IOP PUBLISHING and INTERNATIONAL ATOMIC ENERGY AGENCY NUCLEAR FUSION

Nucl. Fusion 51 (2011) 094009 (17pp) doi:10.1088/0029-5515/51/9/094009

DIII-D contributions towards the scientificbasis for sustained burning plasmasC.M. Greenfield and the DIII-D Team

General Atomics, PO Box 85608, San Diego, CA 92186-5608, USA

Received 5 February 2011, accepted for publication 18 April 2011Published 31 August 2011Online at stacks.iop.org/NF/51/094009

AbstractDIII-D is making significant contributions to a scientific basis for sustained burning plasma operation. Theseinclude explorations of increasingly reactor-relevant scenarios, studies of key issues for projecting performance,development of techniques for handling heat and particle efflux, and assessment of key issues for the ITER researchplan. Advanced scenarios are being optimized in DIII-D via experiments to empirically determine the relationshipbetween transport and the current profile, which in turn can provide essential input to inform improvement of thetheory-based models that do not currently capture the observed behaviour. Joint DIII-D/JET ρ∗ scans in the hybridregime imply Bohm-like confinement scaling. Startup and shutdown techniques were developed for the restrictiveenvironment of future devices while retaining compatibility with advanced scenarios. Towards the goal of a fullypredictive capability, the DIII-D program emphasizes validation of physics-based models, facilitated by a numberof new and upgraded diagnostics. Specific areas include transport, rotation, energetic particles and the H-modepedestal, but this approach permeates the entire research programme. Concerns for heat and particle efflux infuture devices are addressed through studies of ELM control, disruption avoidance and mitigation, and hydrogenicretention in DIII-D’s carbon wall. DIII-D continues to respond to specific needs for ITER. Recent studies havecompared H-mode access in several different ion species, identifying not only isotopic, but density, rotation andgeometrical dependences that may guide access to H-mode during ITER’s non-activated early operation. DIII-D usedan insertable module to simulate the magnetic perturbations introduced by one of ITER’s three test blanket modulesets, demonstrating that little impact on performance is seen at ITER equivalent levels of magnetic perturbation.

(Some figures in this article are in colour only in the electronic version)

1. Introduction

DIII-D research is contributing to the scientific basis forthe optimization of the tokamak approach to fusion energyproduction. Ensuring the success of ITER remains a majorpriority of the programme, as its needs evolve from specificshort-term design issues towards longer term concerns forITER research planning. For longer term success of ITER,and indeed of future devices up to and including a power plant,DIII-D research works towards scientific understanding acrossa range of research areas. The results of these endeavours willbe embodied in reliable predictive tools that can be appliedtowards design of devices and experiments.

In this paper, we will summarize recent researchcontributing to these goals. DIII-D research continues toadvance development, characterization and access to operatingscenarios of interest and importance for burning plasmas(section 2). Relying on DIII-D’s comprehensive set of controltools and diagnostics, a vigorous research programme isunderway to carry out specific tests of physics-based modelsof transport, rotation, energetic particle (EP) effects and theH-mode pedestal (section 3). Methods for handling heat

and particle efflux, including transients (edge localized modes(ELMs) and disruptions) and tritium retention, are describedin section 4. DIII-D continues to be responsive to the specificneeds of ITER. Two of these needs, isotope dependence of theL–H threshold and the plasma response to test blanket modules(TBMs), are described in section 5. In section 6, we will brieflydiscuss new capabilities being added to DIII-D and the futureresearch directions that will be enabled.

2. Scenario development, characterization andaccess

2.1. High-beta steady-state scenario optimization

Steady-state tokamak operation requires that 100% of theplasma current be driven non-inductively (fNI = 1). Thiscurrent is supplied in part by a combination of externallyapplied sources. In DIII-D, these often include electroncyclotron current drive (ECCD) [1] and neutral beam currentdrive (NBCD) [2]. It is desirable to minimize the requirementfor external current drive for reasons of cost and recirculatingpower. Fortunately, high-performance fusion plasmas can

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Nucl. Fusion 51 (2011) 094009 C.M. Greenfield et al

1.0 1.5qmin

Thermal pressurepeaking factor fp

2.0(a)2

3

4 βN=2.8

q95=4.5-6.8

βN=3.1-3.8

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i (ρ=0.7)

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m2/s

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4

5

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m2/s

(b) (c)

Figure 1. (a) The thermal pressure peaking factor in steady-state scenario discharges decreases with increasing qmin and βN. The circularsymbols are from a set of discharges with NBI controlled to maintain βN ≈ 2.8. The discharges shown with square symbols are heated bythe maximum available power PNBI = 13.5 MW, with the resulting βN depending on confinement. The observed behaviour is a consequenceof variations in ion (b) and electron (c) thermal diffusivity, whose variation differs from model-based predictions.

provide a large fraction of the current through self-drivenbootstrap current [3]. In a global sense, the fraction of currentdriven by bootstrap increases as

fBS ∝ qβNfP, (1)

where q is the plasma safety factor, βN is the normalizedplasma beta and fP = (ne(0)Te(0) + ni(0)Ti(0))/⟨neTe + niTi⟩is the thermal pressure peaking factor. The apparent simplicityof this relation, however, is deceiving, since fP is stronglycoupled to both q and βN via transport behaviour that is notfully captured by existing transport models.

To experimentally optimize the q profile for dischargesin DIII-D with 100% of the current driven non-inductively(fNI = 1), the self-consistent response of the plasma profilesto changes in the q profile was studied in high-fNI, high-βN discharges [4, 5]. In discharges where the heating power(neutral beams) is regulated to maintain βN = 2.8 (blue circlesin figure 1(a)), the pressure profile peaking decreases (profileflattens) with increasing qmin. Also shown (red squares) arepoints from a set of discharges where the heating power wasfixed at PNBI = 13.5 MW, so that βN varied between 3.1 and3.8. None of these discharges reached an MHD stability limit;they were all limited in performance by transport. In thesedischarges, the variation of profile shape with qmin was absent,and fP remained at a lower value than seen at βN = 2.8.Referring back to equation (1), we conclude as follows:

• at fixed βN, fP decreases with increasing qmin;• at fixed qmin, fP decreases with increasing βN.

As might be expected from this behaviour, thermaltransport varies significantly as the current profile is varied.Several observations can be made from ion (figure 1(b)) andelectron (figure 1(c)) thermal diffusivities, determined frompower balance analysis of this dataset. First, the transportin both channels varies strongly with the current profile.Second, the electrons and ions do not respond in the same way.Modelling of these cases with a trapped gyro-Landau fluid code(TGLF) [6] show that variations in the linear growth rate donot correlate well with the observed transport changes [7].

The broadening of the pressure profile at high β results ina bootstrap current density profile that is relatively flat betweenthe axis and the H-mode pedestal. Therefore, significantexternally driven current is required in the region inside theH-mode pedestal to match the profiles of the non-inductivecurrent density (JNI) to the desired total current density (J ).In these experiments, the profiles of JNI and J were mostsimilar at qmin ≈ 1.35–1.65, q95 ≈ 6.8, where fBS is alsomaximum, establishing this q profile as the most optimalchoice for fNI = 1 operation in DIII-D with the existing set ofexternal current drive sources.

These experiments focused on establishing the desiredconditions for a sustained scenario, but not on actualsustainment. This will require replacement of inductivelydriven current, driven near the half-radius [8]. In the presentexperiments, this current is primarily provided by ECCD.Modifications now underway will make off-axis NBCD andadditional ECCD available for future experiments.

Plasmas in the steady-state scenario often have βN abovethe no-wall stability limit. Recent research has identifiedkinetic effects as important for resistive wall mode (RWM)stability in this regime, even in non-rotating plasmas [9, 10].Additional progress has also been made in simultaneous activecontrol of error fields and RWMs [11].

2.2. Confinement in advanced inductive scenarios

Advanced inductive (AI) plasmas are a realization of the ITERhybrid scenario, providing high neutron fluence in a prolongedinductive discharge [12]. The performance improvement overconventional (ITER baseline) H-mode is usually ascribed to abroader current profile, but a complete physics basis for hybridscenarios is a subject of continued study. This results in someuncertainty in extrapolation to ITER. DIII-D and JET [13]have recently performed a joint experiment to determine thedependence of confinement on several key parameters [14].The largest extrapolation to ITER is in its size, expressed indimensionless quantities as ρ∗, the normalized ion gyroradiusevaluated at normalized radius ρ = 0.5. In this experiment,the two devices spanned a range of 2.7 in ρ∗, considerablylarger than is possible in any single device and about equal

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Nucl. Fusion 51 (2011) 094009 C.M. Greenfield et al

Mν*

0.0 0.2 0.4ρ0.6 0.8 1.0

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Matched pulses

0.1

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B τE

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1.5

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0.5

H98y,2

Figure 2. (a)–(d) ‘Dimensionally identical’ AI discharges in DIII-D and JET, which are controlled to match normalized gyroradius ρ∗,collisionality ν∗, β and Mach number M in a global sense only, exhibit well-matched profiles of the same over most of the radial extent ofthe plasma. This indicates that the same physics governs the AI discharges in each device. (e) Normalized energy confinement BτE versusρ∗ (closed symbols). The dependence on qcyl and β, from the ITER98 (y, 2) scaling, is included to account for small residual variations in qand β in the data. H98y,2 is shown by open symbols.

to the difference between JET and ITER. To carry out theseexperiments, plasmas with the same shape and aspect ratiowere produced in both tokamaks. Linear dimensions of theplasma in JET were larger by a factor of 1.67 compared withDIII-D. As closely as possible, dimensionless parameters otherthan ρ∗ were matched globally to those expected in ITER andwell matched between JET and DIII-D. Comparison of profilesof these dimensionless parameters reveals a remarkable match(figures 2(a)–(d)) in a local sense, indicating that the scenariosin the two devices are governed by the same physics.

The global energy confinement is found to behave in aroughly Bohm-like fashion (figure 2(e)). This result does notyet provide a sufficient basis to predict performance of AIplasmas in ITER, as the present experiments did not matchall of ITER’s dimensionless parameters. In particular, thecollisionality was different from that expected in ITER, andthere are indications that this might have a significant effecton transport behaviour [12]. These studies are continuing,with the ultimate goal of determining the dependence ofconfinement on all of the relevant dimensionless parameters,thereby enabling accurate extrapolation to ITER.

2.3. Access to and safe shutdown from ITER operatingscenarios

DIII-D is developing and demonstrating startup, ramp-upand ramp-down scenarios that are compatible with ITER’srestrictive environment and capable of producing appropriatetarget conditions for several ITER operating scenarios. Lowbreakdown electric fields E have been achieved with (E ≈0.21 V m−1, lower than the maximum 0.3 V m−1 specified forITER) and without (E ≈ 0.42 V m−1) assistance from electroncyclotron heating (ECH). Transitioning to a large bore divertorconfiguration early in the current ramp reduces heat flux topoloidal limiters, and allows control of the internal inductanceℓi by varying the current ramp rate. Efforts to model thecurrent ramp phase with CORSICA [15], a 2D equilibriumand 1D transport predictive integrated modelling code, havesuccessfully matched the evolution of Te and q near the axis,but have so far not successfully matched the current profile, due

0

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46

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00 1 2 3 4 5 6

Breakdown/Burnthrough

Ramp-up Flat-top Rampdown

136303

Figure 3. An ITER baseline scenario discharge achieved withstartup, ramp-up and ramp-down scenarios optimized for therestrictions anticipated in future devices.

in part to inaccuracies in reproducing the experimental edge Te

evolution in the model [16]. These ramp-up and ramp-downscenarios have been combined into a single discharge [17], witha flattop matched to the ITER baseline (conventional H-mode)scenario (figure 3). Similar demonstrations have been carriedout in ohmic and hybrid scenario discharges.

DIII-D experiments have demonstrated that full ramp-down can be achieved with no additional transformer flux.We find that a controlled elongation ramp-down is effectivein preventing uncontrolled density increases that might triggera disruption. The ITER ramp-down scenario is terminatedby exceeding the vertical control limit, in agreement withcalculations. However, the termination occurs reproduciblybelow the ITER equivalent current level of 1.4 MA, wherearbitrary disruptivity rates are allowed.

DIII-D discharges have been developed to simulate theremaining leading operational scenarios for ITER, and detailedmodelling of these experiments is in progress [18].

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Nucl. Fusion 51 (2011) 094009 C.M. Greenfield et al

Flows, GAMs, ZF Te

ñeTe phase

~

ññ

ññ

~

ñ

ITG TEM ETG

4 400.4(cm-1)θk

∆r (cm) System 60 FIR 1 BES, reflectometer 1-10 PCI 1 DBS 1,20 DBS, HKBS 1 CECE, ECEI 1 BES, DBS 1 DBS/CECE

Figure 4. Validation studies in DIII-D take fluctuations over a wide range of spatial scales and fluctuating fields into consideration. Theoperative drift wave instabilities, in order of increasing kθ , are believed to be the ion temperature gradient (ITG), trapped electron mode(TEM) and electron temperature gradient (ETG) modes. The large suite of diagnostics, deployed to make many of the appropriatemeasurements, includes far infrared scattering (FIR), beam emission spectroscopy (BES), reflectometry, phase contrast interferometry(PCI), doppler backscattering (DBS), high-k backscattering (HKBS), correlation electron cyclotron emission (CECE) and electron cyclotronemission imaging (ECEI).

0 100 200Frequency (kHz)

300 400(a) (b)

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Figure 5. Measured (EXPT) and calculated (GYRO) fluctuation spectra at ρ = 0.6 with NBI-heated (upper row) discharges and thosewhere Te is increased by 50% by adding ECH (lower row). (a) Low-k density fluctuations measured by BES; (b) low-k electron temperaturefluctuations measured by CECE; (c) cross-phase between density and temperature fluctuations as measured by CECE and DBS (at the samelocation).

3. Improved physics basis for performanceprojections

3.1. Transport model validation

Detailed validation studies underway at DIII-D aim to testphysics-based transport models in L-mode [19–21], H-modeand QH-mode [22] plasmas. ‘Validation studies’ here refersto a process comparing measurements with simulated valuesto assess the models’ accuracy as important parameters arevaried. Fluctuation diagnostics for multiple fields (e.g. ne

and Te) spanning a wide range of spatial scales (kθρs ≈ 0.1to >10; kθ ≈ 0.4 to >40 cm−1) (figure 4), and simulations(e.g. GYRO [23]) have advanced sufficiently to make suchstudies viable. Essential to these comparisons are syntheticdiagnostics, digital analogues of real diagnostics applied tothe output of the models.

One such study is a recent L-mode Te/Ti scan [19], whereECH was applied to a neutral beam injection (NBI)-heatedL-mode plasma, with a resulting 50% increase in Te in themid-core region of interest (0.5 ! ρ ! 0.8). Figure 5 shows

the resulting change in measured and calculated fluctuationspectra at ρ = 0.6 [19]. The largest change with theadditional heating is the increase in low-k Te/Te. Although theshape of the resulting fluctuation spectrum is approximatelypredicted by GYRO, the simulation predicts a response thatis considerably larger than that observed in the experiment.Smaller changes, more closely predicted by GYRO, are seenin the density fluctuations ne/ne and the cross-phase 'nT .

GYRO-simulated thermal fluxes (not shown) are inqualitative agreement with experiment, with substantial,but incomplete, quantitative agreement. Recent GYROcalculations at ρ = 0.8 either do not converge or undergonumerical instabilities, indicating a possible need for coupledlow through high-k multiple modes non-linear simulations;this is currently under investigation. These observations fit ageneral trend in DIII-D that simulated transport and fluctuationlevels in the mid-core region (ρ ≈ 0.5) are often in betteragreement with experiment than those in the outer region(ρ > 0.75), where edge effects may be important [19]. Resultssuch as these are noteworthy as they point to potential futureresearch paths.

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Nucl. Fusion 51 (2011) 094009 C.M. Greenfield et al

0 100 200 300(a) (b)

-∇Pedge (kPa/m)

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1.2 Edge intrinsic torque(measured)

N-m

0.99 1.00

-2

-1

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Reynolds stress torque(measured)

Intrinsic torque(measured)

ΨN

N-m

Figure 6. (a) The measured edge intrinsic torque increases with the strength of the edge pressure gradient. (b) The Reynolds stress torque,measured via Langmuir probes, is very small at the edge, and insufficient to be the driver behind this intrinsic torque.

3.2. Edge momentum source for intrinsic rotation duringH-mode

Rotation has a significant, often beneficial, effect on bothtransport and stability in tokamak plasmas. DIII-D is workingtowards a predictive understanding of rotation, with recentemphasis on drive mechanisms [24]. The main source ofmomentum in DIII-D is neutral beams, but driving largerotation in ITER and other future tokamaks via NBI will bechallenging at best.

Experiments on multiple machines over the past severalyears have shown that the plasma can generate its own,‘intrinsic’ rotation. DIII-D experiments previously inferredan ‘intrinsic torque’ in the edge [25]. In order to improvethe extrapolation to ITER, we have investigated the torquedriving intrinsic rotation on DIII-D. We find that the intrinsictorque increases with edge pressure gradient (figure 6(a)).However, the turbulence-driven Reynolds stress [26] measuredin the edge of H-mode discharges produces a negligibletorque (figure 6(b)), inadequate to explain the intrinsic torque[27].

Recently, a strongly co-rotating layer has been observedwith a reciprocating multi-tip Langmuir probe just insidethe separatrix of H-mode discharges [28] (figure 7). The1 cm-wide layer forms independently of injected torque lessthan 50 ms after an L–H transition. In pure ECH plasmaswith no applied torque, the flat core rotation profile spinsup over 600 ms to match the edge rotation layer’s velocity(∼35 km s−1). This indicates that the layer is a cause, ratherthan effect, of core rotation and that viscous transport downthe layer’s gradient can slowly spin up the core.

Ion orbit loss has been suggested as a possible explanationfor the observed behaviour. A simple orbit loss model [29]successfully reproduces the layer’s profile shape, suggestinga link between these losses and formation of the layer. Thetoroidal–radial Reynolds stress [26] torque has also beenconsidered, but it is essentially zero outside the layer’s peakand becomes increasingly negative further inwards. It thus actsto oppose the spin-up of the core by transporting momentumoutwards, directed up the layer’s gradient, thereby acting tomaintain the peaked shape over long timescales.

r-rsep (cm)

0

20

40

-2 0 2

Co-NBIECH

Ctr-NBIohmic

LH + 0.05 sLH + 0.70 s

v φ (k

m/s

)

Figure 7. An edge co-rotation layer forms immediately after theL–H transition and is maintained for the duration of the H-modephase. The sign of this rotation is independent of the direction ofapplied torque.

3.3. Rotation driven by dominantly non-resonant magneticfields

In recent experiments in DIII-D, the first observation ofa theoretically predicted peak in the neoclassical toroidalviscosity (NTV) torque for low toroidal rotation rates hasbeen made [30] (figure 8). We have previously reportedobservations that static, dominantly non-resonant, n = 3fields apply a torque to the plasma that drives it towards acounter-rotating (opposite to the plasma current) finite velocity[31]. This effect is predicted by NTV theory [32], andcan be accessed through application of these fields fromnon-axisymmetric coils (the I-coil) in DIII-D. In the presentexperiment, a series of similar discharges was prepared,scanning over a range of toroidal rotation rates (. The neutralbeams are feedback controlled to maintain constant values ofβN and (, so that they will respond to additional torquesapplied by other sources. An n = 3 field is applied byactivating the I-coil, and the NTV torque determined by thechange in NBI torque required to maintain the programmed

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value of ( : )TNBI = −TNTV((). As shown in figure 8, theexperimental data exhibit a peak in the torque centred at thelocation predicted by theory.

3.3.1. QH-mode with zero net NBI torque. The NTVtorque can have practical benefits. QH-mode [33, 34] exhibitsperformance comparable to the ITER baseline ELMingH-mode, but without ELMs and their damaging effects. Itsattractiveness has been limited by an operational requirementfor strong NBI-driven rotation, which cannot be anticipated inITER or any reactor grade tokamak. In recent experiments,an n = 3 field produced by combining the internal I- andexternal C-coils has been used to apply NTV torque to maintainedge rotation with zero net NBI torque [35]. In these cases,QH-mode was accessed via the usual counter-NBI technique,with the torque removed (by balancing the NBI) over about a1 s interval. Following this, QH-mode operation with zero-netNBI torque was maintained for 1 s, until the n = 3 field wasremoved (figure 9). Previous experiments [36] have shown thatsufficient rotational shear near the edge, rather than rotationitself, is needed to access the QH-mode; in these experiments

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T NTV

(Nm

)

Figure 8. Comparison of measured NTV (diamonds), andcylindrical torque model (line) versus deuterium toroidal rotationrate (obtained from NCLASS) at ρ = 0.67. A least-squares splinefit (dashed) is shown for the data.

2.01.0 3.0 5.0(a) (b)

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Figure 9. (a) n = 3 NTV torque is applied, via the internal I- and external C-coils, to a pre-existing QH-mode. QH-mode phase ismaintained after the neutral beam driven torque is eliminated, with constant βN and confinement quality. (b) Toroidal rotation profilesbefore (t = 2.0 s) and after (t = 4.0 s) the NBI torque is removed.

that shear is provided by NTV rather than applied NBI torque.In addition, the effect of applying predominantly non-resonantmagnetic perturbations (RMPs) did not saturate at the highestperturbation amplitudes. This may open a plausible path toQH-mode utilization in ITER.

3.4. Improved understanding of energetic particle (EP)behaviour

Anomalous EP transport is an important, yet poorlyunderstood, term in the power balance, making transportanalysis challenging when this transport is large. This is amore serious concern for burning plasmas, where energetic α

particles will be the primary heating source. These concernsmotivate efforts to identify, understand and ultimately controlthis transport.

New diagnostics provide data to test models of EPinstabilities and associated transport. Discrepancies betweenmeasured and theoretical EP profiles suggest that the EPpopulation is depleted through both fine-scale turbulenceand large-scale instabilities (e.g. Alfven eigenmodes (AEs))including toroidal AEs (TAE) and reversed shear AEs(RSAE) [37].

Two-dimensional imaging of the modes’ internal structureprovides unprecedented detail on the non-linear interaction ofEPs with the background plasma. In a recent experiment[38], RSAEs were excited by NBI into a plasma with anoff-axis minimum in the q profile. Figure 10 shows acomparison of an n = 4 RSAE calculated by the TAE/FL[39] code (figures 10(a)–(c)) and observed via ECE imaging(figure 10(d)). The distinct poloidal shearing seen here is notreproduced by ideal MHD codes. It is reproduced by TAE/FLsimulations only when coupling between the mode and the fastion population is included (figures 10(c) and (d)), indicatingthat this effect may cause the shearing.

The newly commissioned DIII-D fast-ion loss detector(FILD) observes fluctuations in ion flux that are coherent atfrequencies in the TAE/RSAE range (60–100 kHz) with theenergy and pitch angle of lost ions; the enhanced losses occurwhen AE are present. When the measured pitch angle and

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(a) (b) (c) (d)

Figure 10. Imaging of Te/Te from RSAEs. (a), (b) TAE/FLsimulations; coupling between the mode and the fast ion populationis not included. (c) The same simulation, but now includingcoupling between the mode and the fast ion population, displays adistinct poloidal shearing. (d) Measurements with ECE imagingshow a similar sheared feature, indicating the importance ofcoupling in these simulations.

energy are used to trace backwards from the FILD aperturealong a path, calculated by the ORBIT [40] code, the lost ionsare found to originate from a location where AEs are knownto exist (figure 11(a)). The FILD results, in conjunction withother measurements and calculations, allow us for the first timeto directly associate fast-ion losses with AEs [37]. Figure 11(b)shows the spectrum of the ion flux fluctuations, indicatingboth TAE and RSAE frequencies are present, embedded inthe fast ions striking the detector, as expected from the ORBITcalculations. The FILD also produces an image of the fluxesstriking the detector. In figure 11(c), two features are observed.The feature on the left is identified as being associated withprompt fast-ion loss from the injected neutral beams, and ispresent during the entire duration of NBI. The feature on theright, however, vanishes after the TAE and RSAE activity seenin figure 11(b) disappears. The wealth of data available fromthis new diagnostic is enabling new comparisons with theory-based models of EP behaviour.

3.5. Predicting the H-mode pedestal

Performance predictions for ITER and other future devicesare strongly dependent on the structure and magnitude of theH-mode edge pedestal. Current models are beginning to yieldmore consistent predictions of the pedestal characteristics.Peeling–ballooning (P–B) stability constrains the pedestalprofile with the onset of ELMs, while the kinetic ballooningmode (KBM) has been proposed as a transport drivingmechanism in the pedestal. These two phenomena are includedin the EPED model [41], which has successfully matchedobservations on several tokamaks including DIII-D, JET andJT-60U. The EPED1 model, however, required an empiricallydetermined closure relation, )ψN = cβ

1/2p,ped, where c is an

empirically determined constant, )ψN is the pedestal widthand βp,ped is the poloidal β at the top of the pedestal. Recentimprovements to the EPED model (EPED1.6) have eliminatedthis requirement by using a direct calculation of both P–B andKBM stability with no free parameters [42].

Pedestal heights measured on DIII-D, JET and C-Mod arein good agreement with the first-principles based calculation of

EPED1.6 (figure 12(a)). Both versions of EPED give similarpredictions for ITER, with a pedestal height of βN,ped ≈ 0.6and width )ψ ≈ 0.04 (∼4.4 cm). At the ITER referencedensity, and typical peaking, βN,ped ≈ 0.6 corresponds toTped ≈ 4.1 keV.

As previously mentioned, the KBM is invoked as themechanism of transport in the pedestal. Although decisiveproof of the presence of the KBM in the H-mode pedestalis not in hand, observations have been made which are atleast consistent with this hypothesis. Figure 12(b) showsBES measurements in the H-mode pedestal, exhibiting featuresthat are expected of the KBM. In particular, we observe alarge number of medium-n harmonics, propagating in the iondiamagnetic direction [43].

Additional studies of the pedestal structure and itsdependence on machine size scaling have been underway injoint experiments between DIII-D and JET [44, 45]. Results todate are indicative of a lack of variation of the temperature anddensity pedestal widths with ρ∗ on the two devices (figure 13).Comparison of this dataset with additional data from ASDEXUpgrade is consistent with this observation. Note that thevariation seen in the DIII-D data in the density pedestal widthis believed to be a consequence of neutral penetration effectsand not the size scaling.

4. Heat and particle flux control methods

4.1. Reducing the impact of ELMs

The impulsive heat load from ELMs is a major concern forITER and other next step devices. DIII-D is exploring severaltechniques to eliminate or mitigate ELMs, including ELMsuppression by RMPs and ELM pacing by oscillating magneticperturbation and pellet injection.

DIII-D has pioneered the use of RMPs for ELM control.When a sufficiently large current is applied to an n = 3 internalcoil in DIII-D, ELMs are reliably and reproducibly suppressedwithin a particular range of safety factor (figure 14(a)). Thiseffect is consistent with P–B stability calculations with theELITE code [46] that indicate that the RMP reduces thepedestal gradient and/or current density below the P–B limit.We surmise, then, that the ELM suppression is a consequenceof increased transport within the pedestal. Recent researchin DIII-D is focusing on elucidating the physical causesunderlying that increase, so it can be better predicted andcontrolled.

Upon closer examination, the non-axisymmetric fieldsappear to have a direct impact on particle transport, with themagnitude of the effect being dependent more on the sizeof the applied field than on resonant (q dependent) effects(figure 15(a)). BES measurements indicate that changes in theRMP field have a prompt effect on fluctuations in the plasma,with the change in density occurring considerably more slowly(figure 15(b)) [43].

Resonant effects are, however, seen on energy transport.The electron temperature gradient (ETG) is reduced inside)q95, the range of safety factors where ELMs are suppressed(figure 16) [47]. Future experiments will focus on the physicsunderlying these observations.

During RMP experiments in high-β hybrid plasmas, thecharacteristics of ELM suppression were found consistent with

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Nucl. Fusion 51 (2011) 094009 C.M. Greenfield et al

#142111 FILD

Confined

Lost

FILD

TAEs

RSAEs

Energy and Pitch of Loss

300

20

40

60

80

Freq

uenc

y (k

Hz)

100

120

400 500Time (ms)

600 700 800

(a)(b)

(c)

Figure 11. (a) ORBIT calculations indicate that fast ions lost from the calculated vicinity of TAE and RSAE modes in the core of theplasma can travel to the location of the FILD; (b) FILD spectrum includes features that are strongly indicative of TAE and RSAE modes;(c) FILD image at 525 ms (time slide indicated by vertical line in (b)). Feature on the left-hand side of image is identified as due to promptloss from NBI, and is unrelated to AE activity. The feature on the right-hand side corresponds to the AEs detected in (b), and vanishes laterin the discharge when the AE frequency activity has ceased.

100 101 10

(a) (b)

2100

101

102

EPED1.6 Predicted Pedestal Height (kPa)

Mea

sure

d Pe

dest

al H

eigh

t (kP

a) DIII-D (pedscale + ITER demo)JET (ρ*)C-Mod (1090914)ITER prediction (base)

ñ/n spectrum (BES; au)

50 100f (kHz)

150 200 250150

n=19n=12

Figure 12. (a) Initial tests of the EPED1.6 model against data from DIII-D (squares), JET (circles), and C-Mod (triangles) are promising forimproved confidence in predicting the pedestal in future experiments, such as ITER (diamond). (b) Fluctuations detected by BESmeasurements in the H-mode pedestal are consistent with expectations for the KBM.

(a) (b)

0.0020

1

2

3

∆Te(

r oute

r/a) [

%] 4

5

0

1

2

3

∆ne(

r oute

r/a) [

%] 4

5

0.010

DIII-D low δDIII-D high δJET high δ

ρ*-ped0.002 0.010

ρ*-ped

Figure 13. Widths of the (a) electron temperature and (b) electron density pedestal as a fraction of the plasma radius in ρ∗ scans performedin JET and DIII-D. The variation of the electron density width in DIII-D is believed to be a consequence of neutral penetration effects.

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Nucl. Fusion 51 (2011) 094009 C.M. Greenfield et al

q95

∆q95

Dα (au)

Time (s)

1

2

3

4

0.4

0.6

0.8

Norm

aliz

ed E

dge

Curr

ent D

ensi

ty

Normalized Pressure Gradient

Unstable

Stable

ELMing — no RMPRMP ELM suppressed

Peeling-ballooning stability boundary

2 3

(a) (b)

4 2 3 4

Figure 14. (a) ELMs are routinely suppressed in DIII-D, over a particular range of safety factors, by n = 3 RMP fields driven by currentsflowing in internal coils. (b) Suppression of ELMs occurs when the pedestal conditions move below the calculated P–B stability boundary.

0

0.2

0.4

0.6

(a)3 4 5 6

I-coil = 1.3 kA

I-coil = 3.4 - 4.1 kA

q95

Nominal ELMsuppression window

ELMs suppressedELMy

|∆n pe

d/npe

d|

au

2.62.5Time (s)

4

3

2

1

0

I-coil Current (kA)

I-coilCurrent

ñ/n (BES ρ=0.85)

Density(BES ρ=0.85)

(ELMs remain suppressedthroughout the period shown) (b)

Figure 15. (a) The pedestal density decreases with increasing current in the RMP coils, but the effect is considerably broader in safetyfactor than the range where ELMs are suppressed. (b) During RMP ELM suppression, reduction in the pedestal density is associated withincreased density fluctuations that are in turn correlated with the magnitude of the RMP coil current.

ΨN

q 95

∆q95

0.6 0.7 0.8 0.9

3.4

3.3

3.5

3.7

3.8

3.94.0

Te (keV)

132741

0.5

1.0

1.5

2.0

3.6

Figure 16. The electron temperature gradient is reduced within theELM suppression window as q95 is ramped down.

those in standard H-modes [48]. The shape and magnitudeof the edge bootstrap current were observed to affect theconfinement of magnetic field lines in the pedestal region,so that a larger bootstrap current separates magnetic resonantsurfaces, reduces field line loss from the pedestal top andcorrelates with a lack of ELM suppression.

Although the impulsive heat load is of the greatestconcern, it is desirable to be able to control the steady heat loadas well. Recent experiments have been carried out to explorethe combination of RMP effects with a radiative divertor,created by injecting gas into the divertor region. In thesedischarges, ELMs are mitigated, but not fully suppressed,so that at a high pedestal density the peak heat flux isreduced almost to the level of the ELM-suppressed case(figure 17(a)). The time-averaged heat flux (figure 17(b))paints a very different picture: the discharges with full RMPELM suppression exhibit the highest time-averaged heat fluxobserved, and the radiative divertor discharges are as much asan order of magnitude lower [49].

In recent experiments, DIII-D demonstrated for the firsttime in the ITER baseline scenario that increased ELMfrequency can be achieved via pellet injection [50]. Whensmall pellets are injected from the low-field side of the plasmaat 14 Hz, the resulting ELMs occur at an surprisingly fast rate(up to 25 Hz) (figure 18). This can be compared with a naturalfrequency of about 5 Hz in this scenario. The small pellets donot penetrate the plasma and do not appreciably fuel the core.At the same time as the ELM frequency increases by a factor ofabout 5, the per-ELM energy loss decreases by a factor of 3–4.

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Nucl. Fusion 51 (2011) 094009 C.M. Greenfield et al

0

10

20

30

40

0.0 0.1 0.2 0.3 0.4 0.5 0.6

QPE

AK,IN

(MW

/m2 )

nPED (10 20 m-3)

0

1

2

3

4

QPE

AK,IN

(MW

/m2 )

Attached Detached

(a)

(b)

RMP + gas

RMP + gas

RMP only

RMP only

Pre-RMP

Pre-RMP

Figure 17. (a) Peak heat flux is reduced by RMP ELM suppression(‘RMP only’). Gas puffing in the divertor (‘RMP+gas’) increases thepeak value, since ELM suppression is lost with addition of the gas.(b) The steady, time-averaged, heat flux is the highest with RMPELM suppression, and is reduced with the addition of gas puffing.

Further experiments are planned using an upgraded injectionsystem, capable of introducing pellets at 30 Hz. ELM pacinghas also been observed in phase with a 20 Hz modulation onthe I-coil (the same coil as used for the RMP). The ELMs inthis case are triggered at both the positive and negative peaks,so the frequency is doubled.

Another approach to eliminating ELMs in tokamaks is todevelop operating scenarios that are naturally free of ELMs.QH-mode [35], described in section 3.3.1, is one such scenario,with performance similar to conventional H-mode. The keydevelopment needed in QH-mode has been to eliminate theneed for strong neutral beam torque, which will not be presentin future devices. The present research, using NTV-driventorques to generate the required E × B shear near the edge,has progressed sufficiently towards that goal so that QH-modemust be considered a possible option for future tokamakoperation.

4.2. Avoiding and mitigating disruptions

Rapid plasma shutdown via rapid particle delivery is beingdeveloped as a last-resort measure to mitigate disruptioneffects. This has two roles: first, to rapidly shut down the

Figure 18. ELM pacing via pellet injection has been demonstratedin DIII-D. (a) Plasma density is unaffected by pellet injection at arate of 14 Hz into a plasma with 5 Hz ‘natural’ ELM frequency.(b) The ELM rate increases to approximately 25 Hz, a surprisingfactor of nearly two faster than the pellet injection. (c) Dα emissionin a plasma with no pellet injection.

discharge, and second, to quench the runaway electron (RE)beam that may form. The latter is very challenging due tothe theoretically very high densities required. Also, studieswith the NIMROD code [51] indicate MHD activity may beless effective in deconfining REs in ITER than in presenttokamaks [52].

Previous studies [53] concentrated on massive gasinjection (MGI) from valves outside the plasma (figure 19(a)).These studies show that particle assimilation is only effectiveduring the thermal quench; particles that arrive later, duringthe current quench, are ineffective. Recent experiments haveexplored shattered pellet injection (SPI) as an alternativeparticle delivery technique (figure 19(b)). A large deuteriumpellet is injected into the DIII-D vessel, where it strikes ashatter plate, so that the resulting fragments penetrate theplasma [54]. Although more localized than the effect of theMGI, SPI increases the core density to a higher level (as highas 9 × 1021 m−3, a record density for DIII-D). The deuteriumis delivered more rapidly than with MGI in the present valveconfiguration (figure 19(c)), though earlier experiments haveshown that the MGI performance can be significantly improvedby a higher throughput delivery system [55].

In recent experiments, active control of the current andposition of a RE beam has been demonstrated. The beamis produced by injection of an Ar pellet into an inner walllimited plasma (figure 20). With no further intervention, thebeam dissipates in less than 50 ms. Initial attempts to controlthe radial and vertical position of the RE beam have beensuccessful, while the transformer has demonstrated the abilityto regulate the current. During the extended RE phase, nodirect wall contact, and so no vessel damage, takes place.

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Nucl. Fusion 51 (2011) 094009 C.M. Greenfield et al

0

D2 pellet

MGI SPI

Shatter platesSolid D2 shards

2.02

(a) (b)

(c)

2.01Time (s)

∫ned (1022 m-2)

Ip (MA)

2.00

21

SPIMGI

5

4

3

2

10

0

Injection

Figure 19. (a) In MGI, gas is injected from one or more valves, resulting in deposition in the outer part of the plasma. (b) SPI depositsparticles more centrally. (c) The density effect of SPI is both faster and larger (at least locally) than MGI in experiments done to date inDIII-D.

0

0

20

Loop

Vol

tage

(V)

Toro

idal

Cur

rent

(MA)

400

Ar pellet ablation

142661

142661

142666

142666

1

50 100Incremental Time (ms)

150

1.2-MA thermal 0.2-MA IWL 0.3-MA DND#142666

Figure 20. A RE channel is produced in inner wall limited discharges in DIII-D by the ablation of an argon pellet. In a discharge with nofurther intervention, the runaway channel decays rapidly (solid black line). Addition of inductive current drive, in conjunction with activeposition control, allows sustainment of the RE beam for extended periods of time.

This technique opens the possibility of maintaining the REbeam without risk to hardware, while energy is removed fromthe beam in a controlled manner, such as simply rampingdown the current, or dissipating the energy using the aboveparticle injection techniques. Further experiments are plannedto investigate these possibilities. In addition, the RE beamprovides a laboratory for measurement of RE growth and decayrates, which are found to be qualitatively consistent with thosepredicted by RE collisional drag theories.

4.3. Tritium retention in plasma facing components

Carbon is an appealing plasma facing material due to its heattolerance. However, tritium retention is a serious concern forITER. DIII-D research is working to quantify and developtechniques to mitigate retention.

Dynamic particle balance is used to determine the amountof hydrogenic particles retained in the vessel wall. Here,the sources (gas injection and, where applicable, neutral

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Nucl. Fusion 51 (2011) 094009 C.M. Greenfield et al

0

15

T-L/

s

30

0 1 2 3 4 5

137939Particle inputParticle exhaustWall inventory

Time (s)

H-mode

Figure 21. Time history of particle input and exhaust, and the wallinventory calculated by dynamic particle balance. The wall absorbsa large fraction of the gas introduced during the ohmic and L-modephases, but the entire inventory is removed during an extendedELMing H-mode phase.

beams) and exhaust (cryopumps) are monitored, with theremaining particles assumed deposited in plasma facingsurfaces. This technique has been used in both DIII-Dand Alcator C-Mod, and has been successfully benchmarkedagainst static measurements of vessel pressure before and aftera discharge. Figure 21 shows an example of dynamic particlebalance during an H-mode discharge in DIII-D. Althoughthere is finite loading during the ohmic and L-mode phases,surprisingly, the wall absorbs no additional particles during theH-mode. This result does not depend on the heating source;in both ECH- and NBI-heated H-mode discharges, the vesselinventory returns to its pre-shot level at the end of the discharge.

This does not preclude hydrogen from being bound upwith eroded carbon that redeposits at less exposed locations.DIII-D has tested thermal oxidation as a technique forremoving these co-deposits. Prior to the experiment, anextensive laboratory testing programme was carried out at theUniversity of Toronto. The DIII-D vessel was then filled with a10 Torr He : O2 = 90 : 10 mixture (∼1 Torr O2 partial pressure)and baked for 2 h at 350 C, with pre-characterized sample tilesinserted. Estimates made by monitoring gas removed duringand after the bake indicate about half of the co-deposits wereremoved (figure 22). This result is consistent with expectationsbased on lab tests at the same partial pressure of O2. Furtherlab tests indicate that a larger fraction of the deposits can beremoved from the tiles by increasing the pressure. Followingthe bake, high-performance hybrid plasmas were obtained aftera short clean-up period. O2 baking did not damage internalcomponents, and did not compromise the vessel conditionsneeded for advanced scenarios [56].

5. Specific ITER research plan issues

As ITER approaches its construction phase, DIII-D isincreasing its emphasis on predictive understanding of physicsgoverning ITER’s behaviour. However, DIII-D continues torespond to specific ITER concerns, now increasingly focusedon development of the research plan. We report on two suchareas of focused DIII-D research here.

D Re

mai

ning

/ D

Initi

al

Time (h)0

0.060.080.10

0.20

0.400.600.801.00

2.00

1 2 3 4

DIII-D divertor tile bakeUTIAS LAB 1.6 torr O2 350˚CDIII-D 2 torr O2 350˚CDIII-D divertor tile bakeUTIAS LAB 600 torr O2 350˚C

Figure 22. Approximately half of the deuterium inventory wasremoved from DIII-D during a 2 h, 350 C, bake with a 2 Torr partialpressure of O2 (star), consistent with laboratory tests (circles)carried out at the University of Toronto under similar conditions.Further tests at the University of Toronto (triangles) indicate thatvirtually all of the deuterium can be removed by baking at a higherO2 partial pressure.

5.1. Characterization of the L–H threshold

A set of experiments was performed to determine H-modeaccess requirements for ITER’s first (non-nuclear H andHe) and second (activated D) operational phases. Theseexperiments [57] build on previous work [58] demonstratingthat the L–H transition threshold power PTH increases withpositive (parallel to IP) rotation in both H and D plasmas.Similar behaviour was confirmed in He plasmas. PTH for thethree species occurs approximately at the following ratios: PTH

(D : He : H) ≈ (1 : 1.4 ± 0.1 : 2.1 ± 0.1). An intermediatecase, with H-NBI injected into He plasmas, exhibits thresholdsbetween pure H and He, depending on the species’ relativeconcentrations.

The density dependence of the H-mode power thresholdis a key component of the ITPA2008 scaling [59]. Manydevices, including DIII-D, observe a minimum in density thatcontrasts with this scaling. This minimum is seen in the newresults (figure 23), in both D and He plasmas, albeit at differentdensities. Also, as previously observed, PTH in He is higherthan in D at low densities, with the two cases moving closertogether with increasing density.

Another finding of this study is the steady decrease in PTH

as the X-point moves closer to the divertor. This, combinedwith choosing the ‘right’ density, may provide a path toobtain H-mode during the non-nuclear phase of ITER. Theseexperiments also showed applying resonant magnetic fieldswith the RMP ELM control coils can increase the powerthreshold.

5.2. Predict the impact of TBM on ITER’s performance

ITER will include three pairs of toroidally localized TBMswhose ferromagnetic structural materials will perturb themagnetic field. Theory to predict the impact of these fieldperturbations on performance is lacking, so DIII-D simulatedthe effect with a mockup (figure 24(a)) of a single TBM set

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Nucl. Fusion 51 (2011) 094009 C.M. Greenfield et al

PTH,SCAL08(D)

P TH (M

W)

Deuterium

5 6

4

4

3

3

2

2_ne (×1019 m-3)

1

10

0

Helium

×1.5-2×1-1.5

Figure 23. The H-mode power threshold as a function of theline-integrated L-mode density for D and He plasmas (1.0 MA,1.65 T) with auxiliary heating by ECH. The dotted line correspondsto the ITPA2008 power threshold scaling relation for D plasmas.The results with balanced (torque free) NBI are very similar.

with coils driving fields up to ∼3× the relative amplitudeof those in ITER. The TBM mockup assembly was insertedinto a midplane port on DIII-D for these experiments. Itwas subsequently removed, but can be reinserted for futureexperiments.

Measurements were made of the TBM’s impact onparameters including L–H threshold, confinement, rotationand EPs [60]. The effect on rotation and confinement qualityis shown in figure 24(b). Here, the local ripple δ ≡ (Bmax −Bmin)/(Bmax + Bmin) is the combined ripple arising from boththe TBM and toroidal field coil non-axisymmetries. Theequivalent value for ITER, in front of a TBM, is δ ≈ 1.2%, andthe DIII-D coil at full current (1.2 kA) is capable of reachingapproximately 3.6%. In these experiments, we found onlysmall effects on performance with the TBM mockup current setto match the ITER value. Decreased confinement and toroidalrotation were observed, but only when the coil current wasincreased to produce 2–3 times the ripple anticipated in ITER.Global losses of fast ions were small [61].

These experiments were carried out at βN ≈ 2.2,somewhat higher than βN ≈ 1.8 anticipated in the ITERbaseline scenario. We chose the higher pressure because theeffect on stored energy and density increases strongly with βN,the results were much clearer at the higher β.

Even where detrimental effects are seen, it may be possibleto minimize them by reoptimizing the magnetic error fieldcorrection in DIII-D, a technique that is directly applicable toITER. Calculations with IPEC [62] indicate that amplificationof the n = 1 perturbation is responsible for the majority ofthe NTV torque on the plasma, which in turn, we speculate, isresponsible for the negative impact on rotation (figure 25(a)).Further calculations indicate that the n = 1 component canbe attenuated using error field correction techniques. Initialexperiments in DIII-D appear to confirm this. As shown infigure 25(b), the density at the onset of locked modes, an

indicator of the tokamak plasma’s sensitivity to error fields,increases with increasing TBM coil current. Reoptimizationof the error field correction reduces the threshold to its previouslevel prior to introduction of the TBM field. These experimentswere carried out in low-pressure L-mode plasmas; furtherexperiments are planned to demonstrate the applicability ofthis technique to H-mode.

6. Summary and future research directions

We have described recent DIII-D research highlights in severalareas; the references herein provide the reader a more completediscussion of these results.

Recent efforts to optimize steady-state scenarios focus onthe relationship between transport and the current profile. Thisrelationship is not captured by present transport models, sothe data obtained from these studies can in turn drive furtherdevelopments of those models. Aspects of other scenarioshave been studied, including ρ∗ scaling of advanced inductivescenarios and generation of QH-mode without an externaltorque. Safe access to and exit from ITER operating scenarioshave been studied, with full discharge scenarios demonstratedin several different regimes. Future work will benefit from fastwave heating and additional ECH/CD both for tailoring profilesand establishing more reactor-relevant conditions by primarilyheating electrons. Off-axis NBI will provide additional controlof the current profile.

Validation of physics-based models is a major emphasisacross the DIII-D programme. Here we described such effortsin transport, rotation, energetic particles and the H-modepedestal. These are given as examples; the approach of usingmodelling to design and interpret experiments permeates theDIII-D programme. Future opportunities will take advantageof additional diagnostics, already commissioned and currentlyundergoing construction or upgrade.

Methods to control steady and transient particle andenergy fluxes are extremely important to ensure the reliableoperation and survivability of future devices. DIII-D haspioneered the use of resonant magnetic perturbations tosuppress ELMs and remove their damaging effects. Recentresearch points to the relationship between the appliedmagnetic perturbation and thermal transport as the key. Furtherexperiments and theory are being prepared to focus onthe detailed effects of imposed three-dimensional magneticfields on the plasma, especially the mechanisms under whichtransport is increased. We are also developing alternativestrategies, including pellet ELM pacing. Another approachunder study is to identify naturally ELM-free operationalscenarios. An example is the QH-mode, which has for thefirst time been maintained without an external torque drivenby neutral beams.

DIII-D has also explored techniques to mitigate thedamaging effects of disruptions by introducing large quantitiesof particles to the plasma. A major challenge of this line ofresearch has been to introduce a sufficient number of particlesto prevent the development of a RE beam. Recently, DIII-Dhas demonstrated position control of this beam, opening thepossibility of allowing it to safely dissipate in a controlledmanner.

Experiments continue to evaluate carbon as a first wallmaterial in tokamaks. We have determined that deuterium is

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Nucl. Fusion 51 (2011) 094009 C.M. Greenfield et al

0.0

-0.1

-0.2

-0.3

-0.4

-0.5

-0.60 1% 2% 3%

ApproximateITER level ofperturbation

∆vtor /vtor

∆H98/H98

Bmax BminBmax + Bmin

δ ≡

(a) (b)

Figure 24. (a) The TBM mockup, combining toroidal racetrack coils and a poloidal solenoid, was inserted in a radial port on DIII-D.Currents in the two sets of coils simulate the field perturbation in ITER arising from a single set of TBMs. (b) Relative reductions ofconfinement factor H98 and toroidal velocity as a function of the local ripple δ at βN ≈ 2.2. Below the ITER perturbation level, the effect onconfinement was too small to measure.

Figure 25. (a) Calculations with IPEC indicate that the NTV torque imposed by the TBM mockup on the plasma is primarily due toamplification of the n = 1 field component (black circles). Optimization of the n = 1 field error correction is expected to attenuate thiseffect. (b) Susceptibility of the plasma to low-density locked modes is used to gauge the effect of error fields. The effect increases withcurrent in the TBM, but is cancelled via reoptimization of the error field correction as predicted by IPEC.

actually removed from the wall during H-mode. Hydrogen-bearing co-deposits that are shadowed from the plasma canbe removed by baking the vessel in an atmosphere containingoxygen.

The DIII-D research programme has demonstrated stronginterest and ability to respond to ITER’s short- and long-termneeds. Experiments with D, H and He plasmas are contributingto plans for the initial operation of ITER, while at the same timeproducing a large dataset useful for confinement and transportresearch.

Finally, DIII-D responded to an ITER-identified need fortests of the effects of the TBM on plasma performance, with thefavourable result that at ITER equivalent perturbation strengthand beta, little or no degradation is seen in performance.Further experiments are being planned to verify that even thesmall effects seen in ITER-like conditions can be attenuatedby reoptimizing the error field control.

DIII-D is currently planning experiments to be carriedout during the next campaign, and further into the future.These activities will take advantage of improvements thatare currently underway, including additional EC power,flexible aiming of NBI, and a large number of new andimproved diagnostics. We anticipate continued production of

exciting results in support of ITER and other devices on thedevelopment path to fusion as an attractive power source.

Acknowledgment

This work was supported by the US Department of Energyunder DE-FC02-04ER54698.

Appendix. The DIII-D Team

G. Abla1, M.M. Abrahim1, E.O. Allen1, S.L. Allen2,D.D. Anastasi1, P.M. Anderson1, M.E. Austin6, F.W. Baity4,J.P. Bakalarski1, C.P. Balance5, I. Balboa6, E.M. Bass7,D.J. Battaglia4, C.B. Baxi1, L.R. Baylor4, M. Becoulet8,E.A. Belli7, J.W. Berkery9, L.A. Berry4, M.N.A. Beurskens6,J.M. Bialek9, J.A. Boedo10, I.N. Bogatu11, R.L. Boivin1,P. Bonoli12, A.H. Boozer9, B.D. Bray1, D.P. Brennan5,S. Brezinsek14, N.H. Brooks1, D. Buchenauer15, R.V. Budny16,R. Bulmer2, P. Burratti11, K.H. Burrell1, R.J. Buttery1,J.D. Callen18, R.W. Callis1, G.L. Campbell1, J.M. Candy1,J.M. Canik4, T.N. Carlstrom1, T.A. Carter19, W.P. Cary1,T.A. Casper2, M. Cengher1, C.D. Challis6, F.R. Chamberlain1,

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Affiliations

1 General Atomics, USA2 Lawrence Livermore National Laboratory, USA3 University of Texas, USA4 Oak Ridge National Laboratory, USA5 Auburn University, USA6 EURATOM/CCFE Fusion Association, Culham Science

Centre, UK7 Oak Ridge Institute of Science Education, USA8 Association Euratom CEA, DSM/IRFM, Cadarache, France9 University of California, San Diego, USA

10 Columbia University, USA11 FAR-TECH, Inc., USA12 Massachusetts Institute of Technology, USA13 University of Tulsa, USA

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Nucl. Fusion 51 (2011) 094009 C.M. Greenfield et al

14 EURATOM-FZ Julich, Germany15 Sandia National Laboratories, USA16 Princeton Plasma Physics Laboratory, USA17 ENEA-Frascati, Italy18 University of Wisconsin, USA19 University of California, Los Angeles, USA20 Courant Institute, New York University, USA21 University of Arizona, USA22 ITER Organization, Cadarache, France23 Association EURATOM-FOM Rijnhuizen, The Netherlands24 University of Toronto, Canada25 Association EURATOM-CIEMAT, Madrid, Spain26 University of California, Davis, USA27 Max Plank Institut fur Plasmaphysik, Garching, Germany28 University of Maryland, USA29 Palomar College, USA30 Association EURATOM-VR, Alfven Laboratory,

Stockholm, Sweden31 ASIPP-Hefei, People’s Republic of China32 Consorzio RFX, Italy33 National Fusion Research Institute, Korea34 CompX, USA35 Purdue University, USA36 University of California, Irvine, USA37 Australian National University, Australia38 Japan Atomic Energy Agency (JAEA), Japan39 Association EURATOM, Belgium40 Max Plank Institut fur Plasmaphysik, Greifswald, Germany41 LeHigh University, USA42 Butler University, USA43 Korea Atomic Energy Research Institute, Korea44 University of California, Berkeley, USA45 Association EURATOM-OAW/ATI, Vienna, Austria46 University of Marseille, France47 Helsinki University of Technology, Finland48 Brown University, USA49 ALITRON50 University of Basel, Switzerland51 Association Euratom CEA, DSM/IRFM, Cadarache, France52 EURATOM Lisbon53 Lexam Research, Redwood City, USA54 Association EURATOM-IST, Lisbon55 Pohang University of Science & Technology, Korea56 Institute for Tokamak Physics, Moscow, Russia57 Institute for Plasma Research, India58 CRPP-EPFL-Lausanne, France59 Fusion for Energy Joint Undertaking, Barcelona, Spain60 West Virginia University, USA61 Idaho National Laboratory, USA62 Heinrich-Heine University, Dusseldorf, Germany63 Georgia Institute of Technology, USA64 Association EURATOM-Tekes, Finland65 University of York, UK66 Mesa College, USA67 National Renewable Energy Laboratory, USA68 Ecole Polytechnique, France

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