FIRE Advanced Tokamak Progress C. Kessel Princeton Plasma Physics Laboratory NSO PAC 2/27-28/2003, General Atomics 1.0D Operating Space 2.PF Coils 3.Equilibrium/Stability 4.ECCD/Neoclassical Tearing Modes 5.RWM Stabilization 6.ICRF, FWCD, LHCD 7.TSC-LSC Simulations 8. Further Work ---> PVR

ARIES-AT Provides a Long Term Target for Advanced Tokamaks FIRE-AT will need to show how close we can get to this configuration thru control in a burning plasma Steady state Strong plasma shaping Large bootstrap fraction, minimal CD High  Transport that supports high f BS and high  Plasma edge solution that supports CD, power handling, divertor solution

0D Operating Space Analysis for FIRE AT Heating/CD Powers –ICRF/FW, 30 MW –LHCD, 30 MW Using CD efficiencies –  (FW)=0.20 A/W-m2 –  (LH)=0.16 A/W-m2 P(FW) and P(LH) determined at r/a=0 and r/a=0.75 I(FW)=0.2 MA I(LH)=Ip(1-fbs) Scanning Bt, q95, n(0)/, T(0)/, n/nGr,  N, fBe, fAr Q=5 Constraints: –  (flattop)/  (CR) determined by VV nuclear heat (4875 MW-s) or TF coil (20s at 10T, 50s at 6.5T) –P(LH) and P(FW) ≤ max installed powers –P(LH)+P(FW) ≤ Paux –Q(first wall) < 1.0 MW/m2 with peaking of 2.0 –P(SOL)-Pdiv(rad) < 28 MW –Qdiv(rad) < 8 MW/m2

FIRE’s Q=5 AT Operating Space Access to higher t flat /  j decreases at higher  N, higher Bt, and higher Q, since t flat is set by VV nuclear heating Access to higher radiated power fractions in the divertor enlarges operating space significantly

FIRE’s AT Operating Space Q = 5-10 accessible  N = accessible f bs = accessible t flat /t j = 1-5 accessible If we can access….. H98(y,2) = Pdiv(rad) = P(SOL) Zeff = n/n Gr = n(0)/ =

Examples of Q=5 AT Points That Obtain  flat /  J > 3 nn nnTT TT BTBT q 95 IpH f Gr f BS P cd PP z eff f Be f Ar t/  %.3% %.2% %.2% %.2% %.3% %.3% %.2% %.2% %.3%3.29 HH < 1.75, satisfy all power constraints, Pdiv(rad) < 0.5 P(SOL)

PF Coils Must Sustain AT Plasmas with Low li and High  Ip= MA, Bt= T 100% non-inductive in flattop Ip can not be too low or we’ll loose too many alphas from ripple Inductive + non-inductive rampup ----> consumes V-s, what is final flux state?? Flattop times = 16-50s (from P fusion of MW) and TF coil Low li(3) = 0.42,  N=4.2 Divertor coils are driven to high currents

PF Coil Capability for AT Modes Advanced tokamak plasmas –Range of current profiles: 0.35 < li(3) < 0.55 –Range of pressures: 2.50 <  N < 5.0 –Range of flattop flux states: chosen to minimize heating and depends on flattop time (determined by P fusion ) –Ip limited to ≤ 5.5 MA Lower li operating space led to redesign of divertor coils –PF1 and PF2 changed to 3 coils and total cross-section enlarged Presently examining magnet stresses and heating for AT scenarios

Neo-Classical Tearing Modes at Lower Bt for FIRE AT Modes Bt=6.5 T Bt=7.5 T Bt=8.5 T Ro Ro+a fce=182fce=142 fce=210fce=164 fce=190fce= GHz 200 GHz Target Bt=6.5-7 T for NTM control, to utilize 170 GHz from ITER R&D Must remain on LFS for resonance ECCD efficiency, can local  e be high enough to avoid trapping boundary?? Can we rely on OKCD to suppress NTM’s far off-axis on LFS versus ECCD ?? (enhanced Ohkawa affect at plasma edge) Avoid NTM’s with j profile and q>2.0 or do we need to suppress them??

J. Decker, APS 2002,MIT OKCD allows LFS EC deposition, with similar A/W as ECCD on HFS

Comments on ECCD in FIRE ASDEX-U shows that 3/2 island is suppressed for about 1 MW of power with I ECCD /I p = 1.6%, giving A/W –Ip=0.8 MA and  N =2.5 DIII-D shows that 3/2 island is suppressed for about MW with j EC /j BS = –Ip= MA,  N = OKCD analysis of Alcator-CMOD gives about A/W FIRE’s current requirement should be about 15 times higher than ASDEX-U (scaled by Ip and  N 2 ) –Need about 200 kA, which would require about 35 MW?? Early detection reduces power alot according to ITER –Do we need less current for 5/2 or 3/1, do we need to suppress them?? Is 170 GHz really the cliff in EC technology?? MIT, short pulse results

Updating AT Equilibrium Targets Based on TSC-LSC Equilibrium TSC-LSC equilibrium Ip=4.5 MA Bt=6.5 T q(0)=3.5, q min =2.8  N =4.2,  =4.9%,  p=2.3 li(1)=0.55, li(3)=0.42 p(0)/  p  =2.45 n(0)/  n  =1.4 Stable n=  Stable n=1,2,3 with no wall √V/Vo

Original AT Target Equilibrium for FIRE   = 3.65, f bs < 0.75  N = 2.5, f bs < 0.55 q(min) = r/a(qmin) = 0.8 n(0)/ = 1.5 Ip = 5.5 MA Bt = 8.5 T No wall stabilization  N = 2.5 n=1 RWM stabilized  N = 3.65 This needs to be revisited with Ip=4.5 MA and Bt=6.5 T

Stabilization of n=1 RWM is a High Priority on FIRE Feedback stabilization analysis with VALEN shows strong improvement in , taking advantage of DIII-D experience, most recent analysis indicates  N(n=1) can reach 4.2 What is impact of n=2??

Ideal wall n=1 limit

FIRE Uses ICRF Ion Heating for Its Reference and AT Discharges ICRF ion heating – MHz –2 strap antennas –4 ports (2 additional reserved) –20 MW installed (10 MW additional reserved) –He3 minority and 2T heating –Frequency range allows heating at a/2 on HFS and LFS (C- Mod ITB) Full wave analysis –SPRUCE in TRANSP –Using n(He3)/ne = 2% –n 20 (0) = 5.3, = 4.4 –P ICRF = 11.5 MW,  = 100 MHz –T He3 (0) = 10.2 keV –P abs (He3) = 60% –P abs (T) = 10% –P abs (D) = 2% –P abs (elec) = 26% Antenna design --->D. Swain, ORNL

ICRF/FW Viable for FIRE On-Axis CD PICES (ORNL) and CURRAY(UCSD) analysis f = MHz n|| = 2.0 n(0) = 5x10^20 /m3 T(0) = 14 keV 40% power in good part of spectrum (2 strap) ----> A/W CD efficiency with 4 strap antennas is 50% higher Operating at lower frequency to avoid ion resonances, v ph /v th ?? Calculations assume same ICRF ion heating system frequency range, approximately 40% of power absorbed on ions, can provide required AT on-axis current of MA with 20 MW (2 strap antennas) E. Jaeger, ORNL

Benchmarks for LHCD Between LSC and ACCOME (Bonoli) Trapped electron effects reduce CD efficiency Reverse power/current reduces forward CD Recent modeling with CQL and ACCOME/LH19 will improve CD efficiency, but right now…….. Bt=8.5T ----> 0.25 A/W-m2 Bt=6.5T ----> 0.16 A/W-m2 FIRE has increased the LH power from 20 to 30 MW

HFS Pellet Launch and Density Peaking ---> Needs Strong Pumping FIRE reference discharge with uniform pellet deposition, achieves n(0)/ ≈ 1.25 Simulation by W. Houlberg, ORNL, WHIST P. T. Lang, J. Nuc. Mater., 2001, on ASDEX and JET L. R. Baylor, Phys. Plasmas, 2000, on DIII-D

HFS Launch V=125 m/s, set by ORNL pellet tube geometry Vertical and LFS launch access higher velocities

TF Ripple and Alpha Particle Losses TF ripple very low in FIRE  (max) = 0.3% (outboard midplane) Alpha particle collisionless + collisional losses = 0.3% for reference ELMy H- mode For AT plasmas alpha losses range from 2-8% depending on Ip and Bt ----> are Fe inserts required for AT operation??? Optimize for Bt=6.5T

Fe Shims for Ripple Reduction in FIRE TF Coil Outer VV Inner VV Fe Shims

TSC-LSC Simulation of Burning AT Plasma in FIRE Bt=6.5 T, Ip=4.5 MA q(0) =4.0, q(min) = 2.75, q(95) = 4.0, li = 0.42  = 4.7 %,  N = 4.1,  p = 2.35 n/nGr = 0.85, n(0)/ = 1.47 n(0) = 4.4x10^20, n(line) = 3.5, n(vol) = 3.0 Wth = 34.5 MJ  E = 0.7 s, H98(y,2) = 1.7 Ti(0) = 14 keV, Te(0) = 16 keV  (total) = 19 V-s, P  = 30 MW P(LH) = 25 MW P(ICRF/FW) = 7 MW –Up to 20 MW ICRF used in rampup P(rad) = 15 MW Zeff = 2.3 Q = 5 I(bs) = 3.5 MA, I(LH) = 0.80 MA, I(FW) = 0.20 MA t(flattop)/  j=3.2

TSC-LSC Simulation of Q=5 Burning AT Plasma Ip=4.5 MA, Bt=6.5 T,  N =4.1, t(flat)/  j=3, I(LH)=0.80, P(LH)=25 MW f BS =0.77, Z eff =2.3,

TSC-LSC Simulation of Q=5 AT Burning Plasma

AT Physics Capability on FIRE Strong plasma shaping and control Pellet injection, divertor pumping, impurity injection FWCD (electron heating) on-axis, ICRF ion heating off-axis LHCD (electron heating) off-axis ECCD (LFS, electron heating) off-axis, MHD control RWM MHD feedback control NBI ?? (need to examine for AT parameters!!) t(flattop)/t(curr diff) = 1-5 Diagnostics Control MHD J Profile P-profile Rotation

Ongoing Advanced Tokamak Work ---> PVR Establish PF Coil operating limits Revisit Equilibrium/Stability Analysis Use recent GLF23 update in AT scenarios LHCD efficiency updates EC with FIRE’s parameters Orbit calculations of lost alphas for scenario plasmas RWM coil design in port plugs and RF ports Determine possible impact of n=2 RWM on access to high  N Examine NBI for FIRE AT parameters