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1. Title Euratom Euratom 24th SOFT Conference Sep. 2006, Warsaw, Poland O1A-A-360 JT-60SA Prospective performances in JT-60SAtowards the ITER and DEMO relevant plasmas JT-60SA H. Tamai, T. Fujita, M. Kikuchi, K. Kizu, G. Kurita, K. Masaki, M. Matsukawa,Y. Miura, S. Sakurai, M. Sukegawa, Y. Takase1), K. Tsuchiya, D. Campbell2),S. Clement3), J. J. Cordier4), J. Pamela5), F. Romanelli6), and C. Sborchia7) Japan Atomic Energy Agency, 1) University of Tokyo, 2) EFDA Close Support Unit, 3) EFDA-CSU-Barcelona, 4) CEA Cadarache, 5) JET-EFDA, 6) EURATOM-ENEA, 7) Max-Planck Institut

2. OUTLINE • Mission and Concept • Plasma Performance • Engineering Design • Time Schedule • Summary

3. Mission and Concept • Mission and Concept • Plasma Performance • Engineering Design • Time Schedule • Summary

4. JT-60SA Project JT-60SA (JT-60 Super Advanced) = Combined project • Japanese national project (former JT-60SC or NCT) + • ITER satellite tokamak project Collaboration with Japan and EU fusion community

5. Mission of JT-60SA Support to DEMO Sustain high beta (bN=3.5-5.5) non-inductive CD plasma - Explore high beta regime above no-wall limit - Develop optimized integrated scenario for DEMO for shape, aspect ratio, SN/DN, current profile, MHD control, fuelling, pumping, divertor shape, … Test of Plasma Facing Component - Compatibility test of reduced activation ferritic steel - Test candidate divertor modules - Sample station for plasma-material research Target for JT-60SA bN Exp. in JT-60U Time (s) high-k, d shape for high-beta operation Support to ITER ITER similar configuration A=3.1, k95=1.7, d95=0.33, q95 =3.0 Experimental research with ITER relevant plasma configuration - high density operation - increased heating power, plasma current divertor structure : TBD Support to ITER - ITER construction phase • optimization of operation scenario, auxiliary system • training of scientists, engineers and technicians - ITER operation phase • support further development of operating scenarios and understanding of physics issues • Test of possible modifications before their implementation Support to DEMO - to explore operational regimes and issues complementary to those being addressed in ITER • steady-state operation • advanced plasma regimes (high-beta plasma) • control of power fluxes to wall

6. Basic Machine Parameters of JT-60SA Spherical Cryostat Center Solenoid Poloidal Field Coil Toroidal Field Coil Diagnostics Port Shear Panel Stabilizing Plates NBI Port Vacuum vessel In-vessel Coil Gravity Support high-S for DEMO ITER similar D2 main plasma + D2 beam injection

7. Heating & Current Drive Equipement Ip Perpendicular P-NB Tangential P-NB (ctr) Resonance layer of EC with two-frequency system co Remote Handling System Ip EC N-NB (co) Tangential P-NB (co) • Increased injection power of N-NB, and EC • P-NB : balanced injection for toroidal rotation control • EC : two-frequency system for flexible control of CD, MHD… for 100s P3-B-336 : Y. Ikeda, et al.

8. Plasma Performance Mission and Concept • Plasma Performance • Engineering Design • Time Schedule • Summary

9. Prospective estimation for ITER/DEMO relevant plasmas • Capability to perform operation scenarios - standard operation - hybrid operation - full non-inductive CD operation • Break-even class plasmas • High-beta plasma accessibility - shape and aspect ratio - MHD control • Heat and particle control - divertor plasma performance

10. Feasibility for current drive scenario like an ITER hybrid operation 4.0 beam driven 3.0 bootstrap Current distribution (MA) 2.0 Flattop Vl ohmic 1.0 0.0 3.0 3.2 3.4 3.6 3.8 4.0 4.2 Plasma current (MA) ACCOME-code analysis ITER similar configuration fGW=0.85, HHy2=1.3, q95=3.1, Pin=41MW Hybrid operation up to 3.7MA for 100s will be available.

11. High-b full non-inductive current drive scenario Te Ti ne q • 2.4 MA full current drive with A = 2.65, bN = 4.4, fGW = 0.86, fBS = 0.70 and HH98y2 = 1.3 is possible withthe total heating power of 41 MW. • NNB is shifted down by 0.6 m for off-axis CD in order to form a weak reversed shear q profile. • Normalized parameters are close to those required in DEMO (J05, slim CS). • RWM will be controlled by non-axisymmetric feedback coils (sector coils). fGW fBS

12. Access for breakeven and high-b plasma with ITER and DEMO relevant parameters 40 MW n/nGW=0.8 2.5MA 1.25T 5.5MA 2.8T 3MA 1.5T 3.5MA 1.8T 4MA 2T 25 MW n/nGW=0.8 4.5MA 2.3T 5MA 2.5T break-even class plasma collisionless /small normalized Larmor radius 1021 Self-ignition Condition Break-even Condition DEMO ITER JT-60SA 1020 KSTAR nD(0) tE (sec/m3) JT-60 FTU TFTR C-Mod JET DIII-D 1019 LHD EAST 109 108 107 Ti(0) (K) Accessibility for high QDT and high bN is enhanced with increased heating power. Non-dimensional parameters with ITER and DEMO relevant region are expected. A=2.6, DN, q95~3.5, HH98y2=1.5 A~2.6, k~1.8, q95~5.5, bN~4 (2.4MA, fGW=0.86) 0.010 0.008 0.006 0.004 0.002 0.000 25MW, HH98y2=1.5 2.4MA, fGW=0.86 3MA, fGW=0.56 Normalized Larmor radius ri* 41MW, HH98y2=1.3 ITER (Steady state) DEMO (J05) 0.00 0.02 0.04 0.06 0.08 0.10 0.12 Normalized collision frequency ne*

13. Flexibility in aspect ratio and plasma shape for high-b plasma accessibility JT-60SA JT-60 ASDEX-U JET DIII-D 3 Divertor pumping(m /s) S=2.3-7.4 3.5 ≥100 <100 S=3.1-3.6 S=3.0-5.4 S=2.0-2.2 S=2-8 Double null ITER Single null Target of JT-60SA bN: 3.5~5.5 6 Aspect ratioA Normalized beta bN 3.0 5 DIII-D Experiment* 4 JT-60 ITER 3 2.5 2 2 3 4 5 6 7 Shape parameter S 4 5 6 7 8 Shape parameter S *M. R. Wade, et al., Phys. Plasmas 8 (2001) 2208. Flexibility in S and A is extended, which enhances the research capability for high-b plasma operation. Shape parameter Ip q95µ A-1{1+2(1+22)} S aBT

14. Controllability for resistive wall mode (RWM) Ideal limit Present design Stabiliser plate Analysed model Sector coil RWM stabilisation by feedback control of sector coils (VALEN code analysis*) *G. Kurita, et al., Nucl. Fusion 46 (2006) 383. Achievable bN depends very much on the location of sector coil outside stabliser plates : bN~3.8 inside stabiliser plates : bN~5.6 ・Sector coils are located on the port entrance in the present design (Analysis ongoing) Outside Inside

15. Heat & particle control with semi-closed divertor Divertor plasma simulation with SOLDOR/NEUT2D code Qtotal=12 MW, Gion= 1x1022s-1, Gpuff =0.5 x1022s-1 , Spump = 50 m3/s,e=i=1 m2/s, D=0.3 m2/s ,Cimp=1 % - ~1.83, dicertor leg ~ 0.8 m - Cryopanel under the dome (200 m3/s) - Vertical divertor target (60-80˚) Detachment control will be available with a strong gas puff. H. Kawashima, et al., Fus. Eng. Design 81 (2006) 1613.

16. Engineering Design ☞ Mission and Concept • Plasma Performance • Engineering Design • Time Schedule • Summary

17. Engineering Design and Procurement Allocation Cryostat •Structure design • Structure analysis • Thermal shielding Superconducting Magnet • Cable-in-conduit conductor • Structure analysis • Support structure TF PF Cryogenic System Vacuum Vessel •Structure design • Structure analysis • Baking • Thermal shielding Power Supply ECH System First Wall •PFC Ferrite (F82H) •Structure design • Baking/Cooling Radiation Shielding • R&D of shielding material Boron doped resin etc. • Shielding analysis 2D/3D code Divertor •Target design • Heat removal • Particle pumping • Cooling system Remote Handling System

18. Superconducting Coils TF CS EF strand NbTi Nb3Sn NbTi conductor cable-in-conduit Bmax (T)6.4 10 5.0 Top (K)4.6 5.0 4.8 Iop (kA)26.5 20 20 CS TF EF conductor P1-E-328 : K. Tsuchiya, et al P1-E-286 : K. Kizu, et al.

19. Vacuum vessel consists of 18 sections VV has a double-wall structure. cylindrical: toroidally, polygonal:poloidally 24 24 140mm weight: ~300 ton without in-vessel components one turn resistance: ~15µΩ Shielding water (Boronic acid Water) baking temp. : ~200˚C (TBD) Helium gas Low cobalt SS316L 9926 mm 3mm VV is covered with a thermal shield. SS316 3140 mm VV is supported with 9 legs. Connection plate to restrain the horizontal swing of VV spring plates (AISI660) for baking VV support leg structure Bird’s-eye view of vacuum vessel

20. Plasma facing components ~0.3m Exchangeable heat sink Total thickness ~ 7cm Bolted exchangeable armor tiles ~1.6m Bellows for thermal expansion of heat sink Pipe connection for laser cutter/welder Header (permanent) • First wall, divertor modules will be feasible for the maintenance by remote handling system. • Mono-block target (15MW/m2) will be adopted after the relliability is established bysignficant R&D. • Exchange with full metal plasma facing components will be decided after experimental and computational analyses. Width 10deg, Weight <500kg Heat sink for bolted armor Crank support for allowing large thermal expansion Divertor and dome geometry will be determined. Divertor target Example of FW with exchangeable heat sink Example of divertor cassette with crank support P2-F-341 : S. Sakurai, et al.

21. Radiation Shield SS316L SS316L SS304 SS304 ● DD neutron emission rate P3-J-302 : A. M. Sukegawa, et al.

22. Time Schedule Mission and Concept • Plasma Performance • Engineering Design • Time Schedule • Summary

23. Time Schedule Schedule of construction and operation agreed in JA-EU WG Construction: 7 years + exploitation: 3 years Completeion of Conceptual Design with the collaboration of JA and EU design teams Detailed Design and Starts of Construction

24. Summary • Prospective performance in JT-60SA plasma is estimated on the viewpoint of ITER / DEMO support. • ITER operation scenario will be investigated with the ITER similar configuration (shape, ne, etc.) by increased heating power and plasma current. • Steady-state, high beta plasma controllability will be foreseen (support to DEMO). • Engineering design will be performed with JA and EU, and the construction is planned to start next year.

25. Acknowledgement Related Poster Presentation JT-60SA P1-E-328 : K. Tsuchiya, et al. Superconducting coil system Euratom P1-E-286 : K. Kizu, et al. R&D of superconducting coil conductor Dziekuje !! P2-F-341 : S. Sakurai, et al. Plasma facing components P3-J-302 : A. M. Sukegawa, et al. Safety design P3-B-336 : Y. Ikeda, et al. NBI system Thank you for your attention.