IT/2-2 Experimental studies of ITER demonstration discharges George Sips MPI für Plasmaphysik, EURATOM-Association, Garching, Germany T.A. Casper 2, E.J. Doyle 3, G. Giruzzi 4, Y. Gribov 5, J. Hobirk 1, G.M.D. Hogeweij 6, L. Horton 1, A. Hubbard 7, I. Hutchinson 7, S. Ide 8, A. Isayama 8, F. Imbeaux 4, G.L. Jackson 9, Y. Kamada 8, C. Kessel 10, F. Kochl 11, P. Lomas 12, X. Litaudon 4, T.C. Luce 9, E. Marmar 7, M. Mattei 13, I. Nunes 14, N. Oyama 8, V. Parail 12, A. Portone 15, G. Saibene 15, R. Sartori 15, T. Suzuki 8, G. Tardini 1, S. Wolfe 7. The C-Mod team 7, the AUG team 1, the DIII-D team 9 and JET EFDA contributors 16. 1 MPI für Plasmaphysik, EURATOM-Association, Garching, Germany. 2 LLNL, PO Box 808, Livermore, CA 94550, USA. 3 Physics Dept. and PSTI, Univ. of California, Los Angeles, CA, USA. 4 Ass. Euratom-CEA, Cadarache Saint Paul Lez Durance, France. 5 ITER-IO, Cadarache 13108 Saint Paul Lez Durance, France. 6 FOM Rijnhuizen, Ass. EURATOM-FOM,, The Netherlands. 7 MIT, Plasma Science and Fusion Center, Cambridge, USA. 8 JAEA, 801-1 Muko-yama, Naka, Ibaraki 311-0193, Japan. 9 General Atomics, San Diego, USA 10 Plasma Physics Laboratory, Princeton University, Princeton, USA. 11 Association EURATOM-ÖAW/ATI, Vienna, Austria. 12 EURATOM/UKAEA Fusion Ass., Culham Science Centre, OX14 3DB, UK. 13 Ass. Euratom-ENEA-CREATE, Univ. degli Studi di Reggio Calabria, Italy. 14 Euratom/IST Fusion Ass., Centro de Fusao Nuclear, Lisboa, Portugal. 15 FUSION FOR ENERGY, Joint Undertaking, 08019 Barcelona, Spain. 16 JET-EFDA, Culham Science Centre, Abingdon OX14 3DB, UK. 22 nd IAEA Fusion Energy Conference, Geneva, 13-18 October 2008 Page 1
Motivation ITER: Unique combination of size, high power, long pulse,. stringent demands on the superconducting poloidal field system ( flux l i (3)=0.7-1.0 at 15MA (..OH Holtkamp, OV/2-1 Hawryluk, IT/1-2 Kessel, IT/2-3 ( 2008 ) Mattei Until recently ( IAEA FEC, 2006 ): NO detailed experimental data on the time evolution of ITER like plasma discharges. Dedicated experiments at C-Mod, AUG, DIII-D and JET on all aspects of the ITER discharge scenario. (in part coordinated by the SSO-TG of the ITPA) Romanelli, OV/1-2 Strait, OV/1-4 Zohm, OV/2-3 Marmar, OV/4-4 22 nd IAEA Fusion Energy Conference, Geneva, 13-18 October 2008 Page 2
ITER discharge scenario I p ( 400s ) Flat top 15MA, Std H-mode at q 95 =3 ( 400s ) Flat top I p rise (70s- 100s) I p ramp down ( 200-300s ) Breakdown Experimental verification of discharge evolution, l i (3) Joint,experiments consistent results? 700-800s 22 nd IAEA Fusion Energy Conference, Geneva, 13-18 October 2008 Page 3
Outline Plasma breakdown phase at low voltage. Current rise phase. Flat top phase at q 95 =3 (Q=10 reference scenario). (Poster: q 95 =4-4.5 for the hybrid scenario). Current ramp-down phase. Conclusions & Implications for ITER. 22 nd IAEA Fusion Energy Conference, Geneva, 13-18 October 2008 Page 4
Breakdown phase R 0 [m] B T [T] ECRH Power (type) E (V/m) Ohmic E (V/m) assisted C-Mod 0.68 5.4 - - 1.2-1.6 - AUG 1.65 1.7-3.2 105-140 GHz 0.3-1 MW (X2,O1) 0.6 0.2 EAST 1.70 2.0 - (LHCD) 0.3 MW (LH) 0.5 0.3 DIII-D 1.70 1.9-2.1 110 GHz 1-1.4 MW (X2) 0.43 0.21 KSTAR 1.80 1.5 84 GHz 0.35 MW (X2) - 0.4 TS 2.40 3.85 118 GHz 0.3-0.6 MW (O1) 0.3 0.15 JET 2.96 2.36 - (LHCD) 1.0-2.0 MW (LH) 0.23 0.18 JT-60U 3.32 3.5 110 GHz 0.4-2 MW (O1) (0.43) 0.26 ITER 6.20 5.3 127 (170) GHz 3 (20) MW (O1) 0.33 0.33 Un-assisted (ohmic) breakdown: E with machine size (JET~0.23V/m). Reliable breakdown ~0.2V/m with ECRH assist for all devices (EAST, JET use LHCD). Or 0.3-0.4V/m for a de-conditioned machine. 22 nd IAEA Fusion Energy Conference, Geneva, 13-18 October 2008 Page 5
Breakdown phase (example) AUG Zohm, OV/2-3 0.6 0.3 E axis [V/m] 0.33V/m AUG: Resistor-less operation: NEW for AUG: E=V/m 0.25V/m. 0 0.4 0 P ECRH [MW] D α [a.u] ( HFS ) Resonance: r/a=0.2, high field side ECRH pre-ionisation and assist: ( 2.3T ) X2: 105 GHz (1.7T), 140 GHz O1: 105 GHz (3.2T), HFS (R ec =1.45m). ITER using 170 GHz (O1) at full field (5.3T) or X2 at half field. 0.4 Ip 0.2 [MA] 0 0.0 0.1 0.2 Breakdown at low loop voltage: Successful, reproducible, reliable. Slow, controlled, rise of I p. No MHD reconnection low l i. 22 nd IAEA Fusion Energy Conference, Geneva, 13-18 October 2008 Page 6
Current rise phase Study several aspects of the current rise phase: The optimum plasma shape evolution, ohmic discharges, use of additional heating, I p tools available for l i control. used here l i =l i (3) =2 B p2 dv/((µ 0 I p ) 2 R) Heating? time 22 nd IAEA Fusion Energy Conference, Geneva, 13-18 October 2008 Page 7
Current rise phase (plasma shape) 3 2 JET, Ohmic I p [MA] 1.2 0.4 DIII-D, Ohmic I p [MA] Jackson, IT/P7-6 small bore full bore Small bore: l i ~1-1.2 1 0 small bore large bore full bore 1.5 li 1.0 0.5 X pt forms X pt forms Full bore l i =0.7-1.0 1.2 1.0 l i X pt forms X pt forms X pt forms 0.6 0 2 4 6 8 10 2.0 1.0 T e [kev] Sawteeth begin 16 12 8 q 95 4 0 0.2 0.4 0.6 1.0 1.2 control j-profile at start of the flat top ALL experiments show advantage of using full bore limiter + early divert: Large plasma size, Z eff control, density control, allow early heating.. 22 nd IAEA Fusion Energy Conference, Geneva, 13-18 October 2008 Page 8
Current rise phase (ohmic) l i 1.1 1.0 ohmic 15MA OH flux limit - Full bore rise, ohmic q 95 =3 - <n e >/n GW =0.2-0.4. - Low voltage breakdown 0.9 l i =-5 (stable) using fastest currents ramp rates. 0.7 Scaling ~ a 2, T e 3/2, Z eff ITER: ( 60s ~ Fast ramp: ~70s (limit Slow ramp: ~100s MHD unstable C-Mod AUG DIII-D JET ITER 22 nd IAEA Fusion Energy Conference, Geneva, 13-18 October 2008 Page 9
Current rise phase (example) l i 1.0 0.9 JET, di p /dt=0.28ma/s H-mode: 5MW ICRH H-mode: 6MW NBI H-mode: 10MW NBI Ohmic Flat top JET: Heating allows large variation of l i, independent of heating method. divert L-mode L H H-mode: Low l i =0.63- (broad T e (r), edge pedestal). 0.7 L H Start NBI ITB L H 0.6 Heating phase 0 2 4 6 8 JET and DIII-D: Active control of l i during current ramp phase to q 95 ~3 I p /dt, Heating power (~ 50% extra..) 22 nd IAEA Fusion Energy Conference, Geneva, 13-18 October 2008 Page 10
Current rise phase (heating) l i 1.1 1.0 ohmic L-mode H-mode C-Mod: ICRH AUG: NBI, ECRH DIII-D: NBI JET: ICRH, LHCD, NBI 0.9 Heating during limiter phase the Z eff 2-4, so l i ~0.9-1 For Z eff ~1.6-2 low l i, Heating dominates over current drive effects. 0.7 C-Mod AUG DIII-D JET ITER ITER: Modest ramp rate, with 5-10MW heating in L-mode (50% extra for control) 22 nd IAEA Fusion Energy Conference, Geneva, 13-18 October 2008 Page 11
Flat top, at q 95 =3 The experiments aimed at obtaining H 98 ~1 and β N ~1.8 as required for ITER to achieve Q~10 at 15MA: Evolution during the flat top phase T i (0)~T e (0) in most discharges. Typically P rad is between 0.4-0.6 of P tot No active ELM mitigation or radiation seeding was used. β rises within 2τ E, density rises within 4-6τ E I p [MA] /B T [T] P tot [MW] <n e > [10 19 m -3 ] β p / β N H 98 f GW P tot /P LH ( 1 ) l i (end ( FT of AUG 1.0 / 1.7 5.0 9.8 5 / 1.9 0.95 0.78 1.5-1.7 5 DIII-D 1.5 / 1.9 4.5 8.0 0.65 / 1.8 1.0 0.65 1.0-1.5 0.65 JET 2.5 / 2.35 19.0 6.4 0.7 / 1.8 0.95-0.98 0.70 1.9-2.1! 0 ITER 15 / 5.3 40+80 10.0 / 1.8 1.0 5 1.1-1.5? (1) P L-H [MW] =2.15*n e20 0.782 *B T 0.772 *a 0.975 R 1.0 ( 2008 ) Martin 22 nd IAEA Fusion Energy Conference, Geneva, 13-18 October 2008 Page 12
Flat top, at q 95 =3 (l i evolution) H-mode starts at beginning of flat top, current rise controlled l i =-0.9 1.2 1 AUG 1.2 1 Doyle, EX/1-3 ITER shape, low f ELM DIII-D 1.2 1 JET rise FT rise rise 0.6 0.6 FT 0.6 FT 0 1 2 0 1 2 3 4 0 5 10 15 DIII-D, infrequent ELMs significant edge pedestal contribution! Consistent equilibrium calculations for the different devices? 22 nd IAEA Fusion Energy Conference, Geneva, 13-18 October 2008 Page 13
Flat top, at q 95 =3 (l i evolution) ( tot H-L mode back-transition, in DIII-D and JET (stepping down P AUG 1.2 1.2 Doyle, EX/1-3 DIII-D 1.2 JET 1 1 FT 1 FT H-mode 0.6 0 1 2 0.6 0 1 2 3 4 H L 0.6 H L DIII-D: rise of l i then disrupts, JET: rise of l i to 1 within 3s ITER needs strategy for coping with H L transition at 15MA. New data useful for control simulations in ITER (seen IT-2-3) 0 5 10 15 22 nd IAEA Fusion Energy Conference, Geneva, 13-18 October 2008 Page 14
k Current ramp down 1.4 1.0 0.6 0.2 2.0 1.6 1.2 24 20 16 C-Mod k κ 1.5 κ 1.5 I p [MA] I OH [ka] 10% κ~1.85 12 1.0 1.2 1.4 1.6 1.8 2.0 l i C-Mod: 1MA/s ramp down keeps l i < 1.2 However, this ramp down requires ~10% increase in central OH current. Similar results for AUG, DIII-D, JET: Vertically stable and low l i in ramp down: Elongation reduced from 1.85 to 1.5. For ohmic or L-mode, slow decay (I p /dt) (ITER: ~ 300s, OH flux consumption?) Or use H-mode for significant part of ramp down phase (preliminary). ITER ramp down phase is important! experiments and modelling required. 22 nd IAEA Fusion Energy Conference, Geneva, 13-18 October 2008 Page 15
Conclusions Breakdown at ~0.3V/m: E=0.33V/m in ITER is sufficient for ohmic & using 170GHz system in ITER to provide assist. I p rise phase (test 15MA scenario): Variation of l i =0.63-1.05. Full bore, early X-point in ITER! Heating is essential to prepare burn phase. Flat top at q 95 =3: PF coils system should cope with l i =0.65-5 in H- mode, and cope with loss of H-mode during flat top. I p ramp down: Deserves more attention. Slow ramp down for ohmic or L-mode or use H-mode. Need viable ITER scenario + simulations. More remains to be done: Advanced scenarios, RF dominated plasma, simulation of burn control. 22 nd IAEA Fusion Energy Conference, Geneva, 13-18 October 2008
Extra slides (for Poster) 22 nd IAEA Fusion Energy Conference, Geneva, 13-18 October 2008
Hybrid scenario (q 95 ~4-4.5) ALL, including C-Mod (using LHCD, WILSON, EX/P6-21), produce target q(r) with q(0) 1 using ITER relevant current rise phase. JT-60U: Long pulse capability: β N >2.3 and H 98 ~1 for 23.1 s Low ρ*: New JET results with H 98 =1.2-1.4 DIII-D: ITER Demo Oyama, OV/1-3 Joffrin, EX/1-4Ra Doyle, EX/1-3 22 nd IAEA Fusion Energy Conference, Geneva, 13-18 October 2008 Page 18
Conclusions (I) Verification of ITER scenarios by C-Mod, AUG, DIII-D, JET and JT-60U: Breakdown at ~0.3V/m: Ohmic breakdown reliable for larges devices. Robust with ECRH assist for all devices. E=0.33V/m in ITER is sufficient & 170GHz system in ITER would be suitable to provide assist. I p rise phase (mainly at q 95 ~3, to test 15MA scenario): Best results using full bore, early divert plasma shape evolution. Ohmic: l i =0-1.05. Heated: l i =0.63-1.05. Lowest values for H-mode. Control and advanced scenarios are feasible Allow full bore, early X-point in ITER. Heating is very desirable. 22 nd IAEA Fusion Energy Conference, Geneva, 13-18 October 2008 Page 19
Conclusions (II) Flat top at q 95 =3: Q=10 conditions documented. PF coils system should cope with l i =0.65-5 in H-mode, And cope with loss of H-mode during flat top. PF coil operational range in ITER should cover l i =0.65-1.0 at 15MA. I p ramp down: Deserves more attention (many aspects to safe ramp down phase). If l i <1.6 during first half of ramp down slow decay (300s!) for ohmic or L-mode ramp down (OH flux consumption) or use H-mode. Need viable ITER scenario, backed-up by simulations. More remains to be done: Advanced scenarios, RF dominated plasmas and simulation of burn control. 22 nd IAEA Fusion Energy Conference, Geneva, 13-18 October 2008 Page 20