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Actively enhancing plasma radiated power has been used to reduce divertor heat flux in high power, high-performance double null divertor (DND) and near-DND plasmas in DIII-D, while at the same time maintaining acceptable energy confinement. One form of this approach that we focus on here is the radiating mantle. The effectiveness of the radiating mantle was found to depend strongly on the location of electron cyclotron (EC) heating deposition and on the effect that seed impurities have on triggering inimical MHD activity inside the plasma. In addition, we present experimental support for the predictions made by the ELITE boundary stability code (Snyder et al., 2002, [1]) for ways to improve confinement and fueling in these high-performance plasmas. Representative AT operating conditions were H98 ≈ 1.4–1.7, βN ≈ 3–4, and q95 ≈ 4.7–6.5. JEL classification: Divertor Materials (D0500), Impurities (10100), Plasma Properties (P0600), Keywords: DIII-D, Edge pedestal, Energy confinement, Gas injection and fueling impurity, Power deposition

We review progress made on the advanced tokamak path to fusion energy by the DIII-D National Fusion Facility (Luxon et al. in Nucl Fusion 42:614, 2002). The advanced tokamak represents a highly attractive approach for a future steady state fusion power plant. In this concept, there is a natural alignment between high pressure operation, favorable stability and transport properties, and a highly self-driven (‘bootstrap’) plasma current to sustain operation efficiently and without disruptions. Research on DIII-D has identified several promising plasma configurations for fully non-inductive operation with potential applications to a range of future devices, from ITER to nuclear science facilities, to compact or large scale fusion power plants. Significant progress has been made toward realizing these scenarios, with the demonstration of high β access, off-axis current drive techniques, model based profile control, and stability and ELM control in reactor relevant physics regimes. Radiative techniques have also been pioneered to develop improved compatibility with divertor requirements, and simultaneous access to high performance pedestals. 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 Alfvenic and other instabilities. These advances have been partnered with development of a rigorous integrated modeling framework used to interpret and validate individual physics models of the various aspects of plasma behavior, and to guide development of improved regimes and upgrades. These tools are also being used to develop and validate concepts for future reactors directly. Having established these foundations, DIII-D is now undergoing a substantial upgrade to raise power, current drive, electron heating and 3-D field capabilities in order to validate this physics and test conceptual solutions in reactor-relevant physics regimes, with a goal to resolve the key scientific and technology questions to enable a decision on a future steady state fusion power plant.

The impurity-seeded detached divertor is essential for heat exhaust in ITER and other reactor-relevant devices. Dedicated experiments with injection of N2, Ne and Ar have been performed in DIII-D to assess the impact of the different impurities on divertor detachment and confinement. Seeding with N2, Ne and Ar all promote divertor detachment, greatly reducing heat flux near the strike point. The upstream plasma density at the onset of detachment decreases with increasing impurity-puffing flow rates. For all injected impurity species, the confinement and pedestal pressure are correlated with the impurity content and the ratio of separatrix loss power to the l -H transition threshold power. As the divertor plasma approaches detachment, the high-Z impurity seeding tends to degrade the core confinement owing to the increased core radiation. In particular, Ar injection with up to 50% of the injected power radiating in the core cools the pedestal and core plasmas, thus significantly degrading the confinement. As for Ne seeding, medium confinement with H98∼0.8 can be maintained during the detachment phase with the pedestal temperature being reduced by about 50%. In contrast, in the N2 seeded plasmas, radiation is predominately confined in the boundary plasma, which leads to less effect on the confinement and pedestal. In the case of strong N2 gas puffing, the confinement recovers during the detachment, from ∼20% reduction at the onset of the detachment to greater than unity comparable to that before the seeding. The core and pedestal temperatures feature a reduction of 30% from the initial attached phase and remain nearly constant during the detachment phase. The improvement in confinement appears to arise from the increase in pedestal and core density despite the temperature reduction.

Significant challenges to reducing divertor heat flux in highly powered near-double null divertor (DND) hybrid plasmas, while still maintaining both high performance metrics and low enough density for application of RF heating, are identified. For these DNDs on DIII-D, the scaling of the peak heat flux at the outer target (q⊥P) ∝ [PSOL x IP] 0.92 for PSOL= 8−19MW and IP= 1.0–1.4MA, and is consistent with standard ITPA scaling for single-null H-mode plasmas. Two divertor heat flux reduction methods were tested. First, applying the puff-and-pump radiating divertor to DIII-D plasmas may be problematical at high power and H98 (≥ 1.5) due to improvement in confinement time with deuterium gas puffing which can lead to unacceptably high core density under certain conditions. Second, q⊥P for these high performance DNDs was reduced by ≈35% when an open divertor is closed on the common flux side of the outer divertor target (“semi-slot”) but also that heating near the slot opening is a significant source for impurity contamination of the core. PSI-22 keywords: DIII-D, Divertor geometry, Gas injection and fueling, Impurity sources, Power deposition

Experimental results from the National Spherical Torus Experiment (NSTX), a medium-size spherical tokamak with a compact divertor, and DIII-D, a large conventional aspect ratio tokamak, demonstrate that the snowflake (SF) divertor configuration may provide a promising solution for mitigating divertor heat loads and target plate erosion compatible with core H-mode confinement in the future fusion devices, where the standard radiative divertor solution may be inadequate. In NSTX, where the initial high-power SF experiment was performed, the SF divertor was compatible with H-mode confinement, and led to the destabilization of large Edge Localized Modes (ELMs). However, a stable partial detachment of the outer strike point was also achieved where inter-ELM peak heat flux was reduced by factors 3–5, and peak ELM heat flux was reduced by up to 80% (see standard divertor). The DIII-D studies show the SF divertor enables significant power spreading in attached and radiative divertor conditions. Results include: compatibility with the core and pedestal, peak inter-ELM divertor heat flux reduction due to geometry at lower $n_{e}$ , and ELM energy and divertor peak heat flux reduction, especially prominent in radiative $D_{2}$ -seeded SF divertor, and nearly complete power detachment and broader radiated power distribution in the radiative $D_{2}$ -seeded SF divertor at $P_{\text {SOL}}=3-4$ MW. A variety of SF configurations can be supported by the divertor coil set in NSTX Upgrade. Edge transport modeling with the multifluid edge transport code UEDGE shows that the radiative SF divertor can successfully reduce peak divertor heat flux for the projected $P_{\text {SOL}} \simeq 9$ MW case. The radiative SF divertor with carbon impurity provides a wider $n_{e}$ operating window, 50% less argon is needed in the impurity-seeded SF configuration to achieve similar $q_{\text {peak}}$ reduction factors (see standard divertor).

Feedback control of radiated power from the lower divertor P rad, div, L has been implemented in the DIII-D Plasma Control System (PCS). A realtime sensor for P rad, div, L has been constructed from 12 foil bolometer channels which agrees with standard post-shot analysis to within 20%. Results with the 12-channel sensor are compared to initial proof-of-concept tests with a single channel as a proxy for P rad, div, L , showing that the upgraded sensor is necessary to overcome limitations of the proxy channel strategy in DIII-D. Using N 2 seeding under feedback control, P rad, div, L has been increased by up to 150% above unseeded levels, and a radiated power fraction f rad of 80% has been demonstrated, although feedback controlled gas flow is steadier at f rad = 55%. Spatial coverage is broad enough to enable P rad control during the strike point sweeps which are commonly used to generate pseudo-2D divertor Thomson measurements in DIII-D divertor experiments. Use of this control reveals challenges that may affect next step devices, which will require actively controlled extrinsic impurity seeding in order to manage heat loads. When operating at high f rad , changes in pedestal T e (caused by ELMs in these experiments but could come from other disturbances) resulted in large perturbations to P rad which were destabilizing to the feedback controller.

We survey the results of recent DIII-D experiments that tested the effectiveness of three innovative tokamak divertor concepts in reducing divertor heat flux while still maintaining acceptable energy confinement under neon/deuterium-based radiating divertor (RD) conditions: (1) magnetically unbalanced high performance double-null divertor (DND) plasmas, (2) high performance double-null “Snowflake” (SF-DN) plasmas, and (3) single-null H-mode plasmas having different isolation from their divertor targets. In general, all three concepts adapt well to RD conditions, achieving significant reduction in divertor heat flux ( q ⊥ p ) and maintaining high performance metrics, e.g., 50–70% reduction in peak divertor heat flux for DND and SF-DN plasmas that are characterized by β N ≅ 3.0 and H 98(y,2) ≈ 1.35. It is also demonstrated that q ⊥ p could be reduced ≈50% by extending the parallel connection length ( L ||-XPT ) in the scrape-off layer between the X -point and divertor targets over a variety of the RD and non-RD environments tested.

The snowflake divertor (SFD) control and detachment control to manage the heat flux at the divertor are successfully demonstrated at DIII-D. Results of the development and implementation of these two heat flux reduction control methods are presented. The SFD control algorithm calculates the position of the two null-points in real-time and controls shaping coil currents to achieve and stabilize various snowflake configurations. Detachment control stabilizes the detachment front fixed at specified distance between the strike point and the X-point throughout the shot.

Recent DIII-D experiments assessed the snowflake divertor (SF) configuration in a radiative regime in H-mode discharges with D 2 D 2 seeding. The SF configuration was maintained for many energy confinement times (2–3 s) in H-mode discharges ( I p =1.2 I p = 1.2 MA, P NBI =4–5 P NBI = 4 – 5 MW, and B×∇B B × ∇ B down (favorable direction toward the divertor)), and found to be compatible with high performance operation (H98y2 ⩾ 1). The two studied SF configurations, the SF-plus and the SF-minus, have a small finite distance between the primary X-point and the secondary B p B p null located in the private flux region or the common flux region, respectively. In H-mode discharges with the SF configurations (cf. H-mode discharges with the standard divertor with similar conditions) the stored energy lost per the edge localized mode (ELM) was reduced, and significant divertor heat flux reduction between and during ELMs was observed over a range of collisionalities, from lower density conditions toward a higher density H-modes with the radiative SF divertor.

Upgrades to core and divertor Thomson scattering (DTS) diagnostics at DIII-D have provided measurements of electron pressure profiles in the lower divertor from attached- to fully-detached divertor plasma conditions. Detailed, multistep sequences of discharges with increasing line-averaged density were run at several levels of Pinj. Strike point sweeping allowed 2D divertor characterization using DTS optimized to measure Te down to 0.5 eV. The ionization front at the onset of detachment is found to move upwards in a controlled manner consistent with the indication that scrape-off layer parallel power flux is converted from conducted to convective heat transport. Measurements of ne, Te and pe in the divertor versus Lparallel demonstrate a rapid transition from Te ⩾ 15 eV to ⩽3 eV occurring both at the outer strike point and upstream of the X-point. These observations provide a strong benchmark for ongoing modeling of divertor detachment for existing and future tokamak devices.

Joint DIII-D/EAST experiments in the radio-frequency (RF) heated H-mode scheme with comparison to that of neutral beam (NB) heated H-mode scheme were carried out on DIII-D and EAST under similar conditions to examine the effect of heating scheme on scrape-off layer (SOL) width in H-mode plasmas for application to ITER. A dimensionally similar plasma equilibrium was used to match the EAST shape parameters. The divertor heat flux and SOL widths were measured with infra-red camera in DIII-D, while with divertor Langmuir probe array in EAST. It has been demonstrated on both DIII-D and EAST that RF-heated plasma has a broader SOL than NB-heated plasma when the edge electrons are effectively heated in low plasma current and low density regime with low edge collisionality. Detailed edge and pedestal profile analysis on DIII-D suggests that the low edge collisionality and ion orbit loss effect may account for the observed broadening. The joint experiment in DIII-D has also demonstrated the strong inverse dependence of SOL width on the plasma current in electron cyclotron heated (ECH) H-mode plasmas. Keywords: SOL width, Radio-frequency heating, Neutral beam heating, Low collisionality

The ‘Super H-Mode’ regime is predicted to enable pedestal height and fusion performance substantially higher than standard H-Mode operation. This regime exists due to a bifurcation of the pedestal pressure, as a function of density, that is predicted by the EPED model to occur in strongly shaped plasmas above a critical pedestal density. Experiments on Alcator C-Mod and DIII-D have achieved access to the Super H-Mode (and Near Super H) regime, and obtained very high pedestal pressure, including the highest achieved on a tokamak (p ped ~ 80 kPa) in C-Mod experiments operating near the ITER magnetic field. DIII-D Super H experiments have demonstrated strong performance, including the highest stored energy in the present configuration of DIII-D (W ~ 2.2–3.2 MJ), while utilizing only about half of the available heating power (P heat ~ 7–12 MW). These DIII-D experiments have obtained the highest value of peak fusion gain, Q DT,equiv ~ 0.5, achieved on a medium scale (R β N ~ 2.9, H98 ~ 1.6) has been achieved utilizing n = 3 magnetic perturbations for density and impurity control. Pedestal and global confinement has been maintained in the presence of deuterium and nitrogen gas puffing, which enables a more radiative divertor condition. A pair of simple performance metrics is developed to assess and compare regimes. Super H-Mode access is predicted for ITER and expected, based on both theoretical prediction and observed normalized performance, to allow ITER to achieve its goals (Q = 10) at I p

We report on recent experiments on DIII-D that examined the effects that variations in the parallel connection length in the scrape-off layer (SOL), L||, and the radial location of the outer divertor target, RTAR, have on divertor plasma properties. Two-point modeling of the SOL plasma predicts that larger values of L|| and RTAR should lower temperature and raise density at the outer divertor target for fixed upstream separatrix density and temperature, i.e., nTAR ∝ [RTAR]2[L||]6/7 and TTAR ∝ [RTAR]−2[L||]−4/7. The dependence of nTAR and TTAR on L|| was consistent with our data, but the dependence of nTAR and TTAR on RTAR was not. The surprising result that the divertor plasma parameters did not depend on RTAR in the predicted way may be due to convected heat flux, driven by escaping neutrals, in the more open configuration of the larger RTAR cases. Modeling results using the SOLPS code support this postulate.

A major challenge facing the design and operation of next-step high-power steady-state fusion devices is to develop a viable divertor solution with order-of-magnitude increases in power handling capability relative to present experience, while having acceptable divertor target plate erosion and being compatible with maintaining good core plasma confinement. A new initiative has been launched on DIII-D to develop the scientific basis for design, installation, and operation of an advanced divertor to evaluate boundary plasma solutions applicable to next step fusion experiments beyond ITER. Developing the scientific basis for fusion reactor divertor solutions must necessarily follow three lines of research, which we plan to pursue in DIII-D: (1) Advance scientific understanding and predictive capability through development and comparison between state-of-the art computational models and enhanced measurements using targeted parametric scans; (2) Develop and validate key divertor design concepts and codes through innovative variations in physical structure and magnetic geometry; (3) Assess candidate materials, determining the implications for core plasma operation and control, and develop mitigation techniques for any deleterious effects, incorporating development of plasma-material interaction models. These efforts will lead to design, installation, and evaluation of an advanced divertor for DIII-D to enable highly dissipative divertor operation at core density (n e/n GW), neutral fueling and impurity influx most compatible with high performance plasma scenarios and reactor relevant plasma facing components (PFCs). This paper highlights the current progress and near-term strategies of boundary/PMI research on DIII-D.

To move to a fusion DEMO power plant after ITER, a Fusion Nuclear Science Facility (FNSF) is needed in addition to ITER and research in operating tokamaks and those under construction. The FNSF wil...

The article “DIII-D Research to Prepare for Steady State Advanced Tokamak Power Plants”, written by R. J. Buttery, B. Covele, J. Ferron, A. Garofalo, C. T. Holcomb, T. Leonard, J. M. Park, T. Petrie, C. Petty, G. Staebler, E. J. Strait, M. Van Zeeland, was originally published electronically on the publisher’s Internet portal ( https://doi.org/10.1007/s10894-018-0185-y ) on 3 September 2018 without open access.

Simultaneous control of two proximate magnetic field nulls in the divertor region is demonstrated on DIII–D to enable plasma operations in an advanced magnetic configuration known as the snowflake divertor (SFD). The SFD is characterized by a second-order poloidal field null, created by merging two first-order nulls of the standard divertor configuration. The snowflake configuration has many magnetic properties, such as high poloidal flux expansion, large plasma-wetted area, and additional strike points, that are advantageous for divertor heat flux management in future fusion reactors. However, the magnetic configuration of the SFD is highly-sensitive to changes in currents within the plasma and external coils and therefore requires complex magnetic control. The first real-time snowflake detection and control system on DIII–D has been implemented in order to stabilize the configuration. The control algorithm calculates the position of the two nulls in real-time by locally-expanding the Grad–Shafranov equation in the divertor region. A linear relation between variations in the poloidal field coil currents and changes in the null locations is then analytically derived. This formulation allows for simultaneous control of multiple coils to achieve a desired SFD configuration. It is shown that the control enabled various snowflake configurations on DIII–D in scenarios such as the double-null advanced tokamak. The SFD resulted in a 2.5× reduction in the peak heat flux for many energy confinement times (2–3 s) without any adverse effects on core plasma performance.

The objective of the Fusion Development Facility (FDF) under consideration is to carry forward advanced tokamak physics for optimization of fusion reactors and enable development of fusion’s energy...

Recent DIII-D experiments show that both magnetic balance and particle drifts are important to understanding how argon impurities accumulate in balanced and unbalanced double-null plasmas during ‘puff and pump’ radiating divertor operation. Unbalanced double-null shapes, which are biased in the direction opposite to the ion B × ∇ B drift direction, have produced the best result to-date in terms of coupling a radiating divertor approach with an H-mode plasma. The proximity to balanced double-null that can be used for puff-and-pump may depend on the width of the heat flux profile in the scrape-off layer. A comparison of plasma behavior in the open lower divertor to that of the more closed upper divertor determined that plasma density control was far more sensitive to the ion B × ∇ B drift direction than to the relative closure of the divertor.

Resonant magnetic perturbations (RMPs) have been shown to successfully suppress edge localized modes (ELMs) in the DIII-D tokamak. A previous study of target plate conditions highlighted differences in RMP discharges between low and high electron collisionality, ν* e , operation in DIIIl-D. This paper reports on a systematic study of the electron density pump-out associated with the turn-on of the RMP over a wide range of operating conditions in DIII-D, including shapes and collisionalities similar to those anticipated in ITER. It is shown that the pump-out magnitude, Δn e , has an upper envelope that is inversely proportional to the pedestal ν* e, The particle decay times, which are calculated based on global D 2 particle balances, show an increase as the pedestal ν* e, is increased. Both results are suggestive that the underlying physics mechanism is an increase in edge particle transport and/or that wall depletion is playing a role in the pump-out magnitude.

An improved, self-contained gas injection system for the divertor material evaluation system (DIMES) on DIII-D has been employed for in situ study of chemical erosion in the tokamak divertor environment. To minimize perturbation to local plasma, the Mark II porous plug injector (PPI) releases methane through a porous graphite surface at the outer strike point at a rate precisely controlled by a micro-orifice flow restrictor to be approximately equal as that predicted for intrinsic chemical sputtering. Effective photon efficiencies resulting from CH(4) are found to be 58 +/- 12 in an attached divertor (n(e) similar to 1.5 x 10(13)/cm(3), T(e) similar to 25 eV, T(surf)similar to 450 K), and 94 +/- 20 in a semi-detached cold divertor (n(e) similar to 6.0 x 10(13)/cm(3), T(e) similar to 2-3 eV, T(surf) similar to 350 K). These values are significantly more than previous measurements in similar plasma conditions, indicating the importance of the injection rate and local re-erosion for the integrity of this analysis. The contribution of chemical versus physical sputtering to the source of C(+) at the target is assessed through simultaneous measurement of CII line, and CD plus CH-band emissions during release of CH(4) from the Pill, then compared with thatmore » seen in intrinsic sputtering. (C) 2009 Elsevier B.V. All rights reserved.« less

A radiating divertor approach was successfully applied to high performance 'hybrid' plasmas [M.R. Wade, et al., Proc. 20th IAEA Fusion Energy Conf., Vilamoura, (2004)]. Our technique included: (1) injecting argon near the outer divertor target, (2) enhancing the plasma flow into the inner and outer divertors by a combination of particle pumping and deuterium gas puffing upstream of the divertor targets, and (3) isolating the inner divertor from the outer by a structure in the private flux region. Good hybrid conditions were maintained, as the peak heat flux at the outer divertor target was reduced by a factor of 2.5; the peak heat flux at the inner target decreased by 20%. This difference was caused by a higher concentration of argon at the outer target than at the inner target. Argon accumulation in the main plasma was modest (n{sub AR}/n{sub e} {le}0.004 on axis), although the argon profile was more peaked than the electron profile.

Experimental results from the National Spherical Torus Experiment (NSTX), a medium-size spherical tokamak with a compact divertor, and DIII-D, a large conventional aspect ratio tokamak, demonstrate that the snowflake (SF) divertor configuration (D.D. Ryutov, Phys. Plasmas, 14, 064502, 2007) may provide a promising solution for mitigating divertor heat loads and target plate erosion compatible with core H-mode confinement in future fusion devices, where the standard radiative divertor solution may be inadequate. In NSTX, where the initial high-power SF experiment were performed, the SF divertor was compatible with H-mode confinement, and led to the destabilization of large ELMs. However, a stable partial detachment of the outer strike point was also achieved where inter-ELM peak heat flux was reduced by factors 3–5, and peak ELM heat flux was reduced by up to 80% (cf. standard divertor). The DIII-D studies show the SF divertor enables significant power spreading in attached and radiative divertor conditions. Results include: compatibility with the core and pedestal, peak inter-ELM divertor heat flux reduction due to geometry at lower ne, and ELM energy and divertor peak heat flux reduction, especially prominent in radiative D2-seeded SF divertor, and nearly complete power detachment and broader radiated power distribution in the radiative D2-seeded SF divertor at PSOL = 3 – 4 MW. A variety of SF configurations can be supported by the divertor coil set in NSTX Upgrade. Edge transport modeling with the multi-fluid edge transport code UEDGE shows that the radiative SF divertor can successfully reduce peak divertor heat flux for the projected PSOL ≃ 9 MW case. The radiative SF divertor with carbon impurity provides a wider ne operating window, 50% less argon is needed in the impurity-seeded SF configuration to achieve similar qpeak reduction factors (cf. standard divertor).

The advanced tokamak is considered as the basis for a fusion power plant. The ARIES-AT design has an aspect ratio of A ≡ R / a = 4.0 , an elongation and triangularity of κ = 2.20 , δ = 0.90 (evaluated at the separatrix surface), a toroidal beta of β = 9.1 % (normalized to the vacuum toroidal field at the plasma center), which corresponds to a normalized beta of β N ≡ 100 × β / ( I P ( M A ) / a ( m ) B ( T ) ) = 5.4 . These beta values are chosen to be 10% below the ideal MHD stability limit. The bootstrap-current fraction is f BS ≡ I BS / I P = 0.91 . This leads to a design with total plasma current I P = 12.8 MA, and toroidal field of 11.1 T (at the coil edge) and 5.8 T (at the plasma center). The major and minor radii are 5.2 and 1.3 m. The effects of H-mode edge gradients and the stability of this configuration to non-ideal modes is analyzed. The current drive system consists of ICRF/FW for on-axis current drive and a Lower Hybrid system for off-axis. Transport projections are presented using the drift-wave based GLF23 model. The approach to power and particle exhaust using both plasma core and scrape-off-layer radiation is presented.

The ARIES-AT study was initiated to assess the potential of high-performance tokamak plasmas together with advanced technology in a fusion power plant and to identifying physics and technology areas with the highest leverage for achieving attractive and competitive fusion power in order to guide fusion R&D. The 1000-MWe ARIES-AT design has a major radius of 5.2 m, a minor radius of 1.3 m, a toroidal β of 9.2% (βN = 5.4) and an on-axis field of 5.6 T. The plasma current is 13 MA and the current-drive power is 35 MW. The ARIES-AT design uses the same physics basis as ARIES-RS, a reversed-shear plasma. A distinct difference between ARIES-RS and ARIES-AT plasmas is the higher plasma elongation of ARIES-AT (κx = 2.2) which is the result of a “thinner” blanket leading to a large increase in plasma β to 9.2% (compared to 5% for ARIES-RS) with only a slightly higher βN. ARIES-AT blanket is a simple, low-pressure design consisting of SiC composite boxes with a SiC insert for flow distribution that does not carry any structural load. The breeding coolant (Pb–17Li) enters the fusion core from the bottom, and cools the first wall while traveling in the poloidal direction to the top of the blanket module. The coolant then

Recent experiments on DIII-D point to the importance of two factors in determining how effectively the deuterium particle inventory in a tokamak plasma can be controlled through pumping at the divertor target(s): (1) the divertor magnetic balance, i.e. the degree to which the divertor topology is single-null or double-null (DN) and (2) the direction of the of B × ∇B ion drift with respect to the X-point(s). Changes in divertor magnetic balance near the DN shape have a much stronger effect on the particle exhaust rate at the inner divertor target(s) than on the particle exhaust rate at the outer divertor target(s). The particle exhaust rate for the DN shape is strongest at the outer strike point opposite the B × ∇B ion particle drift direction. Our data suggests that the presence of B × ∇B and E × B ion particle drifts in the scrape-off layer and divertor(s) play an important role in the particle exhaust rates of DN and near-DN plasmas. Particle exhaust rates are shown to depend strongly on the edge (pedestal) density. These results have implications for particle control in ITER and other future tokamaks.

Measurements and modeling of the 2D poloidal D α intensity distribution in DIII-D low density L-mode and medium density ELMy H-mode plasmas indicate that the core plasma is predominately fueled near the divertor X-point region. The neutral deuterium and ion carbon emission were measured in the divertor and inner main chamber scrape-off layer (SOL) using a plasma imaging technique, covering 85% of the poloidal cross-section. Typically, the peak emission in the inner main SOL at the tokamak midplane was three orders of magnitude lower than in the divertor. For discharges with the ion B × ∇ B drift direction toward the lower divertor the UEDGE/DEGAS codes predict strong core plasma fueling from the significantly higher density and lower temperature plasma calculated in the inner divertor leg. The concomitant carbon ion flow reversal in the inner divertor leg enhances the leakage of carbon from the divertor into the main SOL, and hence into the core.

Recent experiments on DIII-D point to the importance of two factors in determining the rate at which deuterium particles can be pumped at the divertor target(s): (1) the divertor magnetic balance, i.e., the degree to which the divertor topology is single-null (SN) or double-null (DN), and (2) the direction of the of B x {del}B ion drift with respect to the X-point(s). Changes in divertor magnetic balance near the DN shape have a much stronger effect on the pumping rate at the inner divertor target(s) than on the pumping rate at the outer divertor target(s). The behavior in the particle pumping observed at the inner and outer divertor target(s) in the DN and near-DN shapes suggests a redistribution of particles that would be expected in the presence of B x {del}B and E x B ion particle drifts in the scrapeoff layer (SOL) and divertor(s).

Edge localized mode (ELM) effects in the DIII-D pedestal and boundary plasmas were measured with multiple fast diagnostics in matched, lower single-null, ELMing H-mode discharges with the ion B × ∇B drift either toward or away from the divertor. Data show a strong dependence of the delay in inner vs. outer divertor ELM Dα on drift direction, and a weaker drift dependence of the inner vs. outer delay of Prad, in addition to the strong density dependence seen in previous work. Time dependent boundary plasma modeling during an ELM was done with the UEDGE code including a six-species fluid carbon model and the effect of B-field induced particle drifts. The ELM was modeled as an instantaneous, outer midplane peaked, increase of diffusion coefficients from the top of the pedestal to the SOL. The simulations show delays in the inner vs. outer divertor ELM transients that are similar to the measured delays at moderate density.

Key DIII-D advanced tokamak (AT) experimental and modelling results are applied to examine the physics and control issues for ITER to operate in a negative central shear (NCS) AT scenario. The effects of a finite edge pressure pedestal and current density are included based on the DIII-D experimental profiles. Ideal and resistive stability analyses demonstrate that feedback control of resistive wall modes by rotational drive or flux conserving intelligent coils is crucial for these AT configurations to operate at attractive βN values in the range 3.0–3.5. Vertical stability and halo current analyses show that reliable disruption mitigation is essential and mitigation control using an impurity gas can significantly reduce the local mechanical stress to an acceptable level. Core transport and turbulence analyses indicate that control of the rotational shear profile is essential to reduce the pedestal temperature required for high β. Consideration of edge stability and core transport suggests that a sufficiently wide pedestal is necessary for the projected fusion performance. Heat flux analyses indicate that, with core-only radiation enhancement, the outboard peak divertor heat load is near the design limit of 10 MW m−2. Detached operation may be necessary to reduce the heat flux to a more manageable level. Evaluation of the ITER pulse length using a local step response approach indicates that the 3000 s ITER long-pulse scenario is probably both necessary and sufficient for demonstration of local current profile control.

The reduction in size of Type I edge localized modes (ELMs) with increasing density is explored in DIII-D [J. L. Luxon, Nucl. Fusion 42, 614 (2002)] for the purpose of studying the underlying transport of ELM energy. The separate convective and conductive transport of energy due to an ELM is determined by Thomson scattering measurements of electron density and temperature in the pedestal. The conductive transport from the pedestal during an ELM decreases with increasing density, while the convective transport remains nearly constant. The scaling of the ELM energy loss is compared with an edge stability model. The role of the divertor sheath in limiting energy loss from the pedestal during an ELM is explored. Evidence of outward radial transport to the midplane wall during an ELM is also presented.

Changes in the divertor magnetic balance in DIII-D H-mode plasmas affects core, edge, and divertor plasma behavior. Both the pedestal density ne,PED and plasma stored energy WT were sensitive to changes in magnetic balance near the double-null (DN) configuration, e.g., both decreased 20–30% when the DN shifted to a slightly unbalanced DN, where the B ×∇B drift direction pointed away from the main X-point. Recycling at each of the four divertor targets was sensitive to changes in magnetic balance and the B ×∇B drift direction. The poloidal distribution of the recycling in DN is in qualitative agreement with the predictions of UEDGE modeling with particle drifts included. The particle flux at the inner divertor target is shown to be much more sensitive to magnetic balance than the particle flux at the outer divertor target near the DN shape. These results suggest possible advantages and drawbacks for balanced DN operation.

Measurements from balanced or slightly unbalanced, double-null, ELMing H-mode plasmas show that the upper, more closed, divertor strike point particle fluxes are larger than the lower, more open, divertor strike point particle fluxes for comparable conditions. Comparison of the upper and lower outer strike point peak particle fluxes measured with Langmuir probes during scans of the magnetic balance shows an asymmetry around the magnetic balance point which is consistent with both the enhanced upper recycling and the expected direction of private region E × B drifts. Comparison of the inner and outer peak particle fluxes in the upper divertor shows a dependence on magnetic field direction that is also consistent with the private region E × B drift direction.

We report on DIII–D data that reveal the underlying processes responsible for transport of energy and particles from the edge pedestal to the divertor target during Type I edge-localized modes (ELMs). The separate convective and conductive transport of energy due to an ELM is determined by Thomson scattering measurements of electron density and temperature in the pedestal. Conductive transport is measured as a drop in pedestal temperature and decreases with increasing density. The convective transport of energy, measured as a loss of density from the pedestal, however, remains constant as a function of density. From the SOL, ELM energy is quickly carried to the divertor target. An expected sheath limit to the ELM heat flux set by the slower arrival of pedestal ions is overcome by additional ionization of neutrals generated from the divertor target as evidenced by a fast, ∼100 μs, rise in divertor density. A large in/out asymmetry of the divertor ELM heat flux is observed at high density, but becomes nearly symmetric at low density.

The 2-D fluid code UEDGE was used to analyze DIII-D experiments to determine the role of neutrals in core fueling, core impurities, and also the H-mode pedestal structure. We compared the effects of divertor closure on the fueling rate and impurity density of high-triangularity, H-mode plasmas. UEDGE simulations indicate that the decrease in both deuterium core fueling (≈15–20%) and core carbon density (≈15–30%) with the closed divertor compared to the open divertor configuration is due to greater divertor screening of neutrals. We also compared UEDGE results with a simple analytic model of the H-mode pedestal structure [Nucl. Fusion 42 (2002) 52]. The model predicts both the width and gradient of the transport barrier in n e as a function of the pedestal density. The more sophisticated UEDGE simulations of H-mode discharges corroborate the simple analytic model, which is consistent with the hypothesis that fueling processes play a role in H-mode transport barrier formation.

Recently, both Alcator C-Mod and DIII-D tokamaks observed fast anomalous cross-field plasma transport in the SOL. These observations supported by theoretical studies have indicated that such transport is essentially non-diffusive and intermittent. We present the results of macroscopic plasma simulations with the UEDGE transport code including anomalous cross-field outward-directed convection for a series of low-power DIII-D shots. Our analysis of experimental and simulation data supports the view that non-diffusive transport is important for the entire edge plasma behavior in both L- and H-mode discharges.

High density, gas fueled, high confinement mode, DIII-D [Luxon, Anderson, Batty et al., Plasma Physics and Controlled Nuclear Fusion Research 1986 (International Atomic Energy Agency, Vienna, 1987), Vol. 1, p. 159] shots that were limited in density buildup by: (1) an H–L (high-to-low confinement mode) transition apparently triggered by a core X-point MARFE (multifaceted asymmetric radiation from edge), following the thermal instability sequence of detachment-divertor MARFE-core X-point MARFE; (2) a density limit apparently triggered by large scale core magnetohydrodynamic activity; and (3) an H–L transition apparently triggered by excessive radiated power from the plasma core are analyzed. The presence or absence of a MARFE was correctly predicted by thermal instability theory. A sharp edge temperature gradient was identified as an important factor inhibiting the usual detachment-MARFE sequence in DIII-D and thus leading to the achievement of higher density.

The energy lost from the pedestal region due to an average ELM in DIII-D is determined from changes to the electron density and temperature profiles as measured by Thomson scattering. The ELM energy due to loss of temperature in the pedestal is associated with conduction and is found to decrease with increasing pedestal density. The ELM energy from lost pedestal density, or convective transport, remains constant as a function of density. The scaling of the two transport channels, conduction and convection, are examined in terms of parallel transport processes in the scrape-off-layer and divertor.

Recent experiments on DIII-D [J. L. Luxon, F. Batty, C. Baxi et al., Plasma Physics and Controlled Nuclear Fusion Research 1986 (International Atomic Energy Agency, Vienna, 1987), Vol. I, p. 159] examined the effect of different operating conditions (“open” and “closed” divertor geometry, active pumping, fueling location) on the maximum achievable density in gas fueled H-mode (high confinement mode) discharges. Several phenomena observed at these higher densities (≈0.8 the Greenwald density)—degradation in energy confinement, detachment of the core plasma from the divertor plate, multifaceted asymmetric radiation from edge formation—are found to be correlated with the predicted onset of various thermal instabilities in the plasma edge or divertor regions. The similarity of the maximum achievable densities under the different operating conditions can be related to a similarity of edge thermal instability characteristics.

Densities of up to 40% above the Greenwald limit are reproducibly achieved in high confinement (HITER89P = 2) ELMing H mode discharges. Simultaneous gas fuelling and divertor pumping were used to obtain these results. Confinement of these discharges, similar to moderate density H mode, is characterized by a stiff temperature profile, and is therefore sensitive to the density profile. A particle transport model is presented that explains the roles of divertor pumping and geometry for access to high densities. The energy loss per ELM at high density is a factor of five lower than the predictions of an earlier scaling, based on data from lower density discharges.

DIII-D has recently demonstrated improved energy confinement by injecting neutral deuterium gas into high performance near-double null divertor (DND) plasmas during high power operation. Representative parameters for these plasmas are: q 95 = 6, P IN up to 15 MW, H 98 = 1.4–1.8, and β N = 2.5–4.0. The ion B × B direction is away from the primary X-point. While plasma conditions at lower to moderate power input (e.g. ≤11 MW) are shown to be favorable to successful puff-and-pump radiating divertor applications, particularly when using argon seeds, plasma behavior at higher powers (e.g. ≥14 MW) may make successful puff-and-pump operation more problematic. In contrast to lower powered high performance plasmas, both τ E and β N in the high power cases (≥14 MW) increased and ELM frequency decreased, as density was raised by deuterium gas injection. Improved performance in the higher power plasmas was tied to higher pedestal pressure, which according to peeling-ballooning mode stability analysis using the ELITE code could increase with density along the kink/peeling stability threshold, while the pedestal pressure gradient in the lower power discharges were limited by the ballooning threshold. This resulted in improved fueling efficiency and ≈10% higher τ E and β N than is normally observed in comparable high performance plasmas on DIII-D. Applying the puff-and-pump radiating divertor approach at moderate versus high power input is shown to result in a much different evolution in core and pedestal plasma behavior. We find that injecting deuterium gas into these highly powered DND plasmas may open up a new avenue for achieving elevated plasma performance, including better fueling, but the resulting higher density may also complicate application of a radiating divertor approach to heat flux reduction in present-day tokamaks, if scenarios involving second-harmonic electron cyclotron heating are used.