Analysis and Simulations of the ITER Hybrid Scenario C. Kessel, R. Budny, K. Indireshkumar Princeton Plasma Physics Laboratory, USA ITPA Topical Group on Steady State Operation and Enhanced Performance Centro di Cultura Scientifica - A. Volta Societa del Casino, Como Italy, May 2005

Contents Groundrules for Hybrid Scenario –Goals –Plasma parameters and operating modes –Constraints 0D Systems analysis of Hybrid operating points –Brief description –Major parameters and effects of constraints –Large scan and operating space 1.5 D simulations of the Hybrid Scenario with TSC- TRANSP –Brief TSC-TRANSP description –TSC simulation results with GLF23 energy transport, comments –TRANSP source modeling –Benchmarking of GLF23 energy transport DIII-D

Hybrid Scenario in ITER Plasma parameter ranges –  E ≈   E 98(y,2) –  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 –Maximize f NI, T e (0) and avoid high P fusion 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)

Hybrid Scenario in ITER Constraints –Fusion power vs pulse length: 350 MW s, 500 MW s, 700 MW s –Installed auxiliary heating/CD power –Divertor heat loads: allowable of 20 MW/m 2, leading to ≈ MW/m 2 conduction heat load to account for radiation and transients –Divertor radiation??: 15% of power entering Scrape Off Layer is assumed radiated in divertor slots or about the X-point –Core radiation requirement to meet divertor heat load: 2% Be only (Z eff = 1.3) unacceptable, 2% Be + 2% C % Ar (Z eff = 2.2) is acceptable –Within volt-second capability of OH+PF: maximum of 300 V-s, with 10 V-s in breakdown, and about V-s spare, based on reference H-mode scenario –First wall surface heat flux??: 0.5 MW/m 2, with peaking factor of 2.0, leading to 0.25 MW/m 2 average

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-Neilsen 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

ITER Hybrid Systems Analysis nn nnTT TTNN q 95 IpH f Gr f BS f NI Z eff f Be fCfC f Ar t/  J %0%.2% %2%0% %1%.2% %0%.2% %0%.1% %1%0% %2%0%15.3 B T = 5.3 T, P CD = 33 MW, P aux < 39 MW, P fusion = 350 MW, N w = 0.49 MW/m 2, t flattop = 3000 s, chosen to meet N w x t flattop > 1475 MW-s/m 2  n = > n(0)/ = 1.04,  n = > n(0)/ = 1.25  T = > T(0)/ = 1.50,  T = > T(0) = 2.50

Results Fusion power pulse length limitation is most significant factor in determining Hybrid operating space –Lowering density does not continuously lead to better operating points –Higher H 98(y,2) allows access to higher fluence and lower n/n Gr –High fusion power is not necessary or desirable –Only low  N ≈ 2 operating points are required Volt-seconds capability appears to be enough to offer few thousand second flattops Divertor heat load limits is next most significant factor for Hybrid operating space –Combination of conducted power, power radiated in divertor, transient conducted power, and core radiated power First wall surface heat load limits do not appear to be limiting Available 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%

TSC TRANSP Discharge simulation with assumed source profiles and evolving boundary Plasma geometry T, n profiles q profile Interpretive rerun of discharge simulation with source models, fast ions, neutrals (TSC as expt.) Accurate source profiles fed back to TSC Analysis with interfaces to TRANSP Analysis with interfaces to TSC Flow Diagram of TSC-TRANSP 1.5D Analysis Combining Strengths of the Two Codes

TSC and TRANSP, a Few* Attributes 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 TSC –Predictive –Free-boundary/structures/PF coils/feedback control systems –T, n, j transport with model or data coefficients ( , , D, v) –LSC for LH (benchmark with other LH codes) –Assumed P and j deposition for NB, EC, and ICRF: typically use off-line analysis to derive these *In addition, both codes have models for bootstrap current, radiation, sawteeth, ripple loss, pellet fueling, impurities, etc. ** TRANSP has predictive capability

TRANSP NBCD Results for Various Conditions in the ITER Hybrid Simulations, t = 500 s I P = 12 MA, P NB = 33 MW, P ICRF = variable, ≤ 20 MW W th = 300 MJ W th = 350 MJ I NB = 2.4 MA I NB = 2.1 MA I NB = 2.2 MA I NB = 2.1 MA I NB = 1.8 MA

ICRF He3 Minority Heating Used as Heating Source to Allow NINB to Drive Current f ICRF = 52.5 MHz n He3 = 2% n DT E He3 up to 120 keV

TSC Simulation Description Density evolution prescribed, magnitude and profile Impurity is 2% Be for reference, and 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  momentum =  i ) Prescribed pedestal 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

TSC ITER Hybrid Scenario I P = 12 MA, B T = 5.3 T, V surf = 0.05V, q(0) = 0.99, q 95 = 3.95, li(1) = 0.8,,  t = 2.2%, n/n Gr = 0.79, W th = 300 MJ, n 20 (0) = 0.77, n(0)/ = 1.05,  N = 2.0, H 98(y,2) = 1.33, T e,i (0) = 22.5 keV, T e,i (0)/ = 2.0,  = 1.83,  = 0.46, ∆  rampup = 150 V-s, P  = 65 MW, P aux = 35 MW, P NINB = 33 MW, Z eff = 1.3, I NI = 5.3 MA, I BS = 3 MA, I NINB = 2.2, T ped = keV GLF23, no stab.

TSC ITER Hybrid Scenario

GLF23, no stabilization Shape control points

Variation of T ped With GLF23 no stabilization

Variation of Tped with GLF23 with ExB shear stabilization TRANSP plasma rotation assuming  mom =  i Lost W th control

Benchmark of 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

TSC Simulation Benchmark of DIII-D Discharge

No  -stabilization GLF23 turned on Profiles from TSC and TVTS and CER data at t = 5 s GLF23 w/o EXB (or  -stab) Shows Lower T(0) Values Than Those in Expt However, these do not agree with GLF23 analysis presented by Kinsey at IAEA Cases with ExB shear and  -stabilization have not been completed

Conclusions Based on GLF23 no stab. energy transport, pedestal temperatures appear high (>6.5 keV) to obtain good performance Including EXB shear stabilization in GLF23, with velocity from TRANSP assuming  mom =  I, does not improve the situation Higher density operating points can improve this Directly applying experimental Hybrid discharge characteristics to ITER may be optimistic –Lower rotation in ITER by 10x –T i ≈ T e in ITER –Density peaking in expts. from NB fueling and ExB shear or other particle transport effects, may not exist in ITER Application of GLF23 to full discharge simulations is continuing –No stabilization application is robust in TSC –With stabilization has start up difficulties, are being resolved –Will apply to L-mode and H-mode phases –Benchmark simulations with hybrid discharges is continuing

(Possible) Future Work Determine engineering constraints for use in 0D systems analysis in greater detail –Is the cryoplant limitation for real?? Complete higher density and higher Z eff Hybrid 1.5D scenarios Examine slight density peaking in 1.5D scenarios Turn off stored energy control, examine Q vs. T ped vs. v ExB Examine higher velocities in ExB shear stabilized cases Examine strategies for coupled stored energy and NICD feedback control Use alternative energy transport models, less stiff models to see the dependence on required pedestal temperatures Expand benchmark simulations to other Hybrid discharges in DIII-D Further development of TSC-TRANSP modeling Etc.