Simulation and Analysis of the Hybrid Operating Mode in ITER C. Kessel, R. Budny, and K. Indireshkumar Princeton Plasma Physics Laboratory Symposium On Fusion Engineering Knoxville, TN September 26-29, 2005

So What is a Hybrid Operating Mode in ITER? Reference H-mode Ip = 15 MA B T = 5.3 T R = 6.2 m a = 2.0 m V loop = 0.09 q 95 = 3  N = 1.8 H 98(y,2) = 1.0 q(0) ≤ 1.0(r saw ≈ 1 m) Q = 10 T flattop = 500 s Hybrid Mode Ip = 12 MA B T = 5.3 T R = 6.2 m a = 2.0 m V loop = q 95 = 4  N = 3.0 H 98(y,2) = 1.5 q(0) ≥ 1.0(r saw small) Q = 5-10 T flattop = 3000 s Steady State (AT) Mode Ip = 9 MA B T = 5.3 T R = 6.35 m a = 1.85 m V loop = 0.0 q 95 = 4  N = H 98(y,2) = 1.6 q(0) > Q = 5 T flattop = ∞

Hybrid Scenario in ITER Plasma parameter ranges –  E ≈   E 98(y,2) –  N NTM <  N <  N no wall (≈ 3) –f NI ≈ 50% –I P ≈ 12 MA –n/n Gr varied –  CD determined from TRANSP, or other analysis –Impurities defined to provide acceptable divertor heat loading Operating Modes –NNBI + ICRF –NNBI + ICRF + LH –NNBI + ICRF + EC Prefer to avoid (or minimize) the sawtooth, q(0) ≥ 1.0 –Maximize f NI off-axis (I BS, I LH, I ECCD ) Maximize neutron fluence –N wall  t flattop –t flattop is minimum of t V-s or t nuc-heat Remain within installed power limitations –NNBI at 1.0 MeV, 33 MW –ICRF at about 52 MHz, 20 MW –EC at 170 GHz, 20 MW –LH at 5 GHz, 30+ MW (UPGRADE)

Integrated Modeling of ITER Hybrid Burning Plasma Scenarios 0D systems analysis to identify operating space within engineering contraints 1.5D discharge simulations –Energy transport (GLF23) –Heating/CD –Free-boundary equilibrium evolution/feedback control –Other control; stored energy, f NI, etc. Energy transport experimental verification Ideal MHD analysis Offline heating/CD source analysis Offline gyrokinetic transport simulations (Budny) Fast particle effects and MHD (Gorelenkov) Particle transport/impurity transport Integrated SOL/divertor modeling Non-ideal MHD, NTM’s

0D Systems Analysis Identifies Device Constraints for Scenario Simulations ITER’s Primary Device Limitations That Affect Scenarios –Fusion power vs pulse length ----> heat rejection system 350 MW for 3000 s 500 MW for 400 s 700 MW for 150 s ----> (maximum P fus cryoplant limits) –Divertor conducted heat load, maximum > 20 MW/m2, nominal MW/m2 ----> allowable divertor heat load Radiation from plasma core and edge, P SOL = (1 - f core rad ) P input Radiation in divertor and around Xpt, P cond = (1 - f div rad ) P SOL Radiation distribution in divertor channel, impurities, transients –Volt-second capability ----> PF coil current limits Approximately V-s –First wall surface heat load limit (not limiting for normal operation) –Duty cycle, t flattop /(t flattop + t dwell ) ----> cryoplant for SC coils Limited to about 25% What device upgrades are required for advanced operating modes, and are they major or minor upgrades?

0D Operating Space Analysis Energy balance Particle balance,  P */  E and quasi- neutrality Bosch-Hale fusion reactivity Post-Jensen coronal equilibrium Albajar cyclotron radiation model Hirshman-Neilson flux requirement (benchmarked with TSC) T(r) = (T o - T a )[1-(r/a) 2 ]  T + T a Same for density profile Etc. I P = 12 MA B T = 5.3 T R = 6.2 m A = 3.1  95 = 1.75  95 = 0.5  P */  E = 5 ∆  total = 300 V-s ∆  breakdown = 10 V-s li = 0.80 C E = 0.45  NBCD = 0.3 x A/W-m 2 P CD = 33 MW  T = 1.75, T a /T o = 0.1  n = 0.075, n a /n o = 0.3 f Be = 2.0% 1.5 ≤  N ≤ ≤ n/n Gr ≤ ≤ Q ≤ ≤ f C ≤ 2.0% 0.0 ≤ f Ar ≤ 0.2% Input parameters Scanned parameters

ITER Hybrid Systems Analysis Fusion power pulse length limitation significantly reduces accessible fluence values, and changes dependence on density

ITER Hybrid Systems Analysis Operating space shows strong dependence on allowable conducted peak heat flux on divertor, which must be low enough to accommodate radiation flux and transients

ITER Hybrid Systems Analysis Increasing the power radiated in the divertor can recover operating space at lower conducted peak heat flux

ITER Hybrid Systems Analysis Large Operating Space Scan 1.05 ≤ n(0)/  n  ≤ ≤ T(0)/  T  ≤ ≤ I P (MA) ≤ ≤  N ≤ ≤ n/n Gr ≤ ≤ Q ≤ % ≤ f Be ≤ 3% 0% ≤ f C ≤ 2% 0% ≤ f Ar ≤ 0.2% Other input fixed at previous values

Results Fusion power pulse length limitation is a significant factor in determining Hybrid operating space –Hybrid operating modes on present tokamaks operate in  N window, close to  N ≈ 3 –Existing pulse length vs fusion power limits indicate optimum  N to maximize neutron fluence is about 2.0 (P fusion ≈ 325 MW) –For ITER to operate close to  N ≈ 3, P fusion ≈ 500 MW, the pulse length would be severely limited by heat rejection system –Hybrid operating modes in ITER require upgrades to heat rejection system Volt-seconds capability of PF coils appears to be enough to offer few thousand second flattops –Depending on precise value of V loop First wall surface heat load limits do not appear to be limiting during normal operation due to large FW surface area

Divertor heat load limits is second most significant factor for Hybrid operating space –Core/edge radiated power (bremsstrahlung, cyclotron, line) –Conducted power –Power radiated in divertor region –Transient conducted power Operating space shows that existing ITER design can provide reasonable fluence levels within a discharge – HOWEVER time between discharges is constrained –Appears that cryoplant limitation sets t flat /(t flat +t dwell ) ≈ 25% –For Hybrid operating modes in ITER to provide significant fluence the cryoplant must be upgraded Systems Analysis Results

Pursuing 1.5D Integrated Modeling of ITER with TSC/TRANSP Combination TRANSP Interpretive Fixed boundary Eq. Solvers Monte Carlo NB and  heating SPRUCE/TORIC/CURRAY for ICRF TORAY for EC LSC for LH Fluxes and transport from local conservation; particles, energy, momentum Fast ions Neutrals Plasma geometry T, n profiles q profile Accurate source profiles fed back to TSC TSC Predictive Free- boundary/structures /PF coils/feedback control systems T, n, j transport with model or data coefficients ( , , D, v) LSC for LH Assumed source deposition for NB, EC, and ICRF: typically use off-line analysis to derive these both codes have models for bootstrap current, radiation, sawteeth, ripple loss, pellet fueling, impurities, etc. TSC evolution treated like an experiment

1.5D ITER Hybrid Simulations Integrate Transport, Heating/CD, and Equilibrium Density evolution prescribed, magnitude and profile 2% Be + 2% C % Ar for high Z eff cases GLF23 thermal diffusivities, no rotation stabilization, and with rotation stabilization (plasma rotation from TRANSP assuming   =  i ) Prescribed pedestal height and location amended to GLF23 thermal diffusivities Control plasma current, radial position, vertical position and shape Plasma grown from limited starting point on outboard limiter, early heating required to keep q(0) > 1, keep P heat < 10 MW Control on plasma stored energy, P ICRF in controller, P NB not in controller since it is supplying NICD

Using TRANSP Monte Carlo NB and SPRUCE Full Wave/FP ICRF Analysis to Model ITER Hybrid Sources I P = 12 MA, P NB = 33 MW, P ICRF = 20 MW W th = 300 MJ W th = 350 MJ I NB = 2.1 MA I NB = 1.8 MA ICRF Heating NINB Heating/CD

Source Modeling in TRANSP ITER’s NBs are large single source beams Plasma rotation produced by NBs is much lower than present devices Minority heating with ICRF shows very centralized absorption slightly off axis Each NB source is 16 MW, although modulation could provide finer power injection

ITER Hybrid at  N ≈ 3 Produces 475 MW of Fusion Power I P = 12 MA B T = 5.3 T I NI = 7.8 MA  N = 2.96 n/n Gr = 0.93 n 20 (0) = 0.93 W th = 450 MJ H 98 = 1.6 T ped = 9.5 keV ∆  rampup = 150 V-s V loop = V Q = 9.43 P  = 100 MW P aux = 53 MW P rad = 28 MW Z eff = 2.25 q(0) < 1, ≈ 0.93 r(q=1) = 0.45 m li(1) = 0.78 Te,i(0) = 30 keV Available t flattop V-s > 4000 s

Shape control points

High T ped Required to Get  N ≈ 3 with GLF23 Core Transport ITER expected to have Low v rot (≈ 1/10 v rot DIII-D ) T i ≈ T e Low n(0)/ Present Expts have High v rot T i > T e n(0)/ > 1.25 Direct extrapolation from present Expts to ITER may be optimistic

Density Peaking Makes Energy Transport Worse with GLF23 Core Transport GLF23 predicts higher thermal diffusivities for more peaked density case Flat n(  ) Peaked n(  )

Efforts to Benchmark GLF23 Transport in DIII-D Hybrid Discharge TSC free-boundary, discharge simulation DIII-D data PF coil currents Te,i(  ), n(  ), v(  ) NB data TRANSP Use n(  ) directly TSC derives  e,  I to reproduce T e and T i Turn on GLF23 in place of expt thermal diffusivities Test GLF23 w/o ExB and w EXB shear stabilization t = 1.5 st = 5.0 s L-mode, i-ITB H-mode

Profiles from TSC and TVTS and CER data at t = 5 s TSC Simulation Benchmark of DIII-D Discharge

Using JSOLVER/BALMSC/PEST2, … to Analyze Ideal MHD Stability of ITER Hybrid Hybrid discharges operate in a  N window  N NTM <  N <  N n=1(no wall) Hybrid discharges have f NI ≥ 40%, from NBCD on-axis and BS off-axis Hybrid discharges prefer q(0) > 1 or small sawtooth amplitude or possibly small r(q=1) Examine Porcelli sawtooth model in 1.5D simulations to determine the sawtooth response to small r(q=1), and local dq/dr and dp/dr  N ’s up to 3.2 are stable to n=1 w/o conducting wall

ITER Hybrid Scenario Requires High T ped, High n/n Gr, and High  E Systems Analysis shows that upgrades to the heat rejection and cryoplant systems will be necessary to achieve long pulses and significant neutron fluence in the Hybrid operating mode 1.5 Discharge Evolution calculations, with GLF23 core transport model, indicate that the Hybrid will require High T ped ≈ 10 keV (making power to divertor too high) High n/n Gr ≈ 0.95 High H 98(y,2) ≈ 1.6 to reach its operating space of  N ≈ 3 Including plasma rotation, determined by TRANSP, does not improve the energy confinement significantly

Present Hybrid experiments have characteristics that give them high performance Strong plasma rotation T i > T e Some level of density peaking However, these features will be missing in ITER, so we must project with caution to the ITER Hybrid Verification of the GLF23 core transport model shows reasonable agreement with experimental Hybrid discharges, work is continuing Ideal MHD calculation indicate the ITER Hybrid discharge simulation cases are stable to n=1 external kink modes without a wall, but they may be unstable to sawteeth, work is continuing