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## Integrated Modeling and Simulations of ITER Burning Plasma Scenarios

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**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, fNI, 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 Pfus cryoplant limits) • Divertor conducted heat load, maximum > 20 MW/m2, nominal 5-10 MW/m2 ----> allowable divertor heat load • Radiation from plasma core and edge, PSOL = (1 - fcorerad) Pinput • Radiation in divertor and around Xpt, Pcond = (1 - fdivrad) PSOL • Radiation distribution in divertor channel, impurities, transients • Volt-second capability ----> PF coil current limits • Approximately 260-280 V-s • First wall surface heat load limit (not limiting for normal operation) • Duty cycle, tflattop/(tflattop + tdwell) ----> 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 Plasma geometry T, n profiles q profile • 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 • 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 evolution treated like an experiment Accurate source profiles fed back to TSC both codes have models for bootstrap current, radiation, sawteeth, ripple loss, pellet fueling, impurities, etc.**1.5D ITER Hybrid Simulations Integrate Transport,**Heating/CD, and Equilibrium • Density evolution prescribed, magnitude and profile • 2% Be + 2% C + 0.12% Ar for high Zeff 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 Pheat < 10 MW • Control on plasma stored energy, PICRF in controller, PNB not in controller since it is supplying NICD**ITER Hybrid with GLF23 Requires High n/nGr, High Tped to**Reach N ≈ 3 t = 170 s t = 1500 s IP = 12 MA BT = 5.3 T INI = 7.75 MA N = 2.90 n/nGr = 0.93 Wth = 450 MJ H98 = 1.56 Tped = 9 keV ∆rampup = 150 V-s Vloop = 0.042 V Q = 9.43 P = 100 MW Paux = 53 MW Prad = 27 MW Zeff = 2.25 q(0) < 1, ≈ 0.9 r(q=1) = 0.45 m li(1) = 0.8 tflattopV-s ≈ 3000 s Shape control points **ITER Hybrid Simulation Shows Rapid q(0) Drop, V-s are Low,**Long Core Relaxation**ITER Hybrid Scenario Needs High Tped for GLF23 w/o and w ExB**• ITER expected to have • Low vrot (≈ 1/10 vrotDIII-D) • Ti ≈ Te • Low n(0)/<n> • Present Expts have • High vrot • Ti > Te • n(0)/<n> > 1.25 • Direct extrapolation from present Expts to ITER may be optimistic • Continuing analysis with = f x i , higher n(0)/<n>, etc. vrot from TRANSP with = i**Using TRANSP Monte Carlo NB and SPRUCE Full Wave/FPICRF**Analysis to Model ITER Hybrid Sources IP = 12 MA, PNB = 33 MW, PICRF = 20 MW NINB Heating/CD ICRF Heating Wth = 300 MJ Wth = 350 MJ INB = 2.1 MA INB = 1.8 MA**Using JSOLVER/BALMSC/PEST2, … to Analyze Ideal MHD**Stability of ITER Hybrid Hybrid discharges operate in a N window NNTM < N < Nn=1(no wall) Hybrid discharges have fNI ≥ 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 104276 Hybrid**Discharge t = 1.5 s t = 5.0 s • TSC free-boundary, discharge simulation • DIII-D 104276 data • PF coil currents • Te,i(), n(), v() • NB data TRANSP • Use n() directly • TSC derives e, I to reproduce Te and Ti • Turn on GLF23 in place of expt thermal diffusivities • Test GLF23 w/o ExB and w EXB shear stabilization 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 Te and Ti 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, Te/Ti,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 ofburning 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