Integrated Modeling and Simulations of ITER Burning Plasma Scenarios C. E. Kessel, R. V. Budny, K. Indireshkumar, D. Meade Princeton Plasma Physics Laboratory IEA Large Tokamak Workshop (W60) Burning Plasma Physics and Simulation Tarragona, Spain July 4-5, 2005

Integrated Scenario Development for ITER Advanced Operating Modes Goals –Provide discharge simulations for self-consistency of plasma configurations –Identify impact of uncertain physics –Identify operating space within device/engineering limitations, and required upgrades –Define auxiliary systems requirements –Examine plasma control issues (equilibrium, stored energy, current profile) –Examine physics/engineering interfaces (PF coils, divertor) –Examine detailed physics modeling outside scope of 1.5D simulations, feeding back constraints for 1.5D –Provide experiment/theory comparison/verification within an integrated physics simulation

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?

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

Shape control points I P = 12 MA B T = 5.3 T I NI = 7.75 MA  N = 2.90 n/n Gr = 0.93 W th = 450 MJ H 98 = 1.56 T ped = 9 keV ∆  rampup = 150 V-s V loop = V Q = 9.43 P  = 100 MW P aux = 53 MW P rad = 27 MW Z eff = 2.25 q(0) < 1, ≈ 0.9 r(q=1) = 0.45 m li(1) = 0.8 t flattop V-s ≈ 3000 s ITER Hybrid with GLF23 Requires High n/n Gr, High T ped to Reach  N ≈ 3 t = 170 st = 1500 s 

ITER Hybrid Simulation Shows Rapid q(0) Drop, V-s are Low, Long Core Relaxation

ITER Hybrid Scenario Needs High T ped for GLF23 w/o and w ExB 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 Continuing analysis with   = f x  i, higher n(0)/, etc. v rot from TRANSP with   =  i

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

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

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 s t = 5.0 s L-mode, i-ITB H-mode

Energy Transport is at Center of Modeling/Projections for ITER Whats wrong : –How is comparing to experiment improving our modeling?? Error are say 20-30% on T e and T i profiles, and maybe 10% on stored energy –Multiple models, for example GLF23 and MMM95, will give reasonable agreement on any given experiment, how good should this be to believe a projection to ITER (remember ITER- EDA, comparison of transport models in Physics Basis 1999, what has changed??) What can we do : –Examine the critical features of transport; external rotation, T e /T i, safety factor, density peaking, etc. and test these –Apply transport models to difficult expts. C-mod ITB, JT60-U high  P, DIII-D/AUG/JET Hybrid and AT discharges…. –Apply transport model to entire discharge, not a single flattop time-slice –Consider what will be present in burning plasma device

Integrated Modeling of Burning Plasmas Integration includes feeding back numerous offline analyses to constrain the core 1.5D transport modeling –0D analysis for operating space limitations –Ideal and non-ideal MHD analysis –Source modeling benchmarks –Detailed SOL/divertor modeling –Particle and impurity transport (fueling) TSC/TRANSP is being used to improve the 1.5D simulations of burning plasma scenarios on ITER --- provides integration of full discharge free- boundary/feedback control and sophisticated source modeling/fast particle treatment Energy transport is at the center of 1.5D transport simulations –Project to ITER with consideration of difference from present expts. –Apply theoretical models to difficult experimental cases and for the entire discharge –Pedestal projections need to transition from empirical to theory based In some areas our integrated modeling needs more effort –Particle/impurity transport –SOL/divertor integrated into core evolution –Non-ideal MHD