1 OVERVIEW OF ITER PHYSICS V. Mukhovatov 1, M. Shimada 1, A.E. Costley 1, Y. Gribov 1, G. Federici 2,A.S. Kukushkin 2, A. Polevoi 1, V.D. Pustovitov 3, Y. Shimomura 1, T. Sugie 1, M. Sugihara 1, G. Vayakis 1 1 International Team, ITER Naka Joint Work Site, Naka, Ibaraki, Japan 2 International Team, ITER Garching Joint Work Site, Garching, Germany 3 Nuclear Fusion Institute, RRC Kurchatov Institute, Moscow, Russia ITER V. Mukhovatov et al., 30th EPS Conf. on Control. Fusion and Plasma Phys., July 7-11, 2003, St Petersburg, Russia

2 Contents V. Mukhovatov et al., 30th EPS Conf. on Control. Fusion and Plasma Phys., July 7-11, 2003, St Petersburg, Russi  Introduction  ELMy H-mode l Operational limits l Confinement l Instabilities  Improved H-mode  Internal Transport Barriers l Formation l Performance l Control  Summary

3 Introduction V. Mukhovatov et al., 30th EPS Conf. on Control. Fusion and Plasma Phys., July 7-11, 2003, St Petersburg, Russia  Predictive methodologies for tokamak Burning Plasma Experiment (BPX) have been summarized in the ITER Physics Basis (IPB) published in 1999 [Nucl. Fusion 39 (1999) ].  In recent years, significant progress has been achieved in many areas of tokamak physics  New achievements have had significant impact on new ITER design (stronger shaping, methods to suppress NTMs and RWMs)  This talk reviews the ITER physics basis taking account of the recent progress in tokamak studies

4 Major ITER-Relevant Confinement Modes V. Mukhovatov et al., 30th EPS Conf. on Control. Fusion and Plasma Phys., July 7-11, 2003, St Petersburg, Russia  H-mode (High Confinement Mode) associated with formation of edge transport barrier (ETB) l Reference mode for ITER inductive high-Q operation  Improved H-mode l Candidate mode for inductive and/or hybrid ITER operation  Advanced Tokamak (AT) mode associated with formation of Internal Transport Barrier (ITB) l Candidate mode for steady- state ITER operation

5 Physics Rules for Selection of ITER Design Parameters V. Mukhovatov et al., 30th EPS Conf. on Control. Fusion and Plasma Phys., July 7-11, 2003, St Petersburg, Russia  Q ≥ 10 Q = 5P  /P aux  ELMy H-mode reference operation mode  ITERH-98P(y,2) scaling for energy confinement time  Safety factor q 95 ≥ 2.5 q 95  (5B/I)(  a 2 /R)  Electron density n e ≤ n G n G = I/(  a 2 ), Greenwald density  Normalized beta  N ≤ 2.5 [  N =  (%)(aB/I)]  Strong plasma shaping  sep = 1.85,  sep = 0.48  Heating power P ≥ 1.3 P L-H P = P  + P aux - P rad P L-H is H-mode power thresh.

6 ELMy H-MODE

7 ELMy H-mode V. Mukhovatov et al., 30th EPS Conf. on Control. Fusion and Plasma Phys., July 7-11, 2003, St Petersburg, Russia  ELMy H-mode: H-mode with bursts of Edge Localized Modes (ELMs) Reference ITER mode for inductive high-Q operation Robust mode observed in all tokamaks under wide variety of conditions at heating power above the threshold, P>P L-H Good prospects for long-pulse operation >20 years of studies Rich experimental database High confidence that ELMy H-mode will be obtained in ITER

8 Energy Confinement Projections for ELMy H-mode in ITER V. Mukhovatov et al., 30th EPS Conf. on Control. Fusion and Plasma Phys., July 7-11, 2003, St Petersburg, Russia  Three approaches (discussed in details in IPB) predict compatible results for ITER reference high Q scenario Transport models based on empirical scalings for the energy confinement time Physics-based transport models Dimensionless analysis

9 ITER Reference Scalings V. Mukhovatov et al., 30th EPS Conf. on Control. Fusion and Plasma Phys., July 7-11, 2003, St Petersburg, Russia ITERH-98P(y.2) confinement scaling ITER:  E = 3.66s ±14% [2.78, 4.83]s 95% nonlinear interval estimate O.Kardaun, Nucl. Fusion 42 (2002) 841 J A Snipes et al PPCF 42 (2000) A299 H-mode power threshold scaling ITER: P L-H = 49 MW [28.4, 84.1]MW 95% interval estimate

10 Effect of Plasma Dilution with Helium  ITER performance depends on plasma dilution with He  B2/Eirene code: Helium content in ITER plasma reduces due to Helium atom elastic collisions with D/T ions  Reduction of He content improves ITER performance V. Mukhovatov et al., 30th EPS Conf. on Control. Fusion and Plasma Phys., July 7-11, 2003, St Petersburg, Russia 1/2D ITINT1.SAS code with P sep ≥ P L-H O.J.W.F. Kardaun NF 42 (2002) 841

11 Theory Based Transport Models V. Mukhovatov et al., 30th EPS Conf. on Control. Fusion and Plasma Phys., July 7-11, 2003, St Petersburg, Russia  WEILAND, MMM, GLF23 and IFS/PPPL transport models  Transport driven by drift wave turbulence  Detailed treatment is somewhat different  Boundary conditions taken from experiments or from empirical or semi-empirical scalings  Reasonable agreement with experimental data for plasma core

12 ITER Predictions by Physics Based Models V. Mukhovatov et al., 30th EPS Conf. on Control. Fusion and Plasma Phys., July 7-11, 2003, St Petersburg, Russia Pedestal scalings (a) J G Cordey, et al 19th FEC Lyon (b) J G Cordey, et al 19th FEC, Lyon (c) M Sugihara, et al NF 40 (2002) 1743 (d) A H Kritz, et al 29th EPS D (e) M Sugihara, et al submitted to PPCF (g) K S Shaing T H Osborne et al 19th FEC, Lyon

13 ITER Predictions by Physics Based Models V. Mukhovatov et al., 30th EPS Conf. on Control. Fusion and Plasma Phys., July 7-11, 2003, St Petersburg, Russia Predictions for ITER by different models at the same input parameters (G. Pereverzev et al. 29th EPS 2002 P-1072)

14 Edge Pedestal in ELMy H-mode V. Mukhovatov et al., 30th EPS Conf. on Control. Fusion and Plasma Phys., July 7-11, 2003, St Petersburg, Russia J.G.Cordey et al IAEA Lyon Conf M Sugihara et al, submitted tp PPCF 2003 Two-term confinement scalings for thermal energy W = W core + W ped Edge temperature gradient limited by thermal conduction ITER: W ped = 174 MW T ped = 5.2 keV Edge gradient limited by ELMs (MHD limit): ITER: W ped = 98 MW T ped ≈ 3.0 keV

15 Non-Dimensional Confinement Scalings V. Mukhovatov et al., 30th EPS Conf. on Control. Fusion and Plasma Phys., July 7-11, 2003, St Petersburg, Russia  GyroBohm like scalings have been found in experiments with ELMy H-mode: B  E  (  *)  0.03 ( * ) in DIII-D B  E  (  *) -2.7  ( * ) in JET (  *=  i /a)  JET DT discharge with all dimensionless parameters,  *, q, R/a,  etc, except  *, the same as ITER: JET #42983:  *= JET-like ITER:  *= ==> Q =

16 High Performance H-Modes at High Density Demonstrated  One of the major achievements in recent tokamak experiments was demonstration of good confinement in H-mode at high plasma density required for ITER, i.e. H 98(y,2) = 1 at n ≥ 0.85 n G  There are several ways to improve confinement at high density Increase in plasma triangularity; gentle gas fuff Impurity seeding High field side pellet fueling V. Mukhovatov et al., 30th EPS Conf. on Control. Fusion and Plasma Phys., July 7-11, 2003, St Petersburg, Russia

17 V. Mukhovatov et al., 30th EPS Conf. on Control. Fusion and Plasma Phys., July 7-11, 2003, St Petersburg, Russia Good Confinement at High Density Energy confinement reduces with density but improves with plasma triangularity  or shaping parameter q 95 /q cyl H (y,2)corr = ln(q 95 /q cyl ) n/n G (n/n ped -1) ITER: H (y,2)corr =0.91 at n/n ped =1; H (y,2)corr =1.05 at n/n ped =1.3 H JET ITER

18 Power and Particle Control in ITER V. Mukhovatov et al., 30th EPS Conf. on Control. Fusion and Plasma Phys., July 7-11, 2003, St Petersburg, Russia  B2/Eirene code: steady state divertor power loads are within the proven limits  He density at the separatrix reduces by 3-5 times due to elastic collisions of He atoms with D/T ions A S Kukushkin, H D Pacher PPCF 44 (2002) 943

19 Major Instabilities in ELMy H-mode V. Mukhovatov et al., 30th EPS Conf. on Control. Fusion and Plasma Phys., July 7-11, 2003, St Petersburg, Russia  Sawteeth  Edge localized modes (ELMs)  Neoclassical tearing modes (NTMs)  Alfven instabilities  Disruptions

20 EDGE LOCALIZED MODES (ELMs)

21 H-mode Regimes with Smaller ELMs  Expected energy fluxes on the ITER divertor associated with ELMs are close to being marginal for an acceptable divertor target life time  There are alternative high confinement modes with small ELMs found at q 95 > and high triangularity H-mode with ‘grassy’ or ‘minute’ ELMs in DIII-D and JT-60U Enhanced D  (EDA) mode in Alcator C-Mod with quasi- coherent density fluctuations Advanced H-mode with Type II ELMs in ASDEX-U Impurity seeded H-mode in JET with reduced Type I ELMs High density H-mode with rear small ELMs in JET Quiescent Double Barrier (QDB) H-mode in DIII-D  ELM mitigation with frequent pellet injection is promising V. Mukhovatov et al., 30th EPS Conf. on Control. Fusion and Plasma Phys., July 7-11, 2003, St Petersburg, Russia

22 ELM Mitigation Using Pellet Injection. A. Herrmann PSI 2002 ? V. Mukhovatov et al., 30th EPS Conf. on Control. Fusion and Plasma Phys., July 7-11, 2003, St Petersburg, Russia 4Hz pellet injection in ITER can reduce the energy loss per ELM to acceptable level (A Polevoi et al 19 FEC Lyon 2002) ELM induced energy loss is reduced in ASDEX Upgrade at sufficiently high frequency of pellet injection (P Lang, 2002)

23 NEOCLASSICAL TEARING MODES (NTMs)

24 Neoclassical Tearing Modes (NTMs) V. Mukhovatov et al., 30th EPS Conf. on Control. Fusion and Plasma Phys., July 7-11, 2003, St Petersburg, Russia  Neoclassical tearing modes (NTMs) are induced by reduction of bootstrap current inside magnetic islands  Deteriorate confinement and determine the lowest beta limit  NTMs methastable: ‘seed’ islands are required  NTM can be stabilized with localized current drive within magnetic island

25 V. Mukhovatov et al., 30th EPS Conf. on Control. Fusion and Plasma Phys., July 7-11, 2003, St Petersburg, Russia Neoclassical Tearing Modes (NTMs)  Complete 3/2 NTM suppression demonstrated (AUG, DIII-D, JT-60U) with localized ECCD  Complete 2/1 NTM suppression demonstrated (DIII-D)  Real-time ECCD position control demonstrated (DIII-D)

26 V. Mukhovatov et al., 30th EPS Conf. on Control. Fusion and Plasma Phys., July 7-11, 2003, St Petersburg, Russia Suppression of NTMs in ITER  Extrapolation to ITER: P ECCD = (30 ± 15) MW (G Giruzzi and H Zohm, ITPA MHD Meering, Naka, Feb 2002)  Early injection would enable NTM stabilization with P ECCD < 20 MW  ITER design: P ECCD = 20 MW A Zvonkov, 2000 m/n = 2/1

27 DISRUPTION MITIGATION

28 Disruption Mitigation  Mechanical loads during disruptions are within the design limits (confirmed by DINA) (M.Sugihara et al, this Conference)  Promising disruption mitigation technique DIII-D: High-pressure noble gas jet injection (D G Whyte FEC 2002, Lyon) V. Riccardo, this Session V. Mukhovatov et al., 30th EPS Conf. on Control. Fusion and Plasma Phys., July 7-11, 2003, St Petersburg, Russia

29  Preliminary modeling: the technique is feasible for ITER  Operation space limited by melting/ablating the first wall Noble Gas Jet Injection in ITER V. Mukhovatov et al., 30th EPS Conf. on Control. Fusion and Plasma Phys., July 7-11, 2003, St Petersburg, Russia t (ms) (10 21 m -3 ) ITER-98 D G Whyte 19th FEC 2002, Lyon

30 IMPROVED H-MODE

31  Regime with lower current (higher q 95 ) would be beneficial to reduce disruption forces and for access to benign (Type II) ELM regime but requires improved confinement  Recently ASDEX Upgrade, DIII-D and JET demonstrated a possibility to obtain plasmas with improved confinement, H98(y,2) = , at q 95 = (correspond to I = MA in ITER) Q=10 Scenario at Reduced Current V. Mukhovatov et al., 30th EPS Conf. on Control. Fusion and Plasma Phys., July 7-11, 2003, St Petersburg, Russia

32 V. Mukhovatov et al., 30th EPS Conf. on Control. Fusion and Plasma Phys., July 7-11, 2003, St Petersburg, Russia Advanced H-mode with Type II ELMs  18 MW / m 2  6 MW / m 2 inner divertor outer divertor No sawteeth q(0) ≥1  N = 3.5 q 95 = 3.6 H 98(y,2) = 1.3 n = n G  t = 40  E Low divertor heat load (Type II ELMs) ASDEX Upgrade

33 INTERNAL TRANSPORT BARRIERS (ITBs)

34 Steady-State Q≥5 Operation in ITER V. Mukhovatov et al., 30th EPS Conf. on Control. Fusion and Plasma Phys., July 7-11, 2003, St Petersburg, Russia  Requirements H 98P(y.2) > High beta  N > 2.6 High bootstrap current fraction,f BS ≥50%  Advanced Tokamak Mode Regimes with Internal Transport Barriers (ITBs) Weak or negative magnetic shear Resistive wall mode stabilization

35 ITB Power Threshold V. Mukhovatov et al., 30th EPS Conf. on Control. Fusion and Plasma Phys., July 7-11, 2003, St Petersburg, Russia  The rarefaction of resonance surfaces at low/zero magnetic shear helps ITB formation while the barrier width is probably controlled by the ExB shear  JET and ASDEX-U indicate importance of rational q in the vicinity of zero magnetic shear [E Joffrin et al 19th FEC Lyon 2002]  The target plasmas with weak or negative magnetic shear require lower heating power for ITB formation [G T Hoang et al, 29th EPS Conf. 2002]

36 V. Mukhovatov et al., 30th EPS Conf. on Control. Fusion and Plasma Phys., July 7-11, 2003, St Petersburg, Russia 50 Real-Time Control of ITBs in JET

37 JT-60U: ITB and Current Hole V. Mukhovatov et al., 30th EPS Conf. on Control. Fusion and Plasma Phys., July 7-11, 2003, St Petersburg, Russia Current hole and ITB at strong negative shear has been sustained for ~5 s in JT60-U at I = 1.35 MA, q 95 =5.2, H H98y,2 ~1.5,  N ~ 1.7 T(r) and n(r) are flat inside the current hole Transiently: I=2.6 MA, q 95 =3.3,  E =0.89 s, Q eq =1.2 H H98y,2 ~1.5,  N ~ 1.6 n e (0) = m -3

38 RESISTIVE WALL MODES (RWMs)

39  DIII-D: Dynamic error field corrections by feedback control allows rotational stabilization of RWMs  N =  N (ideal wall) ~ 2  N (no-wall limit) at w rot > 2%  Alfven  DIII-D: Negative central shear plasma f BS = 65%, f non-ind = 85%,  T ≥ 4% (E J Strait et all 19th FEC Lyon 2002) Suppression of Resistive Wall Modes V. Mukhovatov et al., 30th EPS Conf. on Control. Fusion and Plasma Phys., July 7-11, 2003, St Petersburg, Russia

40  Extrapolation to ITER Model developed taking account realistic vessel and coil geometry and plasma rotation (A Bondeson, next report) Side correction coils will be used for RWM stabilization (similar to that in DIII-D) Suppression of RWM in ITER V. Mukhovatov et al., 30th EPS Conf. on Control. Fusion and Plasma Phys., July 7-11, 2003, St Petersburg, Russia C  = 0.8 is achievable

41 V. Mukhovatov et al., 30th EPS Conf. on Control. Fusion and Plasma Phys., July 7-11, 2003, St Petersburg, Russia Requirements for Power Reactor  Analysis study suggests that it is possible to achieve most normalized plasma parameters in ITER to enable projection to fusion power reactor, i.e. demonstration of P fus ~0.7GW and simulation of P fus ~ 1 GW (M.Shimada, this Conference, Thursday 10 July)

42 Requirements for Plasma Measurements V. Mukhovatov et al., 30th EPS Conf. on Control. Fusion and Plasma Phys., July 7-11, 2003, St Petersburg, Russia  The requirements for plasma and first wall measurements on ITER are well developed and many diagnostic systems have been designed to an advanced level  Solutions to many of the difficult implementation issues that arise on a DT machine have been found, and design and R&D is in progress on outstanding issues  It is believed that the measurements necessary for the machine protection and basic plasma control can be made at the required level of accuracy etc, and also many of those now identified as necessary to support the advanced operation  There are several papers on ITER diagnostics presented in the diagnostic sessions on Thursday and Friday afternoons including an overview oral by A Costley on long pulse issues in ITER diagnostics

43 Summary - I V. Mukhovatov et al., 30th EPS Conf. on Control. Fusion and Plasma Phys., July 7-11, 2003, St Petersburg, Russia  The reference plasma parameters required for inductive high-Q operation in ITER (  N = 1.8, q 95 = 3, H 98(y,2) = 1, n/n G = 0.85) are demonstrated on present machines  The feasibility of achieving Q ≥10 in H-mode predicted by transport model based on empirical confinement scaling is confirmed by dimensionless analysis and theory-based transport modeling  Active control of NTMs and mitigation of ELMs and disruptions may be necessary. Relevant control and mitigation techniques suggested and tested. Extrapolation to ITER needs further work

44 Summary - II V. Mukhovatov et al., 30th EPS Conf. on Control. Fusion and Plasma Phys., July 7-11, 2003, St Petersburg, Russia  Requirements for ITER steady-state Q≥5 operation (  N > 2.6, H 98(y,2) > 1.3, f BS > 0.5, n ~ n G ) developed. Normalized parameters demonstrated in experiments  More sophisticated control schemes (i.e. current and pressure profiles) will be necessary for steady state operation. Such schemes are under development  Achievement of more demanding normalized parameters (  N > 3.6) and high fusion power, 700MW, necessary to facilitate extrapolation of plasma performance to fusion power reactor is under study and looks possible

45 LIST OF ITER IT REPORTS AT THIS CONFERENCE V. MukhovatovOverview of ITER Physics (Wednesday, July 9) I-3.3A M. Shimada High Performance Operation in ITER (Thursday, July 10) P M. Sugihara Examination on Plasma Behaviors during Disruptions on Existing Tokamaks and Their Extrapolations to ITER (Tuesday, July 8) P A.S. KukushkinEffect of Carbon Redeposition on the Divertor Performance in ITER (Thursday, July 10) P A. Costley Long Pulse Operation in ITER: Issues for Diagnostics (Friday, July 11) O-4.1D K. Itami Study of Multiplexing Thermography for ITER Divertor Targets (Friday, July 11) P-4.62 T. Kondoh Toroidal Interferometer/Polarimeter Density Measurement System for Long Pulse Operation on ITER (Friday, July 11) P-4.64 T. Kondoh Prospects for Alpha-Particle Diagnostics by CO2 Laser Collective Thomson Scattering on ITER (Friday, July 11) P-4.65 T. Sugie Spectroscopic Measurement System for ITER Divertor Plasma: Divertor Impurity Monitor (Friday, July 11) P-4.63 C. WalkerErosion and Redeposition on Diagnostic Mirrors for ITER: First Mirror Test at JET and TEXTOR (Friday, July 11) P-4.59 C.I. WalkerITER Generic Diagnostic Components and Systems for Integration (Friday, July 11) P-4.61