�������

���

Wave-Plasma and Advanced Tokamak Research on Alcator C-Mod

Presentation to FESAC Facilities Review June 13, 2005

Naval Research Laboratory

Presented by A. HubbardMIT Plasma Science and Fusion Center,

for the C-Mod team

�������

���Outline

• Description of Programs– RF Tools– Wave physics research– Advanced Tokamak research

• Uniqueness and Importance of C-Mod Research– Wave-plasma – Advanced Tokamak

• Complementarity and coordination with US and international programs

• Summary: Contributions to vitality of fusion science

�������

���

• 8 MW total source power.• Up to 6 MW coupled.• 4-strap Antenna has

variable frequency and phase– Can heat at two radii

simultaneously eg. At 4.5 T, 70+80 MHz heats

at r/a= 0.1 and 0.5

• Up to now, ICRF has been the onlyaux. heating, and is used for most C-Mod experiments in all science areas (not just advanced scenarios).– Places high priority on reliability,

understanding.

C-Mod ICRF Hardware

40-80 MHz~ 80 MHzFrequency

4 Strap2 - 2 StrapAntenna

variablefixedPhase

4 MW2 x 2 MWPower

�������

���

Lower Hybrid Current Drive System:Commissioning has started

• LH system designed for current profile control, deposition far off axis (r/a to 0.8), key to advanced scenarios.

• Launcher developed in collaboration with PPPL.– 4 x 24 guides, for well controlled spectrum.– Flexible parallel refractive index N//,

variable over range 2-4 between or duringdischarges using phase shifters. (Important for controlling localization of CD)

• 12 klystrons, 3 MW source now operational.

• Will add a second launcher in 2007.– Will allow two different N//,

giving higher ηLH, better localized jLH(r).

– 4 more klystrons would raise source to 4 MW total.

�������

���

C-Mod makes major contributions to Wave Physics

-2-1

0

12

MC layer

0.64 0.66 0.68 0.70 0.72 0.74 0.76R (m)

0

1

2

ExperimentalSynthetic

Re(∫ ne dl)~

|∫ ne dl|~

[a.u

.][a

.u.]

• Strong diagnostics, notably Phase Contrast Imaging to detect waves.• Leaders in RF modeling, code improvements e.g. TORIC, via SciDAC.• Example of recent success was first identification of Mode Converted

Ion Cyclotron Wave as well as Ion Bernstein Wave– good example of code verification, validation.

Addresses FESAC Priorities Panel Science Topic T11:How do electromagnetic waves interact with plasma?

• ICW was predicted by F. Perkins in 1970s.

• Same process will occur in D-T in a burning plasma!

10-3

1

IBW FW

LCFS

ICW

ICW

MC

-1

�������

���

Near-term RF Program includes Mode Conversion Current and Flow Drive Expts

• Use improved understanding and modeling to predict MC Current Drive.

• Indications of MCCD near q=1 already seen via sawtooth modification.

• TORIC predicts scenarios with strong on-axis current drive, could be valuable complement to LHCD.

• Also using TORIC to explore scenarios which maximize flow drive, will compare with rotation measurements.

0.0 0.2 0.4 0.6 0.8 1.0r/a

-20

0

20

40

60

80

j (k

A/m

2)

Total driven current for 1 MWIBW + ICW 33 kA/MW

IBW Contribution (x10)

Ohmic profile current formodeled discharge (EFIT)

TORIC Prediction for On-axis MCCD

(5.4 T, 50+80 MHz)IMCCD = 100 kA for PICRF = 3 MW

�������

���ICRF Antenna, Coupling Studies

• Group has iteratively improved antenna design, power handling and reliability.– 5 MW now routinely

available.– Record antenna power

densities.

• Involved in improving, comparing to, antenna-plasma models.– Predictive capability crucial

for ITER RF antennas.

• Experimental studies find that antenna loading is mainly sensitive to density pedestal, in contrast to lower ne experiments.– Expect an intermediate

situation on ITER.

Representation of C-Mod 2-strap antenna in 3-D solid antenna model TOPICA3. (Faraday screen, backplane not shown)Collaboration with U. Torino.

�������

���ICRF minority heating and sawtooth control.

• While D,H is our usual heating scenario, D,He3 heating is also of interest for burning plasmas.

• Sawtooth control, key for ITER, has been observed in both minority heating and MCCD scenarios.– C-Mod typically has lower

energy fast ion tail (E<300keV) than other (lower density) tokamaks.

– This is also expected on ITER.– In future, will test sawtooth and

NTM control via LHCD.

• Alfven modes are also driven by RF-produced energetic particles– Will be covered in

macrostability talk.

2

2.5

3

2

2.5

3

0.86 0.88 0.9 0.92 0.94

2

2.5

3

Te0

(k

eV)

Time [sec]

+90°

-90°

Heating

Te0

(k

eV)

Te0

(k

eV)

Example of sawtooth control via central minority heating

�������

���Initial LH Experiments

• Low power coupling experiments measured reflection as a function of phase, limiter-grill distance.– Low R (<10%) can be achieved.– Confirms expected RF

performance of launcher, klystrons, control systems.

• Plasma-material interactions led to erosion and limited power handling; plan to change Ti material of final coupler section.

• Next phase of experiments will increase power, pulse length, measure CD efficiency, localization.

• A 32-channel spatially imaging X-ray spectrometer to detect bremsstrahlung was recently commissioned– will be a key diagnostic

of fast electron profiles.

�������

���Advanced Tokamak Research Thrust

• AT Thrust integrates elements of transport, wave, macrostabilityand divertor physics.

• Focuses on control of current and magnetic shear as well as transport and kinetic profiles.– RF systems (ICRF +LHCD) provide key control tools. – Also adding new cryopump, important for density control.

• Integrated scenario modeling is a key part of the program.

Addresses FESAC Priorities Panel Science Topics:T1: How does magnetic structure affect fusion plasma confinement?T3: How can external control and plasma self-organization be used to improve fusion performance?

�������

���5 year goals of the AT physics program

1. Demonstrate and model current profile control using LH and ICRF waves, at high densities (>1020 m-3).ωp, ωc are key parameters for wave physics – C-Mod is very similar to ITER.

2. Understanding, control and sustainment of Internal Transport Barriers, with coupled ions and electrons, τe-i << τE (Te~Ti ) and without momentum input (RF only).Electron-ion equilibration is particularly important for transport in barriers, since different channels often respond differently.

3. Achieve full non-inductive current drive (70% bootstrap) and extend pulse length to near steady state (5 sec, 4-6 τCR).Will allow us to study fully relaxed non-ind. current profiles.

4. Attain and optimize no-wall β limits (βN ~ 3). Explore means of achieving higher values. ITER advanced operation currently planned to be in this range.

�������

���

Example of long-term, non-inductive AT target scenario

• One of many optimized scenarios modeled with ACCOME.– ILH=240 kA– IBS=600 kA (70%)

J (M

A /

m2 )

r / a

Ip = 0.86 MA Ilh = 0.24 MA fbs = 0.7

r / a

Saf

ety

Fac

tor

- q(

r)

q(0) = 5.08

qmin = 3.30

q(95) = 5.98

• Double transport barrier • BT=4 T• ICRH: 5 MW• LHCD: 3 MW, N//0=3• ne(0)= 1.8e20 m-3

• Te(0)=6.5 keV (H=2.5)• βN=2.9

Scenarios without barrier, or only an ITB, have similar performance.

P. Bonoli, Nucl. Fus. 20(6) 2000.

�������

���

• Ideal no-wall limit βn~3, with:– strong shaping.– optimized p(r), j(r).

• LHCD far off axis (r/a~0.8) is critical to produce the needed broad reversed shear region and pressure profiles.

• Antennas for active core MHD spectroscopy may be able to measure linear growth rates.– Feedback on power, profiles

to avoid limit.• Study ELM, core MHD

interaction.• Try stabilization of NTMs using

LHCD and/or MCCD.

MHD stability of non-inductive plasmas

ACCOME

• Plan to carry out a design/feasibility study of passive and active stabilization methods to allow β > no-wall limit.

• Will install as results warrant.

PEST-II/KINX

�������

���

Tailored ICRF has been used to control core transport and profiles

• Barriers form from EDA H-mode.• OFF-axis heating alone causes

strong density peaking, ~ const T.– Core thermal transport reduced to

ion neoclassical.– Sawtoothing plasmas, without

reversed shear.– Also seen in ohmic H-mode.

• Addition of ON-axis heating tends to increase transport (D and χeff). – Ratio can be used to control

density and impurity rise!– n is clamped, but T, neutron rate

increase.• Increasing understanding of transport

in terms of ITG, TEM stability through theory collaborations.

Ele

ctro

n P

ress

ure

(MP

asca

ls)

0.15

0.10

0.05

0

0.20

0.0 0.2 0.80.60.4 1.0r/a

1.5 MW central ICRFadded into fully formed ITB

ITB, 2.35 MW Off-axis ICRF

H-mode, No ITB

t=1.294 s

t=1.127s

t=0.894 s

ohmicITB +off-axisITB+off-axis+central

χeff from TRANSP (lower power discharge)

�������

���

Transport control/barrier research will move towards exploring effects of shear, flow

• As current profile control tools (LH and MC) mature, we will focus on studying effects of magnetic shear on transport, ITB formation and properties, eg.– Flat shear, q>1 “hybrid scenario”, – Weakly and strongly reversed shear.

• If RF results warrant, may also begin to explore effects of flow drive on barrier formation, control.

• Hope that we will still be able to control degree of transport reduction, find optimum levels of ion, electron, particle transport.– Control of radiation, avoidance of MHD pressure limits are key

to usefulness of core barrier regimes for long pulse scenarios. Will this be possible in our ITER-relevant regime?

�������

���

Answers to questions in the panel’s charge

�������

���Uniqueness of C-Mod RF Program

• C-Mod is the ONLY US Facility focusing on ICRF minority heating (D, H and D, He3).

• C-Mod is the ONLY high performance divertor expt in the world using RF exclusively for heating and current drive.– Compact ICRF antennas have record power density.– Team has been forced to face and solve the physics and

engineering challenges of reliable RF.

• ITER is counting on such systems for its success.– US may well provide much of the ITER ICRF system, and

maintaining C-Mod RF expertise, capability is crucial. – C-Mod collaborates closely with ORNL, PPPL on ICRF.

�������

���Uniqueness of C-Mod RF Program (2)

• C-Mod is the ONLY US Facility using Lower Hybrid Current Drive.– Highest efficiency technique for far off axis current drive

(r/a ~0.8), which is needed to optimize j, p profiles for high β in advanced scenarios.

– Flexible N// of C-Mod LHCD is ideal for j(r) control.

• Only LHCD experiment in the world operating at the density, field and shape of ITER.– n, B set ωp, ωc, the key parameters for wave physics; most LH expts

at much lower ne.– H-mode density profiles, with nped ~ 1020 m-3, pose a particular

challenge that cannot be addressed in a limiter device such as FTU.

• ITER is considering the addition of a LHCD system, which would greatly enhance its AT capabilities.– Timely results from C-Mod experiments are needed to impact this

decision.

�������

���Uniqueness of C-MOD RF Program (3)

• C-Mod’s combination of RF tools, wave and plasma diagnostics, modeling expertise and experimental focus on RF make it a recognized world leader in wave physics, particularly regarding short wavelength modes.– Many recent publications, invited

and review talks at international RF conferences.

– As an indicator, C-Mod team (including students) were recently invited to help analyze AUG and JET Mode Conversion experiments.

First full wave simulations of LH waves, at PSFC, promise to help resolve long-standing mystery of ‘spectral gap’.

�������

���

Uniqueness of C-Mod Advanced Tokamak Program

• In physics terms, “Steady-state” current drive implies pulse lengths >> current relaxation time τCR.– This is needed to study fully

relaxed RF and bootstrap-driven current profiles.

• C-Mod τCR~ 0.2-1.4 s [Zeff=1.5; Te= 2-7.5 keV ]

• C-Mod routinely runs pulse lengths longer than τCR, even with inductive current drive and at highest Te. j(r) is normally fully diffused early in the flat top.

• This is unique among high-performance divertor experiments.

With non inductive current drive, C-Mod will be able to extend pulse length to TF limit of 5 secs (5 T).– Have already run 3 sec pulses,

~2 sec is typical.– No hardware upgrades

needed.

(MajorRadius)-2

(m-2)

Pu

lse

Le

ng

th /

Re

laxa

tio

nT

ime

C-Mod (5s)

JT60-U (30s)

JET (50 s)

ITER (1000 s)

DIII-D (10s)

AUG (10s)

2 3 / 2

e

CR

eff

a1.4 T

Z

Te=6 keV (ITER 19 keV), Zeff=1.5

τ =

C-Mod (2s)

�������

���

Uniqueness of C-Mod Advanced Tokamak Program (2)

Critical issues for ITER advanced regimes are reflected in ITPA High Priority Research Tasks in Transport Physics : “Improve experimental characterization and understanding of critical issues for reactor relevant regimes with enhanced confinement, by:– Obtaining physics documentation for transport modeling of ITER

hybrid and steady-state demonstration discharges.– Addressing reactor relevant conditions, e.g., electron heating,

Te~Ti, impurities, density, edge-core interaction, low momentum input...”

• Most other AT expts have τe-i > τE, Ti Te, and/or use NBI for core fuelling and rotation drive in barriers.

• C-Mod has strongly-coupled electrons and ions, τe-i << τΕ , uses all RF and has no core momentum or rotation drive.It thus is uniquely equipped to address these ITER and reactor relevant conditions in an integrated way.

≠

�������

���

Uniqueness of C-Mod Advanced Tokamak Program (3)

• C-Mod has record (ITER-like) values of divertor power flux. – Handling such fluxes, while maintaining density low

enough for efficient current drive, is likely to be one of the greatest challenges for applying non-inductive scenarios on a burning plasma.

– C-Mod’s experience will be very valuable (PFC materials and techniques covered in talk by Lipschultz).

• C-Mod’s use of LHCD for far off-axis current drive in the high density ITER regime is unique among advanced tokamak programs. – As noted, ITER is considering adding this capability. A

successful C-Mod program would position the US for leadership in this area.

�������

���

US heating and current drive programs are highly complementary

• Almost no overlap between C-Mod and DIII-D, NSTX tools!• TOGETHER, the US facilities have great breadth and strength

in the physics and technologies of interest for burning plasmas.

NBIEBWHelicity inj.HHFW

ECCDNBIFWCD

LHCDMCCDFWCD

Current drive

NBIHigh Harm FW (~10 ωci)

NBIECHFW (>2 ωci)

ICRH (Minority and Mode conv.)LH

Primary Heating

NSTXDIII-DC-Mod

�������

���

Different tools, parameters lead to complementary research on advanced scenarios

C-Mod focuses on the challenges of integrating non-inductive current drive, including bootstrap current, in high density, strongly coupled regime with high power flux.

• In near term, will push to achieve and optimize ideal no-wall MHD limits, as currently envisaged on ITER.

• Long pulse, equilibrated j(r).

DIII-D is a leader in RWM stabilization, β above no-wall limits. Active stabilization may be considered for ITER upgrade.

• Strong ITB program, mainly with Ti > Te, NB dominated.

NSTX operates at highest β. Future program includes non-inductive operation at increased pulse length and reduced ne.

Not necessary, or feasible, for any one program to cover all of the regimes or issues of concern for advanced scenarios!

�������

���

C-Mod RF and AT research is integrated with the international program and addresses ITER

needs, primarily through ITPASteady State Operation and Transport Physics groups focus attention

on needs for advanced scenarios on ITER:• C-Mod scientists have been improving and benchmarking both ICRH

and LH models for ITER, which has similar ωp, ωc, wave regimes as C-Mod.– Predictions of LHCD efficiency vary by ~30%. C-Mod LHCD data will

be crucial in testing models, projections. e.g., Do jLH(r) agree with predictions? Are there any unexpected nelimits?

• C-Mod is executing and planning many joint experiments, eg: – Scaling of spontaneous plasma rotation.– ICRF-generated barriers.– Hybrid scenarios with q~1.– Assessing beta limits for full current drive plasmas with significant jboot.

• Contributing to ITPA databases, to test commonality of physics in various regimes across devices.

�������

���

C-Mod Contributions to Fusion Science: Waves

What research opportunities would be lost by shutting down C-Mod?

2004 FESAC Priorities Panel recommended six target areas as ‘opportunities for enhanced progress’ (ie. Priorities for more US resources). These included:

“Extend understanding and capability to control and manipulate plasmas with external waves.”

• Without C-Mod, US would have no experimental program on minority ICRH heating. This is the baseline scheme for ITER, and US is slated to play a major role. Many technical and physics issues remain!

• Exciting advances in mode conversion physics would stop; we would not find out whether MC can be used for efficient current or flow drive.

• Would not be able to test the feasibility of Lower Hybrid Current Drive for current profile control, in high performance regimes with transport barriers and at ITER density and field.– This would reduce the likelihood of LHCD being added to ITER,

and certainly the role of US in the advanced tokamak program.

�������

���

C-Mod Contributions to Fusion Science: Advanced Scenarios

FESAC ‘Opportunities for enhanced progress’ cont:“Integrated understanding of plasma self-organization and external

control, enabling high-pressure sustained plasmas.”

• C-mod provides a unique opportunity to advance this integration:– Using LHCD and ICRH for external control of current and transport,

in ITER-relevant wave regime. – In a high-pressure, reactor-relevant regime with equilibrated

electrons and ions, no particle or momentum sources, all crucial for transport barriers and self-organized bootstrap current generation.

– Sustained for many transport and current relaxation times.

• Many integration challenges will arise:– Can we simultaneously drive efficient LHCD at r/a~0.8, sustain core

and/or edge barriers, control impurities and handle heat loads?

The features which make advanced scenarios unique and challenging on C-Mod are the same as will need to be addressed on ITER.

Much cheaper and quicker to learn to deal with them now!