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JT-60SA Euratom 1 Japan Atomic Energy Agency, 1) University of Tokyo, 2) EFDA Close Support Unit, 3) EFDA-CSU-Barcelona, 4) CEA Cadarache, 5) JET-EFDA,

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Presentation on theme: "JT-60SA Euratom 1 Japan Atomic Energy Agency, 1) University of Tokyo, 2) EFDA Close Support Unit, 3) EFDA-CSU-Barcelona, 4) CEA Cadarache, 5) JET-EFDA,"— Presentation transcript:

1 JT-60SA Euratom 1 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 Title H. Tamai, T. Fujita, M. Kikuchi, K. Kizu, G. Kurita, K. Masaki, M. Matsukawa, Y. Miura, S. Sakurai, A.M. Sukegawa, Y. Takase 1), K. Tsuchiya, D. Campbell 2), S. Clement 3), J. J. Cordier 4), J. Pamela 5), F. Romanelli 6), and C. Sborchia 7) JT-60SA Euratom JT-60SA Euratom Prospective performances in JT-60SA towards the ITER and DEMO relevant plasmas O1A-A-360 24th SOFT Conference Sep. 2006, Warsaw, Poland

2 JT-60SA Euratom 2 OUTLINE Mission and Concept Plasma Performance Engineering Design Time Schedule Summary

3 JT-60SA Euratom 3 Mission and Concept Plasma Performance Engineering Design Time Schedule Summary

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

5 JT-60SA Euratom 5 Mission of JT-60SA 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 Experimental research with ITER relevant plasma configuration - high density operation - increased heating power, plasma current ITER similar configuration A=3.1,  95 =1.7,  95 =0.33, q 95 =3.0 Support to ITER divertor structure : TBD high- ,  shape for high-beta operation Time (s) NN Test of Plasma Facing Component - Compatibility test of reduced activation ferritic steel - Test candidate divertor modules - Sample station for plasma-material research Support to DEMO Sustain high beta (  N =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, … Exp. in JT-60U Target for JT-60SA

6 JT-60SA Euratom 6 Plasma Current I p (MA)3.5 / 5.5 Toroidal Field B t (T)2.59 / 2.72 Major Radius (m)3.16 / 3.01 Minor Radius (m)1.02 / 1.14 Elongation,  95 1.7 / 1. 83 Triangularity,  95 0.33 / 0. 57 Aspect Ratio, A3.10 / 2.64 Shape Parameter, S4.0 / 6.7 Safety Factor q 95 3.0 / 3.77 Flattop Duration100 s (8 hours) Heating & CD power41 MW x 100 s N-NBI34 MW ECRH 7 MW PFC wall load 10 MW/m 2 Neutron (year)4 x 10 21 D 2 main plasma + D 2 beam injection Basic Machine Parameters of JT-60SA ITER similar high-S for DEMO Gravity Support NBI Port Shear Panel Center Solenoid Stabilizing Plates Vacuum vessel Diagnostics Port In-vessel Coil Poloidal Field Coil Spherical Cryostat Toroidal Field Coil

7 JT-60SA Euratom 7 Heating & Current Drive Equipement N-NB (500 keV) co (2u)10 MW P-NB (85 keV) co (2u)4 MW ctr (2u)4 MW perp (8u)16 MW EC 110 GHz3 MW 140 GHz4 MW total41 MW 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… P3-B-336 : Y. Ikeda, et al. for 100s IpIp co IpIp N-NB (co) EC Tangential P-NB (ctr) Perpendicular P-NB Tangential P-NB (co) Remote Handling System Resonance layer of EC with two-frequency system

8 JT-60SA Euratom 8  Mission and Concept Plasma Performance Engineering Design Time Schedule Summary Plasma Performance

9 JT-60SA Euratom 9 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 Prospective estimation for ITER/DEMO relevant plasmas

10 JT-60SA Euratom 10 Feasibility for current drive scenario like an ITER hybrid operation Hybrid operation up to 3.7MA for 100s will be available. Current distribution (MA) Plasma current (MA) 4.0 3.0 2.0 1.0 0.0 beam driven bootstrap ohmic Flattop VlVl ACCOME-code analysis ITER similar configuration f GW =0.85, HH y2 =1.3, q 95 =3.1, P in =41MW

11 JT-60SA Euratom 11 High-  full non-inductive current drive scenario 2.4 MA full current drive with A = 2.65,  N = 4.4, f GW = 0.86, f BS = 0.70 and H H98y2 = 1.3 is possible with the 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). f GW f BS TeTe TiTi nene q

12 JT-60SA Euratom 12 Access for breakeven and high-  plasma with ITER and DEMO relevant parameters A=2.6, DN, q 95 ~3.5, HH 98y2 =1.5 3MA 1.5T 3.5MA 1.8T 4MA 2T 4.5MA 2.3T 5MA 2.5T 2.5MA 1.25T 25 MW n/n GW =0.8 40 MW n/n GW =0.8 5.5MA 2.8T Accessibility for high Q DT and high  N is enhanced with increased heating power. Normalized collision frequency e * 41MW, HH 98y2 =1.3 ITER (Steady state) DEMO (J05) 3MA, f GW =0.56 25MW, HH 98y2 =1.5 0.010 0.008 0.006 0.004 0.002 0.000 Normalized Larmor radius  i * 2.4MA, f GW =0.86 Non-dimensional parameters with ITER and DEMO relevant region are expected. A~2.6,  ~1.8, q 95 ~5.5,  N ~4 (2.4MA, f GW =0.86) break-even class plasma TFTR ITER JT-60 DIII-D FTU LHD C-Mod JET T i (0) (K) JT-60SA KSTAR Self-ignition Condition 10 21 10 20 10 19 EAST DEMO Break-even Condition 10 8 10 9 10 7 n D (0)  E (sec/m 3 ) collisionless /small normalized Larmor radius

13 JT-60SA Euratom 13 S IpIp aB T q 95  A -1 {1+  2 (1+2  2 )} Flexibility in aspect ratio and plasma shape for high-  plasma accessibility * M. R. Wade, et al., Phys. Plasmas 8 (2001) 2208. JT-60SA S=2.0-2.2 S=3.1-3.6 JT-60ASDEX-UJETDIII-D 6 5 4 3 2 Normalized beta  N 234567 Shape parameter S DIII-D Experiment * ITER JT-60 Target of JT-60SA  N : 3.5~5.5 S=2-8 S=3.0-5.4 S=2.3-7.4 Shape parameter Flexibility in S and A is extended, which enhances the research capability for high-  plasma operation.

14 JT-60SA Euratom 14 Achievable  N depends very much on the location of sector coil outside stabliser plates :  N ~3.8 inside stabiliser plates :  N ~5.6 ・ Sector coils are located on the port entrance in the present design (Analysis ongoing) RWM stabilisation by feedback control of sector coils (VALEN code analysis * ) Ideal limit Outside Inside Stabiliser plate Sector coil Present design Analysed model Controllability for resistive wall mode (RWM) * G. Kurita, et al., Nucl. Fusion 46 (2006) 383.

15 JT-60SA Euratom 15 -  ~1.83, dicertor leg ~ 0.8 m - Cryopanel under the dome (200 m 3 /s) - Vertical divertor target (60-80 ˚ ) Q total =12 MW,  ion = 1 x10 22 s -1,  puff =0.5 x10 22 s -1, S pump = 50 m 3 /s,  e =  i =1 m 2 /s, D=0.3 m 2 /s, C imp =1 % Heat & particle control with semi-closed divertor Divertor plasma simulation with SOLDOR/NEUT2D code Detachment control will be available with a strong gas puff. H. Kawashima, et al., Fus. Eng. Design 81 (2006) 1613.

16 JT-60SA Euratom 16 ☞ Engineering Design  Mission and Concept Plasma Performance Engineering Design Time Schedule Summary

17 JT-60SA Euratom 17 Cryostat Structure design Structure analysis Thermal shielding Engineering Design and Procurement Allocation First Wall PFC Ferrite (F82H) Structure design Baking/Cooling Divertor Target design Heat removal Particle pumping Cooling system Power Supply Cryogenic System Radiation Shielding R&D of shielding material Boron doped resin etc. Shielding analysis 2D/3D code Vacuum Vessel Structure design Structure analysis Baking Thermal shielding Superconducting Magnet Cable-in-conduit conductor Structure analysis Support structure Remote Handling System TF PF ECH System

18 JT-60SA Euratom 18 conductor EF TF CS TF CSEF strandNbTi Nb 3 SnNbTi conductor cable-in-conduit B max (T)6.4105.0 T op (K) I op (kA)26.52020 Superconducting Coils P1-E-328 : K. Tsuchiya, et al P1-E-286 : K. Kizu, et al.

19 JT-60SA Euratom 19 Vacuum vessel VV support leg structure VV is supported with 9 legs. VV has a double-wall structure. cylindrical: toroidally, polygonal:poloidally 140mm 24 Low cobalt SS316L (Boronic acid Water) Bird’s-eye view of vacuum vessel 3140 mm 9926 mm one turn resistance: ~15µΩ baking temp. : ~200˚C (TBD) Shielding water VV is covered with a thermal shield. Helium gas consists of 18 sections spring plates (AISI660) for baking Connection plate to restrain the horizontal swing of VV SS316 weight: ~300 ton without in-vessel components 3mm

20 JT-60SA Euratom 20 Plasma facing components First wall, divertor modules will be feasible for the maintenance by remote handling system. Mono-block target (15MW/m 2 ) 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. P2-F-341 : S. Sakurai, et al. Header (permanent) Bellows for thermal expansion of heat sink Pipe connection for laser cutter/welder Bolted exchangeable armor tiles Exchangeable heat sink ~0.3m ~1.6m Total thickness ~ 7cm Example of FW with exchangeable heat sink Example of divertor cassette with crank support Crank support for allowing large thermal expansion Width 10deg, Weight <500kg Divertor target Heat sink for bolted armor Divertor and dome geometry will be determined.

21 JT-60SA Euratom 21 SS304 SS316L Radiation Shield P3-J-302 : A. M. Sukegawa, et al. ● DD neutron emission rate

22 JT-60SA Euratom 22 Time Schedule  Mission and Concept Plasma Performance Engineering Design Time Schedule Summary

23 JT-60SA Euratom 23 Time Schedule 2006Completeion of Conceptual Design with the collaboration of JA and EU design teams 2007Detailed Design and Starts of Construction Schedule of construction and operation agreed in JA-EU WG Construction: 7 years + exploitation: 3 years

24 JT-60SA Euratom 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, n e, 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 JT-60SA Euratom 25 Thank you for your attention. Dziekuje !! Acknowledgement P1-E-286 : K. Kizu, et al. R&D of superconducting coil conductor P1-E-328 : K. Tsuchiya, et al. Superconducting coil system 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 Related Poster Presentation JT-60SA Euratom

26 JT-60SA Euratom 26

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