Advanced Tokamak Program: Recent progress and near-term plans Program Advisory Committee Review February 2, 2004 MIT PSFC Presented by A. Hubbard MIT Plasma Science and Fusion Center, for the team
Advanced Tokamak Program Overview of program Results, modeling and near-term plans in Current profile control Internal transport barriers Non-inductive, Long pulse operation Summary of program outlook and relation to US and world research program.
Advanced Tokamak Attractiveness of the tokamak as an ultimate fusion reactor increases if it is steady state and produces power at lower cost. Steady state implies non-inductive current drive. Lower cost implies much of the current drive is self-generated, ie. High bootstrap fraction. (external drive is expensive, in $ and Watts!). Also want high confinement, and high β for lower cost (all are inter-related). Research aimed at demonstrating these features and optimizing the tokamak configuration is a key part of the US fusion research program and of ITER planning. Will they be achievable in ITER? In CTF and/or DEMO?
Unique Features of Active advanced tokamak programs on several tokamaks (eg. DIIID, JT60U, AUG, JET, others). What will contribute that is new? 1. In physics terms, Steady-state current drive implies pulse lengths >> current relaxation time τ CR. Pulse Length/Relaxation Time 1.4a κt τ CR = Z 2 3/2 e eff T e =6 kev (ITER 19 kev) Z eff =1.5 JT60-U (30s) JET (50 s) ITER (1000 s) AUG (10s) DIII-D (10s) (5s) can run 5 second pulses, vs τ CR ~ 0.2-1.4 s [Z eff =1.5; T e = 2-7.5 kev ] (Major Radius) -2 (m -2 ) 2. Most AT expts have T i > T e, τ e-i > τ E and use NBI for core fuelling and rotation drive in ITBs. Reactor scenarios have τ e-i << τ E (T e >T i ), no core fuelling, no momentum input, RF heating and CD. can test feasibility of AT scenarios with all these features simultaneously.
Update on Profile Control Tools 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) Tools available or under development: Current profile: Lower Hybrid Current Drive. (Phase I 2005. Phase II ~2007). 4 MW, 4.6 GHz, 2 launchers with independent phasing, N //. Mode Conversion Current Drive. (initial tests were promising) Bootstrap current drive via pressure profile control. Density profile. Control of core transport, peaking. Cryopump controls edge source. (2006) D 2 and Lithium pellet injectors. Temperature Profiles 8 MW ICRH, 40-80 MHz, 2 independently variable deposition locations. LHCD. Control of core transport via RF deposition, magnetic shear. Shear Flow - MC flow drive (tests and modeling underway).
Main 5 year goals of the AT physics program 1. Demonstrate and model current profile control using LH and ICRF waves, at high densities (>10 20 m -3 ). 2. Understanding, control and sustainment of Internal Transport Barriers, with coupled ions and electrons, τ e-i << τ E (T e ~T i ) and without momentum input (RF only). Global confinement H 89P > 2.5 (Note: Details of understanding, mechanisms to be covered this year by Martin Greenwald in Transport talk tomorrow) 3. Achieve full non-inductive current drive (70% bootstrap) and extend pulse length to near steady state (5 sec, 4-6 τ CR ) - divertor power handling and wall particle issues, RF technology. 4. Attain and optimize no-wall β to MHD limit (β N > 3). Explore means of achieving higher values. Program involves all physics areas (waves, transport, plasma boundary, macrostability) and has broad participation from the team, including theory and modeling. More details in each of the topical science presentations tomorrow.
Current Profile Control: Lower Hybrid system begins operation March 1 1st LHCD grill installed at C-port. 4 x 24 guides. Designed for well controlled spectrum. Flexible N //, variable over range 2-4 between or during discharges using phase shifters. 12 klystrons, 3 MW source (coupled power will be lower). Following PAC advice, we plan to get operating experience in 2005 with this grill before fabricating 2 nd. Will allow two different N //, utilization of full 3 MW. 4 more klystrons would raise source to 4 MW total. LH commissioning and testing is critical to AT program. Therefore have a separate LH topical area this year, plans will be covered by Randy Wilson tomorrow. Aim to establish, study current drive by end of 2005. Hope to use as a j(r) control tool in 2006.
Suitable target plasmas for initial LHCD experiments have been made, modeled New experiments, Optimized parameters Experimental target profiles Sensitivity studies How should target be improved? Modeling w RF, LHCD 2003 experiments demonstrated steady, lower density H-modes. Can also use high power (hot) L-modes, exploiting higher power threshold with inner wall limiter or upper null. Total Power (MW) 5 4 3 2 1 L-H Power Threshold USN DN LSN IW Lim Lower Lim Hubbard, EPS 2004 0 0.0 0.5 1.0 1.5 2.0 2.5 n - e (10 20 m -3 )
Low(er) density H-modes as LH target 5.4 T, 600 ka, steady EDA H-mode n e T e n ped ~1.3x10 20 m -3. ψ ψ ACCOME predicts: I LH 131 ka, r/a 0.5-0.8 I BS 234 ka 62% non-inductive (with 2 launchers) In 2005, will also explore forming lower density ITBs starting from such H-modes (primarily to study turbulence, but would make good AT target plasma). (Note: throughout this talk, RED indicates specific planned near-term experiment from ideas forum or MP; not a complete list!). J (A / m 2 ) r / a J_tot J_oh J_bs J_lh
Mode Conversion Current Drive Current can also be driven by modeconverted IBW and/or ICW waves. Remarkable progress in measuring and modeling the waves in these conversion processes! (to be covered in talks by Paul Bonoli, Steve Wukitch) This means we can now use models with high confidence. ctr-cd co-cd P RF (MW) 2 1 D-port D+J-port D-port D+J-port T e0 (kev) 3.5 3 0.7 0.75 0.8 0.7 0.75 0.8 Time [sec] Time [sec] 8 T MC near q=1 Indications of some local current drive in recent experiment, to be followed up by Waves group in 2005. If successful, the AT program will then explore scenarios using this technique, eg: High l i discharges with near axis MCCD High bootstrap fraction discharges with off-axis MCCD TORIC Prediction for On-axis MCCD (5.4 T, 50+80 MHz) I MCCD = 100 ka for P ICRF = 3 MW j (ka/m 2 ) 80 60 40 20 0 Total driven current for 1 MW IBW + ICW 33 ka/mw IBW Contribution (x10) Ohmic profile current for modeled discharge (EFIT) -20 0.0 0.2 0.4 0.6 0.8 1.0 r/a
Critical diagnostics for profile control As recognized by the PAC last year, diagnostics of the current profile, and of the driven fast electrons, are critical to current profile control program. Good progress this year in routine operation of DNB+MSE diagnostic. However, still ongoing issues regarding calibration and interpretation. Not yet reliable j(r), further tests, better calibrations planned. Longer pulse, higher performance beam to be delivered in February. In parallel, continuing development and testing of polarimeter. Also commissioning new diagnostics for fast electrons: Hard x-ray camera (student project, collaboration with ToreSupra), for localization of fast electrons. ECE diagnostics, to add information on suprathermal energy (new student project). Diagnostic plans to be covered in more detail by Jim Irby Collaboration with Craig Petty (GA) on methods to extract maximum information on driven current.
Control of core transport, profiles (Internal Transport Barriers) 2004 experiments continued to study and improve ITB s triggered by off-axis ICRH, at r/a ~0.5, in plasmas without reversed shear. Core barriers co-exist with edge pedestal (EDA H-mode.) Also seen in ohmic H-mode. Increased power up to 3 MW offaxis, 1.8 MW on-axis (earlier limit ~0.6 MW). Central power still tends to increase transport. Its level is used to control impurity rise. There does not seem to be an absolute limit, rather dependence on relative powers. (Good news!) Fiore, EPS 2004 (invited), APS 2004 We can control the degree of energy and particle transport within the barrier!
High pressures (n and T) achieved 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 ) 0.20 0.15 0.10 0.05 0 0.0 0.2 0.4 0.6 0.8 1.0 r/a 30 20 10 1.5 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 0 0.0 0.2 0.4 0.6 0.8 1.0 r/a 20-3 Electron Density (10 m ) Electron Temperature (kev) 5 4 3 2 1 0 2.0 1.5 1.0 0.5 0.0 n e TS data t=0.761s H-mode t=1.094s ITB t=1.276s ITB with on-axis heating 1040309029 0.70 0.75 0.80 0.85 0.90 Midplane Major Radius [m] 1040309029 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 0.70 0.75 0.80 0.85 0.90 Midplane Major Radius [m] Planned experiments: Further optimize regime. Explore lower density ITBs (better AT target).
The ITB foot location expands with decreasing toroidal magnetic field, increasing plasma current Foot Position (r/a) 0.6 0.5 0.4 0.3 0.2 0.4 0.6 0.8 1.0 1.2 1.4 Plasma Current (MA) Fitting data from 4.5, 5.4, 6.3 T at different I p suggests that the ITB radius decreases linearly with increasing q; Plan to try 2.8 T @50 MHz Expect wider barriers 4 T < B t < 4.6 T 6.3 T < _ B t n sqrt [Z ] (/m 3 e eff ) The foot position can be well inside the region where the ICRF power is deposited (TORIC calculation). Ip=1.22MA Ip=1.12 MA Ip=0.88 MA r/a rf power distribution
Transport control/barrier research will move towards exploring effects of shear, flow To date, we have been studying plasmas with normal shear in contrast to most other experiments. As current profile control tools (LH and MC) become available, we will focus on studying effects of shear on transport, ITB formation and properties, eg. Flat ( hybrid ), reversed, current hole. If RF results warrant, may also begin to explore effects of flow drive (currently rather uncertain if levels will be large enough). Based on experience elsewhere, expect the parameter range for ITBs to expand, location to be more controllable. Hope is that by using our variable and well-controlled RF heating locations, we will still be able to control degree of transport reduction, find optimum levels of ion, electron, particle transport. All these experiments will be in Burning Plasma-relevant regime Coupled electrons and ions, no momentum or particle input, pulse length > τ CR. Our challenges, and results, will be of great interest for planning ITB or hybrid regimes on ITER!
Non-Inductive Current Drive, Long Pulse As our current profile control capability (LHCD + MCCD) matures (late 2005 through 2007), we will be able to drive increasing fractions of current non-inductively and extend the pulse length. C-mod has already run 3 secs, could go to 5 secs at 5 T. Short τ CR, cannot count on rampup to freeze in current profile. (This will also be true of any steady state experiment.) Will need to combine high CD efficiency with high bootstrap fraction (ie. Good confinement). Aim to achieve at least 50% non-inductive current with 1 st LH antenna. This year, plan to gain experience with non-inductive issues, operation, models through High Bootstrap Fraction Experiments at high ICRF power, I p rampdown; collaborations with Politzer, GA and Kessel, PPPL.
Plan to progressively explore several scenarios Hybrid Scenario: Regime features low shear, q ~ 1, moderate bootstrap fraction, density peaking and core confinement. Joint ITPA proposal (Sips et al, SSO-2.1) to compare with other tokamaks; high interest for extrapolating regime to ITER, therefore also in Burning Plasma task force. (An example of many connections between AT and BP). Low density H-modes (model results shown previously) Internal Transport Barriers, both L and H-mode edges. As mentioned, expect operating regimes to increase once shear is controlled.
Safety Factor - q(r) Example of long-term, non-inductive AT target scenario 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.
Integrated Scenario Modeling. Modeling is critical to assess wave accessibility, damping, and CD efficiency, and guide experiments toward more optimal scenarios. Multiple tools available, with several knobs, eg: 2 ICRF frequencies (1 variable), minority and mode conversion LHCD, variable N // (eventually two components). Density control. Simply too time-intensive to explore operating space by trial and error Modeling needs are complex, push state of art Need to consider time evolution to steady state, MHD stability. Complex relationships between transport and RF. Currently available models include: ACCOME: LH, ICRH and bootstrap I p. Consistent MHD equilibria. No transport. CQL3D (R. Harvey): Self-consistent 2-D velocity space Fokker-Planck. (20-30% higher LHCD than ACCOME). TRANSP: Time dependent, predictive or simulation mode. LSC for LHCD. Limitations in reverse N //, 1 poloidal ray. TORIC: Full wave field solver for ICRH, mode conversion, and LH.
Modeling needs, plans Past few years have seen breakthroughs in wave-plasma modeling, largely through SciDac (details in talks by Paul Bonoli, Steve Wukitch). Eg. Mode conversion now understood, predicted in detail. (TORIC). 2-D Focker Planck codes predicting higher LHCD than ACCOME, difference benchmarked through ITPA, and understood. (CQL3D) Starting to get FULL WAVE solutions to LH (TORIC, Bonoli APS Invited). Offers hope of understanding bridging of spectral gap. Challenge now is to incorporate improved RF models in timedependent transport simulation codes. Preferred option, since we are using TRANSP for both predictions and experimental analysis, is to include CQL3D module in TRANSP. Improvements in handling variable energy and particle transport are also needed. Also planning use of TSC. Similar issues with RF and transport models. Collaboration with PPPL, open to others!
Power Handling in Long Pulse AT In 2001, demonstrated 3 second pulses, but with P RF ~ 1 MW. Divertor/SOL power handing will be a major challenge as power, pulse length increase: Parallel flux was already up to 0.5 GW/m 2 with 3-4 MW ICRH! Inner divertor has already been upgraded. Requirement for fairly low edge n e (1-2x10 20 m -3 ) for LHCD makes radiative divertor difficult. experience will again be highly relevant to AT scenarios on ITER. Brief summary of plans (details will be in Plasma Boundary talk by Lipschultz) 2005-06: 6 MW ICRH + 1-2 MW LH (coupled) ~3 second pulses. Plan long pulse experiments using higher RF power. Added IR cameras to monitor LH antenna, hot spots Try strike point sweep. - Test tungsten brush modules. 2007 and beyond: 6 MW ICRH+3 MW LH Progress to 5 second pulses. Upgrade outer divertor, plus other areas as required.
Density Control Density is critical for LHCD accessibility, efficiency and deposition profile! As discussed previously, we have demonstrated L and H-mode target plasmas suitable for initial studies of LH current drive. However, we still see a cryopump as needed for optimizing density, achieving our ambitious non-inductive scenarios. Based on earlier measurements of neutral pressure vs S sep, we are designing a cryopump in upper divertor, expect good pumping for discharges close to double null. This shutdown, we installed a baffle to simulate that with the cryopump. Will measure effect on neutral pressures, including shaped plasmas for AT, compare with prior results. Will also assess impurity accumulation, wall saturation effects as part of long pulse experiments. Now plan to install cryopump in 2006 details in presentations by Bruce Lipschultz and Jim Irby.
Advanced Tokamak program is aligned very well with US priorities and ITER needs From Baker s APS presentation on FESAC panel on US fusion program priorities (preliminary): Topical Questions: T1: How does magnetic structure affect fusion plasma confinement? (includes shear) T3: How can external control and plasma self-organization be used to improve fusion performance? T11: How do electromagnetic waves interact with plasma? Recommended areas of US opportunity for enhanced progress : (ie. Top 6 Priorities for more resources!) Integrated understanding of plasma self-organization and external control, enabling high-pressure sustained plasmas. Extend understanding and capability to control and manipulate plasmas with external waves. C-mod is positioned to make key, unique contributions in these areas!
AT directly addresses ITER needs, as reflected in ITPA high-priority tasks Steady State Operation: Investigate hybrid scenarios for prolonged plasma operation and develop full current drive plasmas with significant bootstrap current: assess beta limits. Develop real time current profile control using heating and CD actuators: assess predictability, in particular for off-axis CD. 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... Contribute to and utilize international experimental ITPA database for tests of the commonality of hybrid and steady state scenario transport physics across devices. This alignment is no accident; The features which make advanced scenarios unique and challenging on are the same as will need to be addressed in a burning plasma. Much cheaper and quicker to learn to deal with them now!
Summary is poised to enter a key new phase of its advanced tokamak program, by demonstrating RF tools for current profile control. Commissioning and learning to use LHCD is top priority this year. Details of this activity will be given by Randy Wilson tomorrow. Continue to make good progress in understanding and optimizing core transport barriers produced by localized ICRF. Achieved higher power, and central n, T. (physics details to be given by Martin Greenwald). As j(r) control becomes available, we will explore influence of shear on core transport and barriers. Program will move progressively towards integrating our various tools to produce high bootstrap fraction, largely non-inductive, long pulse plasmas. Modeling, incorporating latest wave-plasma and transport understanding, is key.