1. Progress on NSTX towards steady state at low aspect ratio D. A. Gates, Princeton Plasma Physics Laboratory on behalf of the NSTX Research Team Supported by * Work supported by US DOE Contract No. DE-AC02-76CH03073 ST Requires high bootstrap fraction, f bs, simultaneous with high  t High f bs and high  t competing requirements (at fixed shape and  N ) Progress for ST and advanced tokamak given by  sus  f bs  t ~ S  N 2 [S = q 95 (I p /(aB t ))] –If  Nmax = C Troyon, then only shape improves  sus –  sus increases linearly with increasing S [S = q 95 (I p /(aB t ))] –Component Test Facility requires  t ~ 20% and f bs ~ 50%,  sus = 10% –ARIES-ST requires  t ~ 40% and f bs ~ 90% Optimized shaping with new PF coils for high triangularity and elongation NSTX has achieved record values of elongation and shape factor –Leads directly to record values of the  sus for the ST For NSTX 100% non-inductive operation with  N ~ 7 only with strong shaping New inboard divertor coils increase accessibility of high-triangularity, high-elongation shapes Highest  now obtained at highest  ≈0.8  S  q 95 I P /aB T = 41 MA/m·T Small (Type V) ELM regime recovered at high  > 2.5 with new coils –Previously observed onset of large ELM-like events when  > 2.2 D. Gates, J. Menard Highest elongation  =3.0 (transient)Sustained elongation  =2.7 (0.1s) Record pulse-lengths achieved at high current by operating with sustained H-mode H-mode with small ELMs  lower flux consumption, slow density rise  T = 20%  N = 5.2%mT/MA  E = 52ms H 89L = 2.0 A = 1.4,  = 2.3,  L = 0.75, l i = 0.49 Long duration discharges reach ~70% non-inductive current TRANSP model agrees with measured neutron rate during high-  phase –Model includes anomalous fast ion diffusion during later phase when low-m MHD activity is present 85% of non-inductive current is  p-driven –Bootstrap + Diamagnetic + Pfirsch-Schlüter 1/1 mode onset causes  drop, fast ion diffusion (Menard, PRL) Expected performance improvements observed as shaping has increased Reduction in control system latency increases elongation (2004) Plasma shaping enhancements Upgrade of PF1A coil enables simultaneous achievement of high  and  Overlay of EFIT boundaries from shots 117707 and 117814 between 0.3 and 0.9s ~3cm includes MHD perturbations Control points Boundary overlay time window Implementation of rtEFIT improves shape control reproducibility Modified PF1A Coil Old PF1A Coil Shot 121241 (record  ) All shots 2001-2003 All shots 2004-2006 2001 2005 2004 2002-3  = 2.75,  = 0.8, S ~ 37  = 1.8,  = 0.6, S ~ 22  = 2.0,  = 0.8, S ~ 23  = 2.3,  = 0.6, S ~ 27  = 3,  = 0.8, S ~ 41 2006 Pulse averaged approximate  sus versus S (S is averaged over the same time window as  sus )  N H89 vs.  pulse /  E Pulse averaged  t versus pulse length Data are sorted by year and by S 2001-2004 2005-2006 External kink mode ultimate limit on  t - long pulse discharges above no-wall limit Troyon diagram showing  tmax vs. I p /aB t PEST eigenfunction Shot 117707 Time history of  t and n=1 (no-wall) kink mode growth rate, , for Shot 117707, calculated by PEST using LRDFIT equilibrium including MSE  Alfven  % Each point in the above plots represents an EFIT equilibrium reconstruction Data span entire NSTX database and are filtered against rapid plasma motion Effect of modifications provide clear increase in NSTX operating space

2. 117407 LSN 117432 DN 117424 high-  DN 117407 LSN 117432 DN 117424 high-  DN Increased triangularity actually reduces peak heat flux to divertor target Flux expansion decreases peak heat flux despite reduced major radius Compare single-null & double-null configurations with triangularity  ≈ 0.4 at X-point and high triangularity  = 0.8 double-null plasmas –Measure heat flux with IR thermography of carbon divertor tiles Peak heat flux decreases as 1 : 0.5 : 0.2 ELM character changes: Type I  Mixed  Type V R. Maingi Changes in X-point balance affect ELM characteristics Very small changes in the plasma boundary reproducibly lead to large differences in ELM behavior ELMs have a major impact on plasma performance, controlling them is crucial Precise plasma control provides an important tool for controlling ELMs - highly ITER relevant  r sep (mm) for 117424 (black) and 117425 (red) Shot 117424 Shot 117425 Shot 117424 Shot 117425  r sep is the radial separation of the flux surfaces which pass through the x-points measured at the outboard midplane 1.0 0.0 Encouraging results from both EBW emission and HHFW current drive experiments Long pulse discharges have elevated q(0) without low frequency MHD modes Experiment (116313) Target The time history of shot 120001 showing: 1) the components of non-inductive current as indicated in the plot legend, 2) The plasma current (black) and the surface voltage (red), 3) the reconstructed q(0) (black) and q min (red) and q(0) as determined by a TRANSP magnetic diffusion calculation (blue) 4)  t (black) and the neutral beam power (red). The profiles in preceding figure are calculated over the time interval indicated in green. A spectrogram of magnetic fluctuations as measured by a Mirnov coil for shot 120001. The colors represent toroidal mode numbers as indicated in the legend. Notice the period of time after 0.6s where there are only small amplitude high-n MHD modes present. Measured EBW uses efficient Ohkhawa current drive Data on efficiency of EBW emission from identical plasmas for which the EBW antenna pointing angle is varied. The colors represent measured efficiencies of - 81-90% (red) - 71-80% (orange), - 61-70% (green) - 51-60% (blue) - 41-50% (purple), - 31-40% (black ). The ellipses are contours of theoretically predicted emission efficiency Measured EBW emission angle matches theoretical predictions Achieved high T e = 3.6keV in current drive phasing for first time using high B T = 5.5kG –Improvement consistent with reduced PDI and surface waves expected at higher B T –Expect similar improvements from higher k || Useful for HHFW-CD during ramp-up Useful for HHFW heating at high-  HHFW heats efficiently in current drive phasing Steady state scenario predicted with 100% non-inductive current using only NBICD and pressure driven currents Plasma shape and profiles for predicted 100% non-inductive scenario I P = 750kA  N < 5.6,  P < 1.5,  T < 17% l i = 0.6, q min =1.3, B T =4.5kG  = 2.3,  X-L = 0.75, q*=3.9 Present high-f NI long-pulse H-modes: Target scenario: Inductive current drive is replaced by: Higher J NBI from higher T e Higher J BS from higher  P-thermal Need 60% increase in T, 25% decrease in n e –Lithium for higher  E & density control? 20% increase in thermal confinement 30% increase in HH 98 –Core HHFW heating Want q 0  q min  2.4  higher with- wall limit Higher  for higher q,  P, f BS High  for improved kink stability (shape parameters already achieved in other discharges) I P = 700kA  N = 6.7,  P = 2.7,  T =15% l i = 0.5, q min = 2.4, B T =5.2kG  = 2.6,  X-L = 0.85, q*=5.6 Neoclassical Analysis of current profile information Full complement of kinetic profile data enables analysis of current profile composition Full neoclassical calculation of ohmic, and pressure driven currents Shot 120001 Black total predicted current Gray total reconstructed current (MSE) Orange - Ohmic current Red - Bootsrap Blue - Beam driven current (TRANSP) Data averaged over 0.7 - 0.8s Loop voltage profile calculated from equilibria constrained with MSE data Neoclassical resistivity and bootstrap current from Sauter, et al., Phys. Plasmas 6 (1999) 2834 1) 2) 3) 4) 2