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The Advanced Tokamak: Goals, Prospects and Research Opportunities Amanda Hubbard MIT Plasma Science and Fusion Center with Thanks to Many Contributors, Including A

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The Advanced Tokamak: Goals, Prospects and Research Opportunities Amanda Hubbard MIT Plasma Science and Fusion Center with Thanks to Many Contributors, Including A The Advanced Tokamak: Goals, prospects and research opportunities Amanda Hubbard MIT Plasma 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 Nuclear Fusion 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. Tokamaks 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 fusion power 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 Attractive features Conventional Optimized — 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 CentralCentral Solenoid Solenoid Superconducting TFToroidal Coils Magnets PFPF CoilsCoils Cryostat VacuumVacuum VesselVessel Optimization of the tokamak Door Door concept is known as the Maintenance Port Advanced Tokamak program VacuumVacuum Vessel Vessel Low Temperature Low Activation Shield First WallWall and and Blanket Blanket HighHigh Temperature Temperature ShieldShield Hardback Structure Structure DivertorDiverter RegionRegion 130–02/TST/wj THE GOAL OF THE ADVANCED TOKAMAK PROGRAM IS TO OPTIMIZE THE TOKAMAK CONCEPT FOR ATTRACTIVE FUSION ENERGY PRODUCTION — Discovering the Ultimate Potential of the Tokamak — Key Elements G Steady state — High self-generated bootstrap current G Compact (smaller) — Improved confinement (reduced heat loss) Fusion Ignition Requirement 21 –3 2 × < τ ∝ } κ 3 10 m keV s n Ti ∼ (H a B ) H = τ /τconv E E Size G High power density — Improved stability 2 µ 〈P〉 ∝ 2 ∝ β2 4 β o PFus (n T) Vol B Vol = B2 DIII–D NATIONAL FUSION FACILITY 130–02/TST/wj SAN DIEGO SIMULATIONS PREDICT SELF-CONSISTENT EQUILIBRIA WITH NEARLY 100% BOOTSTRAP 4 © Steady state with low fBS = 0.92 〈 〉 recirculating power J•B tot G Off-axis current drive 3 to supply missing current — Provided by high power microwaves in DIII–D 2 q 〈 〉 G Other benefits of negative J •B BS Negative central shear profile Magnetic Shear 1 — Reduced transport, improved 0 1 confinement Radius — Improved stability to central unstable MHD modes Ballooning Tearing modes DIII–D Sawteeth NATIONAL FUSION FACILITY 130-02/TST/wj SAN DIEGO NEGATIVE CENTRAL SHEAR AND SHEARED E×B FLOW LEAD TO IMPROVED CORE CONFINEMENT Key physics — Measured turbulence reduction is consistent with theoretical prediction # E×B shearing rate exceeds maximum growth rate of ion temperature gradient mode γ — Negative magnetic shear contributes to reduced ITG Similar reduction is often observed in other transport channels 10 6 10 Ti Ti 5 ω 8 | E×B| 8 4 1.2 s 6 -1 6 s 5 3 keV keV 0.9 s+= 0.9 s 4 10 4 2 γ ITG 2 1 2 0 0 0 0 0.2 0.4ρ 0.6 0.8 1.0 0 0.2 0.4ρ 0.6 0.8 1.0 0 0.2 0.4ρ 0.6 0.8 1.0 DIII–D NATIONAL FUSION FACILITY 130–02/TST/wj SAN DIEGO A COMPACT STEADY STATE TOKAMAK β REQUIRES OPERATION AT HIGH N γ ε β2 P eff 3 Q = fus ∝ cur N B aκ ss P nq (1 – ξ √ q β ) CD A N G High power density ⇒ β high T G P Large bootstrap fraction r e Power Density s ⇒ high β A s p d u v r a e G ⇒ 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 m bootstrap current n a Equilibrium Limit v k e 2 n t β io = 5 n N 2 al 1 + κ 2 To β β ∝ β kam T p N ak ( 2 ) β = 4 N = 3.5 q* 2 β = β (I/aB) εβ Bootstrap Current N T / p DIII–D NATIONAL FUSION FACILITY 130–02/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 magnetic field (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 6 AT 1000 (1 GWe) 5 AT 1500 (1.5 GWe) 4 3 2 1 0 Natural Gas Coal Nuclear Wind Fusion (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
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