1 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
2 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.
3 An advanced tokamak device is, in terms of magnetic configuration, simply a TOKAMAK Pure toroidal field does not confine charged particles Adding poloidal field does confine charged particles. Produced by toroidal current. Tokamak needs a toroidal current for stability. Current conventionally driven by tranformer; - Current is driven around central solenoid. Inherently NOT steady-state.
4 Tokamaks lead other configurations in fusion performance, are approaching breakeven ITER D. Meade, ARIES workshop 4/24/05
5 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!)
6 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.
7 OPTIMIZATION OF THE TOKAMAK CONCEPT LEADS TO AN ATTRACTIVE FUSION POWER PLANT Attractive features Improved power cycle Improved economics Reduced size Higher pressure, reduced heat loss Conventional Optimized Power cycle Pulsed Steady-state COE /kwhr ~13 ~7 Major radius (m) 8 5 The U.S. ARIES system study Central Solenoid PF Coils Vacuum Vessel Door Superconducting TF Toroidal CoilsMagnets Cryostat Maintenance Port Optimization of the tokamak concept is known as the Advanced Tokamak program Vacuum Vessel Low Temperature Shield High Temperature Shield Low Activation First Wall and Blanket Hardback Structure Divertor Diverter Region /TST/wj
8 THE GOAL OF THE ADVANCED TOKAMAK PROGRAM IS TO OPTIMIZE THE TOKAMAK CONCEPT FOR ATTRACTIVE FUSION ENERGY PRODUCTION Key Elements Discovering the Ultimate Potential of the Tokamak Steady state High self-generated bootstrap current Compact (smaller) Improved confinement (reduced heat loss) Fusion Ignition Requirement m 3 kev s < n T i τ (H a B κ) 2 H = τ E /τ conv E } Size High power density Improved stability P Fus (n T) 2 Vol β 2 B 4 Vol β = 2 µ o P B 2 DIII D NATIONAL FUSION FACILITY SAN DIEGO /TST/wj
9 SIMULATIONS PREDICT SELF-CONSISTENT EQUILIBRIA WITH NEARLY 100% BOOTSTRAP f BS = 0.92 J B BS 1 0 Radius DIII D NATIONAL FUSION FACILITY SAN DIEGO J B tot q Negative Magnetic Shear 1 Steady state with low recirculating power Off-axis current drive to supply missing current Provided by high power microwaves in DIII D Other benefits of negative central shear profile Reduced transport, improved confinement Improved stability to central unstable MHD modes Ballooning Tearing modes Sawteeth /TST/wj
10 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 Negative magnetic shear contributes to reduced γ ITG Similar reduction is often observed in other transport channels 10 kev ρ 6 T i 5 ω E B = 2 1 γ ITG 10 5 s ρ 10 kev 0.9 s 0.9 s s T i ρ DIII D NATIONAL FUSION FACILITY SAN DIEGO /TST/wj
11 A COMPACT STEADY STATE TOKAMAK REQUIRES OPERATION AT HIGH β N 2 P fus γ ε cur Q ss = eff β N B 3 aκ P CD nq ( 1 ξ A q β ) N β Power Density T ε 1 + κ 2 2 DIII D NATIONAL FUSION FACILITY SAN DIEGO Current Limit q* = 4 Advanced Pressure Conventional Tokamak Limit Tokamak β N = 5 β N = 3.5 Equilibrium Limit 2 εβ p Bootstrap Current High power density high β T Large bootstrap fraction high β p Steady state high β N β N power density bootstrap current ( 1 + κ 2 β T β p 2 )β N β N = β T /(I/aB) /TST/wj
12 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
13 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.
14 Evolution of ARIES Designs 1 st Stability, Nb 3 Sn Tech. High-Field Option Reverse Shear Option ARIES-IA ARIES-I ARIES-RS ARIES-AT Major radius (m) β (β Ν ) Plasma pressure/magnetic p 2% (2.9) 2% (3.0) 5% (4.8) 9.2% (5.4) Peak magnetic field (T) Avg. Wall Load (MW/m 2 ) Current-driver power (MW) Recirculating Power Fraction Thermal efficiency Cost of Electricity (c/kwh) Approaching COE insensitive of power density
15 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) Mid 80's Physics Early 90's Physics Late 90's Physics Advance Technology Mid 80's Pulsar Early 90's ARIES-I Late 90's ARIES-RS 2000 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/m 2 COE 5 c/kwh
16 ARIES-AT is Competitive with Other Future Energy Sources Estimated range of COE (c/kwh) for 2020* Natural Gas Coal Nuclear Wind (Intermittent) Fusion (ARIES-AT) AT 1000 (1 GWe) AT 1500 (1.5 GWe) EPRI Electric Supply Roadmap (1/99): Business as usual Impact of $100/ton Carbon Tax. * Data from Snowmass Energy Working Group Summary. Estimates from Energy Information Agency Annual Energy Outlook 1999 (No Carbon tax). Annual Energy Outlook 2005 (2025 COE, 2003$)
17 Advanced Tokamak Research on current experiments What is needed? key issues What results have already been obtained? Near-term plans and prospects.
18 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, β N. 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 τ E, current relaxation time τ CR.
19 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.
20 Advanced Tokamak Goal is Steady-state Operation Combined with High Fusion Performance Steady state operation 100% non-inductive current High β P, high fraction of bootstrap current High fusion gain High β, high τ E High normalized fusion performance: G = β N H 89 /q 95 2 Fusion power 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
21 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
22 High Normalized Beta (β N ~4) Sustained for ~2 s at High Safety Factor and High Confinement β N > 6l i for ~2 s Relies on wall stabilization of the n=1 external kink mode (conventional stability limit ~4l i ) High performance phase generally terminated by current profile evolution (m,n) = (2,1) tearing mode
23 High β Discharge Profiles Show NCS and ITBs Strong gradients typical of ITBs are observed in T i, n e and rotation profiles, but not in T e Pressure peaking factor, P(0)/<P> varies in range during high beta phase P(0)/<P>=2.9 f GW ~
24 High β N Discharge with Constant Plasma Current Driven Noninductively for ~0.5 s Surface voltage < 0 after I p ramp ends Internal loop voltage profile shows noninductive current fraction 100 %, although not fully relaxed 100 ms triangular smoothing Current drive analysis
25 With Improved Confinement, f ni =100% Achieved with Good CD Alignment 2 Local toroidal current density (A/cm ) J φ (R) Midplane major radius, R (m) MSE Array Tangential Radial Edge 2.4 Flux Surface Averaged Toroidal Current Density (A/cm 2 ) J(ρ) J ind Measurement J tot (Eq. Measurement) RADIUS, ρ Equilibrium measurement: J ind = neo E neo pol / t f NI = 1 f ind f ind = 0.5%, f NI = 99.5% Inductive current is locally & globally close to zero NI current aligned well to desired J tot good CD alignment T = 3.5%, N = 3.6, q 95 = 5.0 G = N H 89 /q 2 95 = 0.3 ITER steady state scenario requirements satisfied /mm/jy
26 Transport Code Carries Out Data Analysis Based on Equilibrium Reconstruction with Kinetic Profile Information Local toroidal current density (A/cm 2 ) J φ (r) J (calc.) EC Midplane major radius, R (m) MSE Array Tangential Radial Edge Analysis (EFIT) 2.4 Flux Surface Averaged Toroidal Current Density (A/cm 2 ) J(ρ) J NB J EC J bs J tot Analysis F Radius, ρ Measurements: f ind = 0.5%, f NI = 99.5% Analysis shows: f BS =59% f NB =31% f EC = 8% f NI = 98% Equilibrium reconstruction (EFIT) lacks spatial resolution Makes the current balance calculations problematic /mm/jy
27 Internal transport barriers, and core transport control, have been produced in C-Mod with normal shear, by varying heating profile OFF-axis heating alone causes density peaking, ~ const T. ON-axis heating clamps n, but increases T, neutron rate. Electron Pressure (MPascals) 3 RF Power Density (MW/m ) r/a MW central ICRF added into fully formed ITB t=1.294 s ITB, 2.35 MW Off-axis ICRF t=1.127s H-mode, No ITB t=0.894 s On-axis + off-axis, 4 MW total rf power at t=1.3 s Off-axis alone, 2.3 MW total rf power at t=1.1 s r/a 20-3 Electron Density (10 m ) Electron Temperature (kev) n e TS data t=0.761s H-mode t=1.094s ITB t=1.276s ITB with on-axis heating Midplane Major Radius [m] t=0.794s H-mode t=1.094s ITB t=1.261s ITB with added central rf power TS data Edge Thomson GPC2 data FRC data T e Midplane Major Radius [m] Levels of heat and particle diffusivity can be reduced to neoclassical, or increased to stabilize density and impurities.
28 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 < τ CR ), 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. Uncoupled (τ e-i >> τ Ε ) vs coupled (τ e-i << τ Ε ) 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.
29 Experiments at Higher q min and Higher β N Will Address Steady-state Demonstration New divertor and improved density control will slow down q min evolution By allowing higher temperature at lower density By allowing higher ECCD at lower density Additional ECCD power will improve current profile control Higher β N at higher q min will give higher bootstrap current Will reduce Ohmic current at large radii Will overdrive at small radii Compensate overdrive using ECCD, FW, Counter NBI
30 Advanced Tokamak Research on C-Mod AT research is an increasingly important focus on C-Mod, which is a compact (R=.68 m), high B (5-8 T), high n e ( m -3 ) 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) Strongly coupled ions and electrons (τ e-i << τ Ε ) 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 β N ~3. Longer term, would like to add active stabilization.
31 New LHCD system on Alcator C-Mod 12 klystrons 0.25 MW = 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!
32 Safety Factor - q(r) Example of non-inductive AT target scenario on C-Mod One of many optimized scenarios modelled with ACCOME. I LH =240 ka I BS =600 ka (70%) Ip = 0.86 MA Ilh = 0.24 MA fbs = 0.7 Double transport barrier B T =4 T ICRH: 5 MW LHCD: 3 MW, N //0 =3 n e (0)= 1.8e 20 m -3 T e (0)=6.5 kev (H=2.5) β N =2.9 Scenarios without barrier, or only an ITB, have similar performance. J (MA / m2) q(0) = 5.08 q min = 3.30 q(95) = 5.98 r / a r / a P. Bonoli, Nucl. Fus. 20(6) 2000.
33 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 B T =5.3 T. Still some flexibility/uncertainty in sources, parameters. 1. Conventional H-mode: Baseline Scenario, Q=10.Positive shear, q 95 =3, β N =1.8, H H ~1, n~10 20 m -3. f NI ~ MAIN ITER GOAL! 2. Hybrid Scenario: Q=10 Weak core shear, q 95 =4, q min ~1, β N =2.8, H H ~1.2, f NI ~ 0.5. j(x) 3. Steady-state: Q=5, long pulse Weak or negative shear, q 95 ~5, q min ~2, β N =3.0, H H ~1.2, f NI ~ 1.0. SECONDARY GOAL TRANSP/TSC simulation of ITER S-S scenario. Houlberg&ITPA, IAEA04.
34 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 BPrelevant 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.
35 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.
36 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.
37 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
38 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.
39 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.
40 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.
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