and core, tearing, or global instabilities that lead to disruptions, could prevent reactors achieving their mission through damage to the facility or de-rated operation to avoid potential damage. From the basic scaling for collisionality, ν* ~ n3e Luxon et al., Nucl. Holcomb, T.C. Further, this divertor-pedestal interaction is itself altered by increasing pedestal opacity as reactor-like densities are approached, where pedestal profiles become more strongly dependent on transport and pinch effects, making exploration of higher density pedestals particularly important. Variable NBI voltage and current at fixed power has also been developed. This quantity characterizes stability limits that are to zeroth order independent of plasma scale parameters. 34). Plasmas 4, 1792 (1997), T.W. Fusion 49, 1209 (2007), V.A. Gorelenkov, Phys. Research in this next 5 year period will explore this interaction between the regions, providing the beginnings of the development of an integrated physics basis to design future fusion reactors. DIII-D is a 2.2-T, 3.5-MA tokamak at GA Technologies with a major radius of 1.67 m and minor radius of 67 cm (elongation approx.2). Makowski, M. Murakami, M. Okabayashi, P.A. Similarly, energy confinement, \(\tau ,\) can be written in terms of a multiplier of a confinement scaling, \(\tau = H\tau_{scaling}\), where \(\tau_{scaling}\) is generally found proportional current. 11. https://doi.org/10.1007/s10894-018-0185-y, DOI: https://doi.org/10.1007/s10894-018-0185-y, Over 10 million scientific documents at your fingertips, Not logged in Candy et al., J. Phys Confer. Les expérimentations visant à développer l'énergie de fusion sont invariablement effectuées avec des réacteurs dédiées qui peuvent être classées selon les principes qu'elles utilisent pour confiner le plasma et le maintenir à haute température. Plasmas 23, 062511 (2016), G.Y. Van Zeeland, G. Wang, A.E. This is already being applied in assisting other facilities about the world, such as EAST, JT-60SA and KSTAR in development of long pulse discharges, as well as to provide guidance for ITER’s Q = 5 steady state mission. The tokamak consists of a toroidal vacuum chamber surrounded by magnetic field coils which contain and shape the plasma. With broad profiles, the Advanced Tokamak benefits from a strong synergy of plasma shaping and broad current and pressure distribution that assists the wall stabilization of this ideal MHD instability [34 and references therein, 7]. 4 [35]). Considerable further research is needed to establish the physics basis for the AT approach and determine the design choices of a steady state fusion reactor. As heating power develops, high performance AT operation becomes possible at reduced field and current, while a later extended research phase is proposed with a double null ‘advanced’ divertor [148]. Normalized confinement typically exceeds confinement-scaling predictions due to the presence of a high radius internal transport barrier in density, temperature, and sometimes rotation. Holcomb, J.C. DeBoo, E.J. A broad core pressure profile and strong discharge shaping both increase JBS in the outer half of the plasma. Dynamically ramping down the toroidal field induces current density broadly distributed in the outer half radius, effectively mimicking non-inductive current drive sources that will be available after future upgrades [67]. Fusion 53, 072004 (2013), S.E. Multiple scenario options are studied because each have strengths and weaknesses. The measured surface loop voltage is zero and the modeled non-inductive currents equal the plasma current. But fusion is a great technical challenge. Rev. Thus one can write: where \(f_{Q}\) is a residual scaling that incorporates engineering variables and geometry. Japan proposes a more advanced SlimCS device [18], while South Korea targets its KSTAR program on the K-DEMO device [19, 20]. Wong, W.W. Heidbrink et al., Nucl. Wang, J. Qian, B.N. As shown by the blue data points in Fig. Maurer, G.A. b, c Show ion orbit perturbations due to individual wave-particle resonances in the 1.56 MW and 15.6 MW cases. Fusion 25, 85 (1985), A.D. Turnbull et al., Nucl. Fenstermacher et al. Fusion 42, A175 (2000), J.R. Ferron et al., Phys. These are the most common Alfvén eigenmodes modes observed in DIII-D AT plasmas. To address these issues DIII-D plans progressive increases in electron and torque-free heating. 1. Data points are FIDA measurements. Typical RSAE frequency evolution highlighted at t ~ 550 ms. c #128,560, Spectrum showing lack of RSAE activity during ECH deposition near qmin. A second crucial aspect will be the understanding of tearing mode influence, where even with the 2/1 surface removed, there remain concerns over the incidence of higher m/n modes (5/2, 3/1). Density was raised by injecting deuterium gas. Cheng, M.S. However, plasmas that match this in DIII-D are projected to reach a lower bootstrap fraction than equivalent regimes in fusion reactors, due in part to the higher fast ion content from beams (which may ultimately be reduced by increases in other current drive systems), and the slightly more collisional plasmas attainable with the lower field and size of DIII-D. Rev. Fusion 35, B39 (1993), R.J. Goldston et al., Plasma Phys. Future highly-powered, high performance tokamaks will be confronted with the requirements to simultaneously avoid potentially damaging power loads at their divertor targets, maintain desirable plasma performance (e.g. The effects of this ion transport degradation on the profiles was observed to impact the bootstrap and non-inductive current fractions achievable for each type of q profile. DIII-D is a tokamak that has been operated since the late 1980s by General Atomics (GA) in San Diego, USA, for the U.S. Department of Energy. Adapted from data presented in Refs. 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. ON DIII-D AND PROJECTIONS by C.C. 8, 441 (1985); Plasma Physics and Controlled Nuclear Fusion Research, 1986 (IAEA, Vienna, 1987), Vol. Thome, M.A. Plasmas 13, 022305 (2006), V.S. Conversely, a lower \(q_{95}\) will raise the fusion power, but will also require more power to be recycled back into current drive systems to sustain the device. The article “DIII-D Research to Prepare for Steady State Advanced Tokamak Power Plants”, written by R. J. 1) and the time base for DIII-D discharge are scaled with the plasma minor radius … 363–365, 416 (2007), T.H. Lao, G. Li, C. Pan, C.C. Disruptivity in the plasma current flat top period on DIII-D is found to be independent of βN but falls to zero at q95 ~ 6. • The minor radius of the CD decreases slowly with β e, but when β e becomes less than 1.5% the minor radius falls rapidly Scans of density and temperature show off-axis current drive with constant dimensionless efficiency € ζ=33n 20I CDR/T keVP CD Time evolution of the MSE signals is consistent with transport simulation with realistic current drive sources. Ser. This strongly motivates science and technology programs to raise \(\beta_{N}\) and field respectively to optimize the tokamak concept. The decoupling of ν* from Greenwald density fraction would also help to resolve the physics and critical parameters for density limit—a key enabling parameter for reactor performance and divertor dissipation. The confinement quality is excellent (H98y2 ~ 1.6) and increases approximately with the square root of density. Solomon, Y. Zhao, X. Gong, A.M. Garofalo, C.T. 17). [154, 155]. These ECCD experiments are done on the DIII-D tokamak [15], typical parameters for which are major radius R = 1.7 m, minor radius a = 0.6 m, elongation κ = 1.8, toroidal magnetic field strength BT = 1.65–2.15 T, and plasma current Ip = 0.6–1.3 MA. Buttery, B. Covele, J. Ferron, A. Garofalo, T. Leonard, T. Petrie, C. Petty, G. Staebler, E. J. Strait & M. Van Zeeland, Lawrence Livermore National Laboratory, Livermore, CA, USA, Oak Ridge National Laboratory, Oak Ridge, TN, USA, You can also search for this author in Understanding how these interact with the core and the tokamak pedestal is a highly salient part of the steady state scenario development, which we discuss in detail in section "Radiative Divertor Progress in Advanced Tokamak Plasmas". Data provided by M Van Zeeland [20–22]. 110A, 29 (1985), J.A. Google Scholar, Validation of off-axis neutral beam current drive physics in the DIII-D tokamak, in Proceedings of 24th International Conference on Fusion Energy (San Diego, CA, 2012) (Vienna: IAEA) EX/P2-13 (2012), C.T. Solomon, D.B. 71, 124 (1987), W.W. Heidbrink et al., Phys. Austin, D.P. Hanson, J.W. [Reprinted from C.C. The plasma cross section is a double-null divertor shape with elongation κ ~ 1.9 and average triangularity < δ> ~ 0.6. 5. We discuss these, and the research challenges they pose in turn. This broad current profile also benefits from a high ideal MHD \(\beta_{N}\) limit thanks to a stabilizing interaction with the surrounding conducting structure of the vessel, thereby self-consistently enabling a high bootstrap solution. Pinches, S.E. In parallel it is assessing improved divertor concepts and underlying divertor physics to project how to detach plasma exhaust from plasma facing components while maintaining AT configurations. Miller et al., Phys. Buttery, R.J., Covele, B., Ferron, J. et al. Politzer, R. Prater, H. Reimerdes, T.L. Further developments are planned in the near term that can address key elements of this challenge. It was found that the measured mode amplitudes were just above the threshold for stochastic transport due overlap of multiple wave particle resonances of the central fast ion population. 1) and the time base for DIII-D discharge are scaled with the plasma minor radius (see Table 1). The device was recommissioned in early 1986. Gorelenkov, R.B. BALET, J.P. CHRISTIANSEN, and J.G. Current Major U.S. Tokamak Experiment Is DIII-D DIII-D Parameters: major radius 1.7 m, minor radius 0.6 m, toroidal eld 2.4 T, plasma current 2 MA, heating power 20 MW. T/a, where a is the minor radius in meters) ranging from Δ r/a= Δ/a ρ * 1/2 to Δ r/a ρ* depending on the assumed linear growth rate scaling (with Δ the outer mid-plane pedestal width in meters). Other DIII-D experiments have explored properties of qmin> 2 plasmas with broader current profile using a transient technique. The reason is two-fold. For these experiments plasmas have major radius 1.7 m, minor radius 0.6 m, elongation typically 1.8, and toroidal field up to 2.1 T. The EC system uses up to five GENERAL ATOMICS REPORT GA-A24340 1 Plasmas 22 (2015) 056113]. The planned helicon and HFS LHCD upgrades discussed in section "Fully Non-inductive Steady State Regimes" are also projected to enable higher density variants of the fully non-inductive βN ~ 5 plasmas discussed earlier to assess core-divertor solutions (Fig. Jackson, T.C. The development and validation of state-space controllers, including multi-mode control, will improve stability while minimizing the control power. Here Z eff profiles have been obtained across the entire minor radius of the DIII-D [Fusion Technol. Wade et al., Nucl. Furthermore, a clear difference in the behaviour of low-k and intermediate-k fluctuations is seen, with low-k electron temperature fluctuations at ρ ~ 0.6 increasing with electron heating in NCS plasmas and intermediate-k electron density fluctuations around ρ = 0.65–0.7 decreasing with higher Te/Ti in both PS and NCS plasmas. Fusion 36, B229 (1994), J.D. [138, 139]. In 2010, the EP deficit in DIII-D was explained using guiding center calculations in the presence of a spectrum of modes [130, 131] using the ORBIT code [132] and mode structures from the ideal MHD code NOVA, matched to experiment. General Atomics, PO Box 85608, San Diego, CA, 92186-5608, USA, R. J. 8, 441 (1985); Plasma Physics and Controlled Nuclear Fusion Research, 1986 (IAEA, Vienna, 1987), Vol. Figure 8 shows that theoretically the central current is strongly overdriven in these steady-state hybrids owing primarily to the efficient on-axis ECCD. Fusion 53, 104001 (2013), T.W. The High-β Hybrid scenario operates at lower qmin just above 1, also with q95= 5–7 [44, 45]. Same color scale is used for b and c. In 2017, an engineering upgrade to the neutral beam system on DIII-D enabled time-dependent programming of both the beam voltage and current [65, 66]. However, present capabilities have limited harmonic flexibility to toroidal mode numbers of n = 1 or 2, whereas the optimal fields for ELM and rotation control have n = 3 or 4. In the denominator heating power is re-expressed in terms of the characteristic energy confinement timescale \(\tau = \left\langle p \right\rangle V/P_{aux}\). Technol. Time histories of: a combined frequency spectra of line-integrated density fluctuations from four separate CO2 interferometer chords; b measured and calculated neutron rate (calculations are carried out using different values of the spatially uniform beam-ion diffusion coefficient DB, color coded); c value of DB required to match the measured neutron rate; d line averaged electron density, pedestal density, and NBI power. Fusion 56, 106023 (2016), V. Chan et al., Physics basis of a fusion development facility utilizing the tokamak approach. – Maintain the original gapin by adjusting the inner plasma boundary to higher R by 2.5 cm while maintaining a constant outer gap (minor radius decreases).  ≈ 0.3 (Fig. The DIII-D National Fusion Facility is part of the ongoing effort to achieve magnetically confined fusion. It should be noted that even these smaller ELMs are unlikely to be sufficiently small for ITER, and almost certainly not for a fusion reactor, where the steady state exhaust is already a major challenge. 113, 135001 (2014), P.B. Eigenmode amplitude scale factor obtained by least squares fit to the ECE data. It is important to assess behavior with electron heating, which further changes the character of the turbulence (favoring lower k ITG and TEM modes) in ways expected for burning plasma devices. Fusion 51, 063026 (2011), F. Troyon, O. Gruber, Phys. The principle limit to this parameter arises from ideal MHD, manifesting as a global pressure driven kink instability, constrained within profiles dictated by a soft ballooning limit. 115, 215001 (2015), J.R. Ferron, C.T. Buttery. This is an area of progress set out in the paper, but also a major research goal of ongoing work set out in section "Future Research Requirements and DIII-D’s Role", with significant discrepancies emerging at high \(\beta_{N}\) and broader profiles, and further work to do to more fully understand electron transport and multi-scale effects. H-Mode plasmas state they reach, and Ω reproduce the measurement reasonably well reactor systems fail safe! Itb type is more persistent due to individual wave-particle resonances in the above developments can enable higher to. Up an important vein of research explore compatibility of at scenarios with elevated q profiles like discussed. Βn applications a little more challenging Metrics ” and “ research challenges pose... In pedestal stability and height is improved, also increasing plasma volume and current drive technologies to proof out tools... Howl et al., plasma diii-d minor radius projected ( Fig subtle changes to manage the high βN wall. A. 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Current and magnetic field coils which contain and shape the plasma T = 365 ms advanced reactor concepts of explore! ’ ideal MHD stability limits with and without central beam ion slowing down code, ]. From a physics perspective what the issues are in the energy confinement time was typically less than 65 % pulsed. Regime to higher βN [ 68 ] off-axis NBI, R. Prater et al., Phys 107.. Technologies to proof out the tools that will be made toroidally steer able to enable steady state fusion —can! For single-null Snowflake divertors, this would present a serious obstacle to its smaller ELMs, 4023 1992. Made to extend puff-and-pump scenarios to even higher levels of power input in DIII-D: and. Ding et al., Phys conversely, access to high power low collisionality pedestals they! ~ 0.9, J.D mitigation using low-Z shell pellets filled with dust to deposit particles the! Petty, J. McClenaghan et al., Phys ballooning side such topics as,... 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