Large Tokamaks Large Tokamaks Thomas J. Dolan ASIPP Hefei 2011 - PowerPoint PPT Presentation

High Power D-T Operations in JET 3.6 T, 3.6 MA NBI = 22.3 ICRF = 3.1  16.1 MW fusion 60% thermonuclear 40% beam-plasma Q DT ~ 0.62 Allowed < 2.5x10 20 neutrons dolan asipp 2011 Figures are from J. Wesson, 43 Tokamaks, 2004

Central Temperature vs. Alpha Power Heating of T e by P  good alpha confinement P  = 1.3 MW   T e (0) ~ 1.3 keV (% tritium) Open diamonds: ICRF heating in pure deutrium pure deutrium. dolan asipp 2011 Figures are from J. Wesson, 44 Tokamaks, 2004

JT-60U Tokamak, Naka, Japan dolan asipp 2011 Figures are from J. Wesson, 45 Tokamaks, 2004

JT-60U Tokamak, Naka, Japan Single null R/a = 3.4/1.1 4.2 T 4.2 T 5 MA NBI = 40 (120 keV) (120 keV) ICRH = 7 LH = 8 ECH = 3 Negative ion NBI Negative ion NBI under development dolan asipp 2011 Figures are from J. Wesson, 46 Tokamaks, 2004

JT-60U Operation Usually L-mode H-mode at low n, high P in H mode at low n, high P in Pellet injection, peaked profiles  better  E  better Graphite targets Radiative cooling in divertor He ash removal effective Boronization  f Boronization  f ox < 1% < 1% dolan asipp 2011 Figures are from J. Wesson, 47 Tokamaks, 2004

Limiter H-Mode in JT-60U dolan asipp 2011 Figures are from J. Wesson, 48 Tokamaks, 2004

Recycling Flux Decreases  E Recycling increase  P rad increase  P rad increase   E decrease Open circles have Open circles have no gas puffing. dolan asipp 2011 Figures are from J. Wesson, 49 Tokamaks, 2004

W-shaped divertor W shaped divertor Better pumping  P(H-mode) reduced 30% dolan asipp 2011 Figures are from J. Wesson, 50 Tokamaks, 2004

Helium residence time Achieved  He* /  E ~ 3-4, with f with f Heplasma ~ 4%. ~ 4% But with ITB  He* /  E ~ 15. dolan asipp 2011 Figures are from J. Wesson, 51 Tokamaks, 2004

Current Drive LHCD ~ 10 MW LHCD 10 MW  3.6 MA  CD = Rn I CD /P in  LHCD  LHCD = 0.34x10 20 A/m 2 W  NNBI = 0.16x10 20 A/m 2 W  ECCD = 0.05x10 20 A/m 2 W dolan asipp 2011 Figures are from J. Wesson, 52 Tokamaks, 2004

Bootstrap Current f b   p Neoclassical  Neoclassical  ITB core and H mode edge H-mode edge High  p case: p 1.8 MA for 2 s (half NBI, half bootstrap) half bootstrap) dolan asipp 2011 Figures are from J. Wesson, 53 Tokamaks, 2004

Current Hole Reversed shear  current hole  current hole At r/a < 0.4 L Lasts 5 s t 5 dolan asipp 2011 Figures are from J. Wesson, 54 Tokamaks, 2004

MHD Stability Diagram for JT-60U Disruptions: density limits current rise current rise error field high- ℓ i during current rampdown current rampdown kink-ballooning tearing modes If J(r) = J o (1-r 2 /a 2 )  then ℓ i = ln(1.65+0.89  ) ln(1.65 0.89  ) then ℓ i ℓ i = <B  2 >/B  a 2 dolan asipp 2011 Figures are from J. Wesson, 55 Tokamaks, 2004

Internal Inductance vs. J(r) ℓ i = <B  2 >/B  a 2 If J(r) = J o (1-r 2 /a 2 )  2 )  If J( ) J (1 2 / then ℓ i = ln(1.65+0.89  ) i Large   narrow J(r)  large ℓ i  disruption  large ℓ i  disruption. Note: J(r) may have other shapes shapes. dolan asipp 2011 Figures are from J. Wesson, 56 Tokamaks, 2004

Mitigation of Disruptions Neon pellets  0.2(divertor heat flux) Saddle coils  3/2 perturbations  suppress Saddle coils  3/2 perturbations  suppress runaway electrons Optimal vertical position   (vertical instability)  bili )  O i l i l i i  ( i l i Low p o /<p>  ELMS limit edge pressure p o p g p Ligh p o /<p>  internal  p collapses High triangularity  and ECCD  higher stable  High triangularity  and ECCD  higher stable  . dolan asipp 2011 Figures are from J. Wesson, 57 Tokamaks, 2004

Confinement Boronization  4.5 MA H-modes H-mode threshhold P th  B  n = 5-10  type I ELMs, bad 5 10  t I ELM b d n > 10  type II “grassy” ELMs, yp g y , lower heat flux to divertor dolan asipp 2011 Figures are from J. Wesson, 58 Tokamaks, 2004

Ripple Loss Fraction of P NBI P P ripple loss /P NBI /P at midplane dolan asipp 2011 Figures are from J. Wesson, 59 Tokamaks, 2004

High-  p H-mode 2 4 MA 4 3 T 2.4 MA 4.3 T reversed shear Triple product = 15x10 20 m -3 keV-s Q DT ≈ 0.4 dolan asipp 2011 Figures are from J. Wesson, 60 Tokamaks, 2004

High-  p H-mode dolan asipp 2011 Figures are from J. Wesson, 61 Tokamaks, 2004

NBI  ITB at r/a ~ 0.53 Steep gradients of p g T i and toroidal rotation speed dolan asipp 2011 Figures are from J. Wesson, 62 Tokamaks, 2004

NBI  Reversed Shear 2 6 MA 4 4 T 2.6 MA, 4.4 T 15 MW NBI  ITB at r/a ~ 0.6 <n> = 10 19  n o = 8x10 19 m -3 T io  20 keV n T  = 8 6x10 20 m -3 keV-s n o T io  E = 8.6x10 m keV-s Q DT = 1.25 Surpasses “breakeven” conditions B t di But disrupts quickly. t i kl Low-n ideal kink ballooning modes. dolan asipp 2011 Figures are from J. Wesson, 63 Tokamaks, 2004

DIII-D, General Atomics Company 1.67/0.67 m 2 2 T 2.2 T 5 MA 5 MA NBI/ICRF/ECRH = 20/4.4/3 graphite+boroniz. pellet injection pellet injection helps H-mode dolan asipp 2011 Figures are from J. Wesson, 64 Tokamaks, 2004

DIII-D Divertors single null P (H mode)  B P th (H-mode)  B double null P th independent of B dolan asipp 2011 Figures are from J. Wesson, 65 Tokamaks, 2004

ELMs in DIII-D ELMs in DIII D Type I "giant" ELMs --> losses > 10% of plasma ions n=5-10 ballooning modes Type II "grassy" ELMs higher frequency, lower amplitude 2 < 0.15 occur when s/q 95 s = shear = d(ln q)/d(ln  ) ~ dq/dr MHD activity n = 1 to 13. MHD activity n 1 to 13. dolan asipp 2011 Figures are from J. Wesson, 66 Tokamaks, 2004

H-Mode Transition dolan asipp 2011 Figures are from J. Wesson, 67 Tokamaks, 2004

Effect of E r on H-Mode H-mode is associated with a change of E r at plasma edge. E r = (Zen z ) -1 dp z /dr –v  B  + v  B  S Steepening of dp z /dr could i f d /d ld change sign of E r and trigger H-mode. gg E r change precedes other signals during H-mode transition signals during H-mode transition. dolan asipp 2011 Figures are from J. Wesson, 68 Tokamaks, 2004

Thermal Diffusivity Scalings dimensionless parameters:  * =  i /a  * =  ii /  e *   * / * / * L-mode scalings: H-mode scalings  e   * (Gyrobohm)  e   * (Gyrobohm)  i   * -1/2  i    i   *  i   (Gyrobohm) (Gyrobohm)  eff   * 0.49 dolan asipp 2011 Figures are from J. Wesson, 69 Tokamaks, 2004

Energy Confinement Scaling Ratio of H-mode to L-mode confinement time : H 89 = (   (   /   H /   ) where   where   H 89 ITER89-P ) ITER89-P is L-mode is L mode Usually H 89 ~ 2 (n/n gr ) ↑ → H 89 ↓ Neon injection → H 89 ↑ Impurities stabilize drift wave turbulence  N =  / ( I MA /aB) "Normalized beta" EAST:  =0.01, I MA =0.5, a = 0.48, B = 2.4 →  N = 2.3 dolan asipp 2011 Figures are from J. Wesson, 70 Tokamaks, 2004

Figures of Merit Fusion reactor needs high pressure  , long confinement  E "Figure of Merit" =  N H 89 "Triple Product" = n T i  E Triple Product = n o T io  E dolan asipp 2011 Figures are from J. Wesson, 71 Tokamaks, 2004

VH-Mode low impurities, recycling strong plasma shaping toroidal rotation “second stability regime” terminated after ~ 1 s Triple product ~ 5x10 20 m -3 keV-s  ~ 12.5%  5%  N =  / ( I MA /aB)  N > 2.5 dolan asipp 2011 Figures are from J. Wesson, 72 Tokamaks, 2004

Magnetic Braking Hurts VH Mode ExB rotation shear ExB rotation shear → stabilization E Error field fi ld → braking of rotation  E ↓ → E High p o /<p>  E ↑  E ↑ → → ℓ i ↑ ℓ i ↑ → → H 89 ~ 4.5 triple prod ct triple product ~ 6.2x10 20 m -3 keV-s I BS broadens J(r) → ℓ i ↓ dolan asipp 2011 Figures are from J. Wesson, 73 Tokamaks, 2004

Figure of Merit  N H 89 Best plasmas initially Best plasmas initially formed with negative central shear.  N =  / ( I MA /aB) H 89 =  E /  L-mode dolan asipp 2011 Figures are from J. Wesson, 74 Tokamaks, 2004

Maximum Stable Elongation vertical displacement instability b/a b/a injection of neon or argon stops vertical instability, reduces damage. dolan asipp 2011 Figures are from J. Wesson, 75 Tokamaks, 2004

Correction Coil Stabilizes n=1 Modes stable unstable dolan asipp 2011 Figures are from J. Wesson, 76 Tokamaks, 2004

Beta Values in DIII-D “Normalized beta”  N =  /( I /aB) Rotation helped suppress resistive wall modes (RWM) wall modes (RWM) dolan asipp 2011 Figures are from J. Wesson, 77 Tokamaks, 2004

Feedback Stabilization of RWM feedback to saddle coils dolan asipp 2011 Figures are from J. Wesson, 78 Tokamaks, 2004

TAE Modes  Fast Ion Losses  Neutron Emission Decrease Neutron Emission Decrease dolan asipp 2011 Figures are from J. Wesson, 79 Tokamaks, 2004

ICRF Fast Wave Current Drive L Mode L Mode dolan asipp 2011 Figures are from J. Wesson, 80 Tokamaks, 2004

Single and Double Null Divertors single null double null SOL flow target plates g p Inner outer dolan asipp 2011 Figures are from J. Wesson, 81 Tokamaks, 2004

Power Asymmetry to Divertor Targets More P rad on inner leg  lower heat flux to inner target. Separatrix position controls power deposition between upper and lower targets. Higher gas injection plus cryopump  lower target heat flux T ~ 2 eV recombination significant T e ~ 2 eV, recombination significant dolan asipp 2011 Figures are from J. Wesson, 82 Tokamaks, 2004

ELMy H-Mode with Gas Injection dolan asipp 2011 Figures are from J. Wesson, 83 Tokamaks, 2004

Helium Pumping Argon frosting on cryopump  He */  E ~ 8-13 dolan asipp 2011 Figures are from J. Wesson, 84 Tokamaks, 2004

ASDEX Upgrade R/b/a = 1 65/0 85/0 5 R/b/a = 1.65/0.85/0.5 3.9 T, 1.4 MA 10 s flattop NBI NBI = 20 20 ICRF = 6 ECH = 2 dolan asipp 2011 Figures are from J. Wesson, 85 Tokamaks, 2004

Divertor Design Goals reduce target heat flux reduce He accumulation in core plasma Methods ionize hydrogen neutrals in SOL increase P rad in SOL  T e < 5 eV increase neutral pressure near target increase neutral pressure near target “compression” n m (pump duct)/n i (midplane)  He */  E ~ 4-6 dolan asipp 2011 Figures are from J. Wesson, 86 Tokamaks, 2004

Radiation in ASDEX Divertor Data unavailable dolan asipp 2011 Figures are from J. Wesson, 87 Tokamaks, 2004

ASDEX Tungsten Tiles Gradually C tiles  W tiles. Keep f w << 10 -4 in core Usually f w ~ 2x10 -5 Sputtered W redeposits nearby. dolan asipp 2011 Figures are from J. Wesson, 88 Tokamaks, 2004

Operating Regimes 1 MA, 2.5 T Avoid Type I ELMs. Type III are OK. Type III are OK. Neon puff detaches divertor lowers heat flux divertor, lowers heat flux. triangularity  ballooning stable at higher p 2 cm inside separatrix dolan asipp 2011 Figures are from J. Wesson, 89 Tokamaks, 2004

Neon reduces divertor heat flux dolan asipp 2011 Figures are from J. Wesson, 90 Tokamaks, 2004

H-Mode T i (r) Profile Stifness Ion Temperatue Gradient (ITG) mode --> turbulence  i keeps same shape r/a = 0.3 to 0.8 T i / ∇ T i ~ constant limited by ITG mode limited by ITG mode e / � T T e ~ constant dolan asipp 2011 Figures are from J. Wesson, 91 Tokamaks, 2004

Fueling and Density Profile gas puffing  cooler edge  lower core parameters (density stiffness) (density stiffness) central fueling better Pellet injection from time low-field side  gas cloud. g diamagnetic plasmoid drifts back to low field side. Injection from high-field side is good. dolan asipp 2011 Figures are from J. Wesson, 92 Tokamaks, 2004

Avoidance of NTM Nearly double null Triangularity  = 0 42 Triangularity  = 0.42 Type II ELMs  only 0.5% energy l 0 5% loss.  N =  / ( I MA /aB) dolan asipp 2011 Figures are from J. Wesson, 93 Tokamaks, 2004

Bootstrap Current Maximize  p f bs + f NBI ~ 100% NBI during ramp-up  ITB ECCD counter to plasma current can sustain ITB p dolan asipp 2011 Figures are from J. Wesson, 94 Tokamaks, 2004

NTMs cause energy losses Tearing mode energy loss 4/3 <10% 3/2 10-30% 2/1 50% and disruption (q a < 3) p (q a )  N for onset   i /a ECCD can generate helical current within islands and stabilize tearing modes dolan asipp 2011 Figures are from J. Wesson, 95 Tokamaks, 2004

Reactor Issues dolan asipp 2011 Figures are from J. Wesson, 96 Tokamaks, 2004

Q vs. Triple Product Q = (fusion power) / (input power) input power = 3nTV/  E – (1/4)n 2 <  v>V(  /5) V=volume i t 3 TV/ (1/4) 2 < >V( /5) V l Underlines denote radial averages  = 2.82x10 -12 J fusion power = (1/4)n 2 <  v>V  <  v> ≈ 1 1x10 -24 T 2 <  v> 1.1x10 T m 3 /s m /s T in keV T in keV Assuming parabolic profiles: Q= 5 / [ 60nT/(n 2 <  v>  E ) – 1] ≈ 5 / [ 5x10 21 /(n o T o  E ) – 1] o  E = 5x10 21  Ignition Attained 1.5x10 21 n o T dolan asipp 2011 Figures are from J. Wesson, 97 Tokamaks, 2004

Reactor Requirements Confinement:  and  E (or  N H 89 ) To confine alphas plasma current I = 30 MA / H To confine alphas plasma current I 30 MA / H Fuel dilution N He /N = 0.012   /  E < 0.1     E < 8 Disruption prevention and mitigation Disruption prevention and mitigation Heat removal : Heat flux at target < 10 MW/m 2 Technology: magnets, structure, heating, current drive, tritium, vacuum, cryogenics, fuelling, diagnostics, tritium, vacuum, cryogenics, fuelling, diagnostics, feedback control, … dolan asipp 2011 Figures are from J. Wesson, 98 Tokamaks, 2004

International Thermonuclear E Experimental Reactor (ITER) i l R (ITER) Ignition 1988 g High-Q 2005 g P f , MW 1500 500 Burn, s 1000 400 R/a, m R/a m 8.1/2.8 8 1/2 8 6 2/2 0 6.2/2.0 I, MA 21 15 B  , T 5.7 5.3 # TF coils # TF il 20 20 1989 G$ 5.9 2.8 (~ 6 G$ 2008) dolan asipp 2011 Figures are from J. Wesson, 99 Tokamaks, 2004

ITER ITER dolan asipp 2011 Figures are from J. Wesson, 100 Tokamaks, 2004