The Advanced : Goals, prospects and research opportunities Amanda Hubbard MIT Science and Fusion Center with thanks to many contributors, including A. Garafolo, C. Greenfield, C. Kessel, D. Meade, M. Murakami, F. Najmabadi, T. Taylor Opinions are my own…

GCEP Fusion Energy Workshop on Opportunities for Fundamental Research and Breakthrough in Princeton, NJ May 1-2 2006 The Advanced Tokamak

• Introduction: What is an ‘advanced tokamak’?

• The AT vision for fusion energy – Drawing heavily on ARIES studies. • Current results and near-term prospects – Focusing here on US program.

• AT on ITER: What we will (and won’t) learn.

• Research Opportunities: ideas to advance and accelerate fusion energy prospects. To start the discussion…. An “advanced tokamak” device is, in terms of magnetic configuration, simply a TOKAMAK

Pure toroidal field does not confine charged particles

Tokamak needs a toroidal current for stability. Adding poloidal field does Current conventionally driven by tranformer; confine charged particles. - Current is driven around central solenoid. Produced by toroidal current. Inherently NOT steady-state. lead other configurations in fusion performance, are approaching ‘breakeven’

ITER

D. Meade, ARIES workshop 4/24/05 “Conventional” tokamak operation will be primary mode of operation on ITER

• Heating applied mainly on-axis, inductive current drive, profiles relax to ‘natural’ state. • Much experience worldwide, good confidence in extrapolation to burning plasma conditions. – This will allow critical exploration of burning plasma physics.

• Could probably be used to make a plant. – Advantages of relative simplicity, staying away from performance limits. – BUT projected power plant not seen as economically attractive (at least in prior assessments with low cost oil!) Tokamak current does not have to be driven by a transformer!

• Alternative means of current drive: – External current drive, by neutral beams, or microwaves (various ranges from ion cyclotron (~100 MHz), Lower Hybrid (~ 5 GHz), electron cyclotron (~100 GHz))

–“Bootstrap current”: Self-generated current due to temperature, density, pressure gradients in the plasma.

• All of these are fairly well understood, and have been demonstrated to work on many experiments. – Gives potential for steady-state operation. • “The crucial distinguishing feature of an Advanced Tokamak over a conventional tokamak is …the use of active control of the current or shear profile, and of the pressure profile or transport characteristics” (AT Workshop, GA, 1999) – Same tokamaks can (and do) operate in both conventional and advanced regimes. OPTIMIZATION OF THE TOKAMAK CONCEPT LEADS TO AN ATTRACTIVE FUSION POWER PLANT

Optimized Attractive features Conventional — Improved power cycle Power cycle Pulsed Steady-state — Improved economics COE¢ /kWhr ~13 ~7 — Reduced size Major radius (m) 8 5 Higher pressure, reduced heat loss

The U.S. ARIES system study

Central Solenoid Superconducting Toroidal Magnets Central Solenoid TF Coils PF Coils Cryostat PF Coils Cryostat VacuumVacuum VesselVessel Optimization of the tokamak Door concept is known as the Maintenance Maintenance Port Port Advanced Tokamak program

Vacuum Vessel

VacuumLow Temperature Vessel Shield First Wall and Blanket High Temperature Shield Hardback Structure Low Activation FirstDiverter Wall and Region Blanket

High Temperature Shield Hardback Structure

Divertor Region

130Ð02/TST/wj AINLFSO FACILITY FUSION NATIONAL DIII Key Elements A DIEGO SAN THE GOAL OF ADVANCED TOKAMAK PROGRAM – D Compact (smaller) Steady state High power density — — — IS TO OPTIMIZE THE TOKAMAK CONCEPT FOR — ATTRACTIVE FUSION ENERGY PRODUCTION Discovering the Ultimate Potential of Tokamak P Fusion Ignition Requirement Improved confinement (reduced heat loss) High self-generated bootstrap current Improved stability Fus 3 H = × 10

∝ τ 21

E (n T) / m τ E conv Ð 3 2 keV s Vol

∝ < β n T 2 B i

4 τ

Vol ∝ ∼

(H a B Size } κ ) 2

β

= 2 µ B o 2

〈 P — 〉 130 Ð 02/TST/wj SIMULATIONS PREDICT SELF-CONSISTENT EQUILIBRIA WITH NEARLY 100% BOOTSTRAP

4 Steady state with low fBS = 0.92 recirculating power 〈J¥B〉tot Off-axis current drive 3 to supply missing current — Provided by high power microwaves in DIIIÐD 2 q

Other benefits of negative 〈J ¥B〉BS 1 Negative central shear profile 0 Magnetic Shear 1 Radius — Reduced transport, improved confinement — Improved stability to central unstable MHD modes

Ballooning Tearing modes Sawteeth DIII–D NATIONAL FUSION FACILITY 130-02/TST/wj SAN DIEGO 10

0 2 4 keV 6 8 AINLFSO FACILITY FUSION NATIONAL DIII . . . . 1.0 0.8 0.6 0.4 0.2 0 . 0.9 s 0.9 s A DIEGO SAN

— Key physics — Similar reduction is often observed in other transport channels NEGATIVE CENTRAL SHEAR AND SHEARED E – Negative magnetic shear contributes to reduced Measured turbulence reduction is consistent with theoretical prediction D ρ E LEAD TO IMPROVED CORE CONFINEMENT ×

B shearing rate exceeds maximum growth of ion temperature gradient mode T i +=

105 s-1 0 1 2 3 4 5 6 . . . . 1.0 0.8 0.6 0.4 0.2 0 | ω E × B | γ ρ ITG γ ITG 10

0 2 4 keV 6 8 . . . . 1.0 0.8 0.6 0.4 0.2 0 × B FLOW 130 1.2 s Ð

ρ 02/TST/wj T i A COMPACT STEADY STATE TOKAMAK REQUIRES OPERATION ATβ HIGH N γ ε β2 Pfus eff N 3 Q = ∝ cur B aκ ss P nq (1 Ðξ √ qβ ) CD A N

High power density ⇒ β high T P Large bootstrap fraction r e Power Density Power s ⇒ highβ A s p d u v r a e ⇒ β Current Limit n L Steady state high β c i N T e m d it β ∝ power density× 2 To N 1 +κ C ka ε o ma bootstrap current n k Equilibrium Limit v e 2 n t β io = 5 n N 2 al 1 +κ 2 To β β ∝ β kam T p N ak ( 2 ) = 4 β N = 3.5 q* β = β 2(I/aB) εβ Bootstrap Current N T / p DIII–D NATIONAL FUSION FACILITY 130Ð02/TST/wj SAN DIEGO Advanced Tokamak concept of fusion power plant.

• Embodied in ARIES design studies, ARIES-RS and ARIES-AT. Japan has similar studies.

• Material courtesy of F. Najmabadi, UCSD ARIES-AT is an attractive vision for fusion with a reasonable extrapolation in physics & technology

∗ Competitive cost of electricity (5c/kWh); ∗ Steady-state operation; ∗ Low level waste; ∗ Public & worker safety; ∗ High availability. Evolution of ARIES Designs

1st Stability, High-Field Reverse Shear Option Nb3Sn Tech. Option

ARIES-IA ARIES-I ARIES-RS ARIES-AT

Major radius (m) 8.0 6.75 5.5 5.2

β (βΝ) Plasma pressure/magnetic p 2% (2.9) 2% (3.0) 5% (4.8) 9.2% (5.4) Peak (T) 16 19 16 11.5

Avg. Wall Load (MW/m2) 1.5 2.5 4 3.3

Current-driver power (MW) 237 202 81 36

Recirculating Power Fraction 0.29 0.28 0.17 0.14

Thermal efficiency 0.46 0.49 0.46 0.59

Cost of Electricity (c/kWh) 10 8.2 7.5 5

Approaching COE insensitive of power density Our Vision of Magnetic Fusion Power Systems Has Improved Dramatically in the Last Decade, and Is Directly Tied to Advances in Fusion Science & Technology

Estimated Cost of Electricity (c/kWh) Major radius (m)

10 14 9 12 8 7 10 6

8 5 6 4 4 3 2

2 1

0 0 Mid 80's Early 90's Late 90's Advance Mid 80's Early 90's Late 90's 2000 Physics Physics Physics Technology Pulsar ARIES-I ARIES-RS ARIES-AT Present ARIES-AT parameters: Major radius: 5.2 m Fusion Power 1,720 MW Toroidal β: 9.2% Net Electric 1,000 MW Wall Loading: 4.75 MW/m2 COE 5 c/kWh ARIES-AT is Competitive with Other Future Energy Sources Estimated range of COE (c/kWh) for 2020*

7 AT 1000 (1 GWe) 6 AT 1500 (1.5 GWe) 5 4 3 2 1 0

Natural Gas Coal Nuclear WindFusion (Intermittent)(ARIES-AT) EPRI Electric Supply Roadmap (1/99): Estimates from Energy Information Agency Business as usual Annual Energy Outlook 1999 Impact of $100/ton Carbon Tax. (No Carbon tax). Annual Energy Outlook 2005 * Data from Snowmass Energy Working Group Summary. (2025 COE, 2003$) Advanced Tokamak Research on current experiments

• What is needed? – key issues

• What results have already been obtained?

• Near-term plans and prospects. Physics Requirements for Advanced Tokamak

•For STEADY STATE, want – 100% non-induction current drive (external + self-generated “Bootstrap”

•For low recirculating power, good economics, want – High “bootstrap fraction” 80-90% self-generated.

• To get this, need high normalized pressure, βN. This requires low transport, to get high gradients, which in turn are enabled by optimized current profile. High pressure itself improves economics. Highly coupled control of current, transport profiles needed – for times long compared to plasma time scales,

eg. energy confinement time τE, current relaxation time τCR. Many of these requirements have been demonstrated in present expts

As examples, show recent results from DIII-D tokamak, San Diego at 2005 APS-DPP meeting (A. Garafolo, Univ. Columbia, M. Murakami, ORNL) and from C-Mod, MIT DIII-D results rely heavily on MHD stabilization techniques to reach high β. This important aspect of AT research will be covered this afternoon by G. Navratil.

Other world tokamaks, in particular JT-60U (Japan) and ASDEX- Upgrade (Germany) also have strong AT programs, range of control tools. Will not attempt a comprehensive review here!

Also important work on advanced scenarios in spherical (low aspect ratio) tokamaks NSTX (PPPL) and MAST (UK), which will (I presume) be covered in talk by Martin Peng. Advanced Tokamak Goal is Steady-state Operation Combined with High Fusion Performance

• Steady state operation – 100% non-inductive current

– High βP, high fraction of bootstrap current

• High fusion gain r e

– High β, high τE w o

– High normalized fusion p

performance: n o 2 i G = βNH89/q95 s u F

• Negative Central Shear Bootstrap current – Stability to high-n ballooning modes and neoclassical tearing modes – Suppression of transport – Good alignment of bootstrap current with total current – Hollow current profile, wall-stabilization of low-n kink modes Recent DIII-D Experiments Achieved High Fusion Performance at High Bootstrap Current Fraction

• Combination of high confinement, high beta, and high bootstrap fraction sustained for ~2 s • Multiple control tools needed, including – Simultaneous ramping of plasma current and toroidal field – Simultaneous Feedback Control of Error Fields and Resistive Wall Mode • Transport analysis confirms presence of internal transport barriers (ITBs) in high β discharges • Stability analysis indicates potential for higher beta operation • High noninductive current fraction (~100%) has been achieved • Steady-state sustainment will be pursued with new DIII-D tools High Normalized Beta (βN ~4) Sustained for ~2 s at High Safety Factor and High Confinement

122004

• βN > 6li for ~2 s – Relies on wall stabilization of the n=1 external kink mode (conventional stability limit ~4li)

• High performance phase generally terminated by current profile evolution – (m,n) = (2,1) tearing mode High β Discharge Profiles Show NCS and ITBs

• Strong gradients typical of ITBs are

observed in Ti, ne and rotation profiles, but not

in Te • Pressure peaking factor, P(0)/

varies in range 2.6-3.2 during high beta phase • P(0)/

=2.9

• fGW~ 0.4-0.6 High βN Discharge with Constant Plasma Current Driven Noninductively for ~0.5 s

• Surface voltage

< 0 after Ip ramp ends • Internal loop voltage profile shows noninductive current fraction ≥100 %, although not fully relaxed

100 ms triangular smoothing

Current drive analysis With Improved Confinement, fni=100% Achieved with Good CD Alignment

2 200 MSE Array 150 (R) Tangential Measurement φ 2 〈J(ρ)〉 J Radial 150 Edge 100 Jtot

100 50

Jind 50 0

Current Density (A/cm ) Current Density (A/cm (Eq. Measurement) 0 Flux Surface Averaged Toroidal Flux Surface Averaged

Local toroidal current density (A/cm ) density (A/cm current Local toroidal 1.6 1.8 2.0 2.2 –50 2.4 0.0 0.2 0.4 0.6 0.8 1.0 Midplane major radius, R (m) RADIUS, ρ

• Equilibrium measurement: Jind = neoE||  neopol/t  fNI = 1 – find

• find = 0.5%, fNI = 99.5% • Inductive current is locally & globally close to zero

 NI current aligned well to “desired” Jtot  “good CD alignment” 2 • T= 3.5%, N = 3.6, q95 = 5.0  G = N H89/q95 = 0.3  ITER steady state scenario requirements satisfied 211-05/mm/jy Transport Code Carries Out Data Analysis Based on Equilibrium Reconstruction with Kinetic Profile Information 2 200 MSE Array 150 Analysis Tangential 〈 ρ 〉 2 J( ) Radial 120096F05 150 Edge Analysis (EFIT) 100 Jtot Jφ(r) Jbs 100 JNB 50 JEC 50 0

〈 〉 ) Current Density (A/cm J ECE C ((calc.)calc.) 0 Flux Surface Averaged Toroidal Flux Surface Averaged Local toroidal current density (A/cm ) density (A/cm current Local toroidal 1.6 1.8 2.0 2.2 –50 2.4 0.0 0.2 0.4 0.6 0.8 1.0 Midplane major radius, R (m) Radius, ρ

• Measurements: find = 0.5%, fNI = 99.5%

• Analysis shows: fBS=59% fNB=31% fEC= 8% fNI= 98% • Equilibrium reconstruction (EFIT) lacks spatial resolution  Makes the current balance calculations problematic

211-05/mm/jy Internal transport barriers, and core transport control, have been produced in C-Mod with normal shear, by varying heating profile

1040309029 • OFF-axis heating alone causes 5 density peaking, ~ const T. 4 ) -3 ne 20 • ON-axis heating clamps n, but m 3 increases T, rate. 2 0.20 TS data t=0.761s H-mode

Electron Density (10 t=1.094s ITB 1.5 MW central ICRF 1 added into fully formed ITB t=1.276s ITB with on-axis heating 0.15 t=1.294 s 0 ITB, 2.35 MW Off-axis ICRF 0.70 0.75 0.80 0.85 0.90 Midplane Major Radius [m] t=1.127s 1040309029 0.10 H-mode, No ITB 2.0 t=0.894 s TS data Edge Thomson GPC2 data 1.5 0.05 FRC data

Electron Pressure (MPascals) 0 1.0 T 0.0 0.20.40.6 0.8 1.0 e r/a

Electron Temperature (keV) 0.5 t=0.794s H-mode 30 t=1.094s ITB 3 On-axis + off-axis, 4 MW t=1.261s ITB with added central rf power total rf power at t=1.3 s 0.0 0.70 0.75 0.80 0.85 0.90 20 Off-axis alone, 2.3 MW Midplane Major Radius [m] total rf power at t=1.1 s

10 Levels of heat and particle diffusivity can be reduced to

RF Power Density (MW/m ) RF Power Density (MW/m 0 neoclassical, or increased to 0.0 0.20.40.6 0.8 1.0 stabilize density and impurities. r/a Issues and Near-term plans for Advanced Tokamak Research

• While much has been achieved, much more remains to be done to realize potential of advanced regimes on burning plasmas, and fusion reactors.

• Most scenarios are still non-stationary (t < τCR) , and/or rely on current profile control techniques (eg, tailored heating during current rampup, central NBI) which don’t extrapolate to steady state.

• Most present experiments have plasma conditions quite different from burning plasmas. Eg. e-i e-i – Uncoupled (τ >> τΕ ) vs coupled (τ << τΕ ) electrons and ions (lower vs higher density) – Core particle and momentum sources (vs RF, alpha htg.) – Both factors strongly affect transport barrier formation.

• Handling of high heat loads in divertor – common to all attractive configurations and will be covered in talk by Mike Ulrickson. Experiments at Higher qmin and Higher βN Will Address Steady-state Demonstration

• New divertor and improved density control will slow down qmin evolution – By allowing higher temperature at lower density – By allowing higher ECCD at lower density

• Additional ECCD power will improve current profile control

• Higher βN at higher qmin will give higher bootstrap current – Will reduce Ohmic current at large radii – Will overdrive at small radii • Compensate overdrive using ECCD, FW, Counter NBI Advanced Tokamak Research on C-Mod

• AT research is an increasingly important focus on C-Mod, which is a 20 21 -3 compact (R=.68 m), high B (5-8 T), high ne (10 -10 m ) tokamak. • Unique among world divertor tokamaks, can test AT physics and scenarios – At ITER field and density (key wave physics parameters). – Without core particle or momentum sources (all RF heating) e-i – Strongly coupled ions and electrons (τ << τΕ ) – Pulse lengths >> current relaxation times, routinely. (ie., steady-state, relaxed j(r)). – ITER-level divertor fluxes. Important challenge and test: Will AT regimes scenarios work as well in these conditions, typical of ITER and reactors??

• Program focuses on control of current and magnetic shear as well as transport and kinetic profiles with various shear profiles. – RF systems (ICRF +LHCD) provide key control tools. LHCD is highest efficiency technique for current drive far off axis. – Also adding new cryopump, important for density control. – In near term, rely on shape and profile control to maximize no-wall limit βN~3. Longer term, would like to add active stabilization. New LHCD system on Alcator C-Mod

12 klystrons × 0.25 MW = 3 MW @ 4.6 GHz

Transmission waveguides

Coupler grill – MIT/PPPL collab’n. 96 waveguide outlets, allowing flexible phasing to launch spectrum Initial experiments in progress and first, significant, LHCD recently seen! Example of non-inductive AT target scenario on C-Mod

• One of many optimized • Double transport barrier

scenarios modelled with •BT=4 T ACCOME. • ICRH: 5 MW

• LHCD: 3 MW, N//0=3 –ILH=240 kA 20 -3 •ne(0)= 1.8e m –I =600 kA (70%) BS •Te(0)=6.5 keV (H=2.5)

• βN=2.9 Scenarios without barrier, or only an Ip = 0.86 MA Ilh = 0.24 MA fbs = 0.7 ITB, have similar performance.

q(0) = 5.08 q(95) = 5.98

) qmin = 3.30 2

J (MA / m J (MA Safety Factor - q(r) - Factor Safety

r / a r / a P. Bonoli, Nucl. Fus. 20(6) 2000. Advanced Scenarios on ITER

As a burning plasma experiment, ITER will explore a range of physics parameters and scenarios. To guide planning, currently focusing

on three main target scenarios, all at BT=5.3 T. Still some flexibility/uncertainty in sources, parameters.

1. Conventional H-mode: “Baseline” j(x) Scenario, Q=10.Positive shear, q95=3, 20 -3 βN=1.8, HH~1, n~10 m . fNI~ 0.25. MAIN ITER GOAL!

2. “Hybrid” Scenario: Q=10 Weak core shear, q95=4, qmin~1, βN=2.8, HH~1.2, fNI~ 0.5. 3. Steady-state: Q=5, long pulse Weak or negative shear, q95~5, qmin~2, β =3.0, H ~1.2, f ~ 1.0. N H NI TRANSP/TSC simulation of SECONDARY GOAL ITER S-S scenario. Houlberg&ITPA, IAEA04. What we can (and can’t) expect from ITER

• Demonstration of advanced, high non-inductive scenarios on ITER would be an extremely important step towards an AT DEMO reactor! – Would resolve many uncertainties about applicability with BP- relevant plasma parameters and control tools, as outlined above. – Would start addressing key control issues with self-heating. • A key goal of the current US program is to conduct research that will support AT on ITER, and to push for needed hardware. BUT… • Because steady-state mission on ITER is secondary, it is NOT an optimized machine for AT. For example, – Shaping flexibility is limited. – Heating and current drive likely underpowered.

• Much depends on hardware decisions not yet made, eg. – Will ITER have LHCD, needed for off-axis CD? How much? – Will ITER have active control coils to reach highest β ? – US will be pushing for these, but we won’t call all the shots! Coming year or two is critical. Research Opportunities

• What are key questions/topics which would take advanced tokamak from interesting and attractive scientific research to a fusion energy source?

• Which are NOT likely to be funded in near-term US-DOE program?

• As a general principle, I assume that issues of direct application to ITER will likely get priority, and (hopefully) adequate funding.

• More general – but still important – issues or those aiming at steps beyond ITER, and fusion energy application, are likely to get less resources. “CONTROL” ISSUES TOP MY LIST

• A fusion reactor or DEMO would need to run for long periods, close to stability limits, and with all profiles well optimized and controlled. • In high-bootstrap scenario, these are tightly coupled – Current profile derived mainly from n, T profiles, which in turn depend on both sources and transport. MANY interactions! – Limited external control of j(r). How much is needed? – Pressure/current profiles need to be aligned for stability. • Heat mainly coming from fusion burn – reduces controllability.

• I would like to see a more focussed effort on demonstrating active, integrated profile control – not just ‘tailoring’ of profiles applicable to a specific machine or experiment (eg by adjusting heating times in rampup). – This likely won’t ‘just happen’, even with ITER coming. – Would benefit from an interdisciplinary approach, from plasma physics experimentalists and theorists, plus engineering, controls, power systems experts. Actuators and nonlinear couplings in a bootstrap-dominated steady state burning plasmas

Figure from P. Politzer et al., ITPA meeting Lisbon 2004 D. Moreau IEA W60 Burning Plasma Physics and Simulation, Tarragona, July 2005 Can we develop a transport control tool?

• Part of the difficulty in profile control is the indirect nature of controlling temperature and density profiles. – We control heat and particle sources and, if we are fortunate, current profiles. – Plasma transport determines T, n profiles. While there is good progress in understanding transport (to be covered in talks by Tynan, Dorland), it is highly complex, with gaps in our knowledge; not easily amenable to control algorithms! χ D have been shown to be affected by j(r), by heating profile, by shear flow… – The goal is NOT minimum transport, but optimum transport – otherwise pressure limits exceeded, impurities and ash accumulate.

• A “holy grail” of transport and control research is an active control tool for transport, independent of heat sources. Best hopes are for RF tools, which could eg. drive flow shear, modifying χ at specific location.

• There are ideas (eg, Ion Bernstein Waves), but currently not a focused experimental and theoretical effort. – Could I think be done with modest funding, including expt-theory collaboration and small-scale lab tests. – Would be high leverage for AT fusion development. Other pressing issues…

Most of these will I expect come up in other talks, and are active areas of international fusion research • Improved divertor solutions and materials for steady state, reactor-level heat fluxes. • Compatibility of core and edge plasmas in advanced modes (closely related to divertor issue) • MHD control tools for sustaining high beta, suppressing code instabilities. • Disruption avoidance and mitigation. • Extension to longer pulse lengths – EAST and KSTAR will play important roles, though experience suggests it will take several years to develop needed AT tools. Summary: The Advanced Tokamak

• The “advanced tokamak” is a tokamak operational scenario characterized by high degree of control of current and pressure profiles. • Leads to optimized fusion reactor designs (eg ARIES-AT, RS) which are steady-state, have low recirculating power and lower size and cost than conventional tokamaks. – Extrapolates to competitive cost of electricity. • Current experiments have demonstrated many key needed features, such as high β, reduced transport and non-inductive current drive. – Near-term research aims at extending such results to steady-state and demonstrating in more burning-plasma relevant conditions. • ITER will be an important test of advanced scenarios in a burning plasma! But, this is a secondary mission and the device may not have optimal design and tools. • Further research is needed to go from current experiments to confidence in an advanced tokamak DEMO reactor, in particular more tools, understanding and experience in active control.