Colloquium of Max Planck Institute fur Physics March 8, 2001

JT-60 Modification Plan for Long Pulse Advanced Research

Presented by M. Kikuchi

Japan Atomic Energy Research Inst. Naka Fusion Research Establishment OUTLINE OF TALK

• Fusion Development Strategy • Progress of Confinement and ITER as one step to DEMO • Advanced tokamak researches • Scientific achievements of JT-60 and its Infrastructures • Proposed Modifications • New research area and Significance of Long Pulse Experiments - Q~1 sustainment - Long sustainment of full CD plasma

- Long sustainment of high bN plasma - Advanced Divertor - Technological Research Subjects • Summary

1 Fusion Development Strategy

The subcommittee on fusion development strategy (N. Inoue chair) identified fusion development strategy in its report to the ITER special committee.

Commercializa Second phase Third phase Fourth phase tion pease

Helical device, Reverse field pinch, Mirror, Inertial confinement Advanced confinement

Advanced and complementary Research and Development

Electricity Generation

Ignition & Tokamak Tokamak Long-burn Experimental Review Prototype Economic JT-60 Reactor Reactor aspects (ITER) Decision system of prototype reactor Development of major components with enlarge sizes and improved performances Test using Experimental Reactor

Long-term development necessary for a fusion reactor

• Reprot on the technical feasibility of fusion energy, May17,2000 2 Progresses of plasma confinement is extremely fast (similar to DRAM) and is ready to high Q burning expl.

15 Self-ignition 10 Break-even 23 12 10 ITER SSTR 10 ITER Tokamak '90

TFTR(US) JT-60U(J) Q=1.25 JT-60U (J) 14 Break-even 10 JT-60U(J) 21 JET (EU) 16G FTU(EU) Highest Temp. 10 DIII-D(US) TFTR 4G C-Mod(US) JT-60U(RS)0.8s Doublet-III 1G 9 LHD('00) TFTR(US) 10 JT-60UFull CD 256M JET(EU) 19 13 Break-even n T PLT 10 10 E 64M Heliotron E Tokamak '80 ASDEX-U(EU) 16M Memory size (Kyoto U.) GAMMA-10 TFR (Tsukuba U.) of DRAM JFT-2M T-4 4M 17 6 1012 10 1M 10 Tokamak '70 256K JFT-2 64K JFT-2a 15 16K 11 10 4K 10 Tokamak '60 1K 3 Helical device 10 Mirror device 13 RFP device 10 Achieved regime of RFP 2020 1010 Others are 1960 1970 1980 1990 2000 2010 Fiscal year 7 8 106 10 10 109 Central ion Temperature (deg.) 3 ITER-FEAT as one step to DEMO

[1] Tokamak is ready for a -dominant burning experiments". ITER-FEAT design is sufficient to sustain burning plasma with Q=10- infinity.

[2] Tokamak is most advanced in scientific understanding. ITER-FEAT provides new scientific regime of self-heated plasma.

Tokamak still needs improvement to become an attractive fusion reactor. ITER-FEAT design places more emphasis on steady-state operation expecting future 0 5 10 15 20 R (m) advances in tokamak researches. 4 Advanced Tokamak Researches

[1] AT research : Significance of large number of tokamaks [2] JT-60 research : High bootstrap full CD, ITB for steady-state tokamak [3] Continuation of JT-60 program : important to keep Japanese potential [4] JT-60 Modification : advance SS Tokamak further ( ITER, SSTR)

European Union United States JET(1983-): 6MA/3.6T Japan DIII-D (1978-) : 3MA/2.2T JT-60 (1985-) : 5MA/4T JFT-2M (1983-) : 0.5MA/2.2T NSTX (1999-) : 1MA/0.3T Tore Supra(1988-):2MA/4.5T TRIAM-1M (1986-) : 0.5MA/8T TST-2 : 0.2MA/0.4T C-MOD (1990-) : 2.5MA/9T TEXTOR(1982-):0.8MA/3T HYBTOK-I : 20KA/0.5T CSTN : 20KA/0.12T 0 1 2 3 4 5 ASDEX-U (1990-):1.6MA/3.9T 0 1 2 3 4 5 Major Raduis (m) Major Raduis (m) MAST (1999-) : 2MA/0.5T

TCV (1992-) : 1.2MA/1.5T

FTU (1989-) : 1.2MA/7.5T China RTP : 0.15MA/2.5T HT-7U:1MA/3.5T(>2005-)

0 1 2 3 4 5 HL-2A : 1.5MA/3T(>2003-) Major Raduis (m) Korea HT-7 (1995-) : 0.5MA/3T Russian Federation KSTAR: 2MA/3.5T(2004-) HL-1M (1984-) : 0.35MA/2.5T T-10 (1975-) : 0.65MA/ 5T

0 1 2 3 4 5 TUMAN-3M (1990-) :0.2MA/1T 0 1 2 3 4 5 Major Raduis (m) Major Raduis (m) 0 1 2 3 4 5 Major Raduis (m) 5 Scientific Achievements of JT-60

WR : World Record

FY '85 '86 '87 '88 '89 '90 '91 '92 '93 '94 '95 '96 '97 '98 '99 '00 '01

JT-60 (1985) JT-60 (1988) JT-60U (1991) JT-60U (1997) •Significant progresses has been made Outer X point Lower X point Lower X point W-shaped Pumped Divertor Divertor Divertor Divertor on fusion performance and steady Geometry state operation in these 16 years .

•Operation of present JT-60 will be In vessel

completed with 17 years of research, First Wall;TiC First First Wall; First Wall; Coarted Mo Wall:Graphite Graphite Graphite equ. end of 2001. QDT High bp H-mode 1.25(WR) Fusion 1.05 Energy Reversed shear Gain 0.6

Ti(0) 45keV (WR) Major parameters of JT-60 Ion 12keV Temp. High bp H-mode Plasma major radius 3.4m n t T (0) Plasma minor radiu 0.9 m D E i 20 -3 Toroidal field coil 1.4x10 keV.s.m RF launcher Fusion 1.5x1021keV.s.m-3(WR) Plasma current 5MA P-NB injector Triple High bp H-mode Toroidal field 4 T Poloidal field coil Product Discharge duration 15s Ip(CD) 3.6 MA(WR) LHRF 1.7 MA 2 MA

1 MA P-NB// injector Ip(CD) NBI

h CD 3.5x1019A/W/m2(WR) N-NB injector Plasma LHRF LHRF 19 2 hCD 1.55x10 A/W/m (WR) NBI NBI Bootstrap Vacuum vessel Vacuum pumping system current 80%(WR) 70% 80%(WR) fraction under Full CD Full CD 2x1020keV.s.m-3 (WR) 6 nDt ETi(0) JT-60 Infrastracture is valuable Properties for Fusion

Motor-generators 2ndary cooling system Central sub-station(275kV, 183MW) H-MG: Evaporation: Power from central substation (MW) 400MVA, 2.6GJ 200 100ton/hr 150 P-MG: 40MW 500MVA, 1.3GJ 100 56.6MW 183 MW T-MG: 50 215MVA, 4GJ 0 0 15 time (min) 30 Transformer yard

Secondary cooling bldg. Central substation Motor-generator bldg. Transformer yard P-P/S: 1GW Power supply bldg. T-Diode P/S : 340MW Primary cooling bldg. JT-60 experimental bldg.

High pressure gas Heating power supply bldg. and machine bldg. Diagnostics system Ion Source Ion Source Tank Central control room (14MeV, activation, Neutralizing Cell spectrometer) JT-60 Heterodyne reflectometer ECE(fourier, poly., heterodyne) High Voltage Table Ion Source Hard X-ray, Soft X-ray, Bol. VUV, Grazing incidence mono. JT-60 N-NBI system(0.5MeV, 10MW) RF heating system Ruby laser TS X-ray TV, Zeff(V.B.) ECRF LHRF ICRF YAG laser TS CO2 collective scattering 120GHz 2GHz 112MHz FIR/CO2 interferometer Magnetic Probes(Eq. Bdot) 4MW 16MW 8 MW CXRS(Ti,Vt,Vp) Divertor MSE(Bp,q,J) Langumuir probe:fixed, recipro CX-NPA (MeV) Interferometer, Vis. spectro. g-ray PHA VUV spectrometer, Da /Ha , IRTC(ripple loss, divertor) Impurity, neutral pressure Thermo couple 7 Proposed JT-60 Machine Modifications

Major Modifications [1] All superconducting magnets for long pulse. [2] 4MA plasma current for Q~1 plasma. [3] Improved shaping, ECCD and RWM coils for AT operation. [4] N-NBCD+ECCD for reactor relevant current drive. [5] use of RAF(in-vessel components) for DEMO. Poloidal Field Coil (SC)

NBI Cryostat

Parameter JT-60SC ITER-FEAT JT-60U Pulse Steady-State Toroiddal Field Coil (SC)

Pulse length 15 s 100 s 400 s Steady Maximum input 40 MW (10 s) 40 MW (10 s) 73 MW 73 MW

5 6

3 2 V 7

H2 8 power ³ 10MW (100 s) 9

10

F11 V1 F12 F12

F13

F14

15

1 16 8 Plasma current Ip 3-5 MA 4 MA 15 MA 7.8 MA 17 18 6

F19 3 1 D2 Toroidal field Bt 4 T 3.8 T (Rp=2.8 m) 5.3 T 4.98 T Major radius Rp 3.4 m 2.8 -3 m (2.8 m*) 6.2 m 6.6 m Minor radius ap 0.9 m 0.7-0.9 m (0.85*) 2.0 m 1.6 m £ Elongation k95 1.8 (d95=0.06) 1.9 (1.7*) 1.7 2.0 8-M80 £ Triangularity d95 0.4 (k95=1.33) 0.45 (0.35*) 0.35 0.35 * Nominal NBI 8 JT-60U JT-60SC New Research Area of JT-60 Modification

[1] Long sustainment of Q~1 plasma (100s, t dis.> t skin) . (r p* and n*close to ITER ss operation) [2] Long sustainment of Full CD plasma (100s). (RS, weak positive shear, N-NBI) [3] Long sustainment of high beta (bN~3-4.2) plasma. (shaping (high k & d and local pitch-dn/dr), ECCD for NTM, Mode-control-coil for RWM) [4] Divertor optimization (compatibility with high d and b, metallic PFC, forced cooling divertor, long particle exhaust) [5] Reduced Activation Ferritic steel for PFC components

9 Significance of Long Pulse Experiments

[1] Long sustainment is key mission of ITER and it will be realized in ITER. [2] Extension of long sustainment of high fusion performance (n T~1020-1021keVsm-3) will certainly contribute to optimize long pulse operation of ITER. [3] Furthermore, long sustainment of non-inductively driven (high bootstrap) discharges will have a great impact to the realization of steady-state tokamak reactor.

2 ITER DEMO(SSTR) 10 Ignition JT-60 Ip limited Break-Even PCD limited 1 0.80MA RS

JET LHD Tore-Supra -2 10 H-E W7AS ATF CHS ATF Triam-1M 10-4 ATF min hr d mon yr 10-6 8 10-2 0.1 1 10 102 104 105 106 10 t Pulse (s) 10 New research area #1 Q~1 sustainment much longer than t skin

[1] JT-60SC : t dis. > t skin of Q~1 plasma

( JT-60U, JET: limited pulse length , t dis.< t skin ) [2] r p* and n* : close to ITER steady-state operation

JT-60U JT-60 modification 4MA 3MA t skin~40s

0 50 time (s) 100 100

0.2n TRIAM-1MU(R=0.8 m) JT-60U skin GW t skin 0.1 0.7n GW • ITER-FEAT(Q=5, 10) 10 n=1x1020 m-3 • JT-60SC JT-60SC (15-40 MW, 4 MA/3.8 T) KSTAR,R=1.8 m 0.4nGW • KSTAR 1 ASDEX-U JT-60U : 3.0m 0.85n (15-40 MW,2 MA/3.5 T) E GW JT-60SC : 2.8m Steady state • TRIAM-1MU JT-60SC,R=2.8 m 0.1 DIII-D : 1.67m ITER-FEAT (1-2 MW, 0.5 MA/8 T) JFT-2M 0.01 ASDEX-U : 1.65m HH=1 (HH=1.5 only Inductive JFT-2M : 1.3m R=6.2 m for steady state ITER) 0.01 1 10 100 1000 0.01 0.1 1 10 0.1 n* 11 dis (s) Steady-state Research

Historical remarks on steady-state researches 1 ) p

1971: Theory of bootstrap current /I (Nature Phys. Sci., R.J. Bickerton) 0.8

1988: Observation of bootstrap current bootstrap SSTR , ARIES-I (TFTR:M. Zarnstorff) 0.6 1990: 80% bootstrap current (JT-60) 1990: SSTR : M. Kikuchi 0.4 JT-60 data ARIES-I :R. Conn 0.2 Measurement 1992: Theory of reversed shear operation Theory

for SSTR, T. Ozeki Bootstrap Fraction (I 0 1993: "Prospect of a Stationary Tokamak Reactor", 0 0.5 1 1.5 2 2.5 3 3.5 *dia EPS invited paper, PPCF, M. Kikuchi p 1995: "Experimental evidence for the bootstrap current", PPCF review paper, M. Kikuchi, M. Azumi

1992-2000; For all IAEA fusion energy conferences, JT-60 team reported progresses towards steady-state operation of tokamak.

12 High Performance Full CD Regime in JT-60U Both weak shear and negative shear regimes are proposed for SSTR. JT-60U results are quite promising for both scenarios. 1013 Negative shear for SSTR: JT-60 Break-even Weak positive shear: T. Ozeki, IAEACN-56/D4-1(1992) Reversed Shear (original SSTR) Q equ=1.25 DT JET JT-60 M. Kikuchi, N.F. 30(1990)265. Negative shear with 80% bs: DIII-D(US) High bp RS Q equ~0.5, 0.8s Weak shear with 51% bs: T. Fujita, IAEACN-77/EX4-1(2000) DT Ti(0)=45keV T. Oikawa, IAEACN-77/EX8/3(2000) (DT) 8 E35037, 8.5 s 1 JT-60 TFTR 6 E36715, 6.4 s 1.2 7 1.6 High bp Full CD (DT) 5 q RS Full CD q 4 NNB 0.8 6 meas. 1.2 q q jtot 3 ECCD 5 0.8 1012 Bootstrap 2 0.4 4 cal 0.4 jBD 1 3 0 0 PNB 0 0 r/a 1 8 9 0 r/a 1 10 Ti (0) (deg.) 10 Ip Full non-inductive CD Ip Full non-inductive CD STEADY STATE OPERATION OH 2 Heating E36715 I q 1 beam Reversed 1.5 Ibeam(NNB) shear q ECCD Profile r/a I (PNB) Ibs optimization 1 beam q r/a 0 Current 0.5 Ibs 4 6 8 10 12 Weak profile shear r/a control q time (s) Heating 00 1 2 3 4 5 6 7 8 Current profile r/a relaxation t ime (s) q Positive Heating 0.8MA/ 3.4T, q95=9.3, dx~0.43, off-axis NBCD INDUCTIVE shear OPERATION 1.5MA/ 3.7T, q95=4.8, NNB (360keV,~4MW), ECH (25%), BS~80-88%, H89P3.3-3.8, HHy22.1- r/a 19 -2 0 1 2 (~1.6MW),INNB~0.61MA(hCD=1.5x10 m A/W), 2.3, bp=2.6-3.2, bN =1.9-2.2- for 2.6s p IPNB~0.26MA,IBS~0.76MA, H89P~2.9, HHy2~1.4, bp=1.9, bN =2.5 for 1.3s 13 Heating and CD System

Current Drive with N-NB (& ECRF) is a unique feature of JT-60 which is most reactor-relevant from engineering point of view.

Present H&CD system becomes powerful for long pulse experiments through minor modifications.

NBI 3 NBI current drive 10 sec 30 sec 100sec Tang. NBI: N-NBI Perpendicular NBI (85keV) 20MW 13.3MW 6.7MW off-axis J(r) 500 keV, 10 MW 2 Tangential NBI (85keV) 10MW 6.7MW 3.3MW N-NBI Tang. Tan. P-NBI center J(r) Negative Ion Tang. NBI (500keV) 10MW 7MW 3MW 1 85 keV, 10 MW ECH (110GHz) 4MW 3.1MW 1.7MW Perp. NBI P(r) 0 Total 44MW 30MW 14.7MW 0 0.5 1 r/a Tangential NBI Perpendicular NBI EC current drive

Plasma Negative Ion Tang. NBI Movable Tangential NBI ECH antenna Perpendicular NBI 14 r/a New research area #2 Long Sustainment of Full CD plasma(>100s)

Full non-inductive sustainment of high-bootstrap plasma sufficiently exceeding skin. Physics of non-linear loop (Jbs-P-Er') in high-bootstrap plasma. 2 Ip=1.5MA, Ibs/Ip~70%, Ibd/Ip~30%,HHy2~1.6, bN~2, q95~9, Bt=3.8T JT-60SC Ip(MA) Ip q q Feedback profile optimization JT-60U r/a r/a 0 ~100 s

spatial gradient of plasma external (variation of pressure Ip=0.80 MA, PNBI=5 MW, bN~2 (~3 s) : Reversed shear N-NBCD: ~3MW control system gradient) nonlinear loop N-NBCD: ~3MW • heating • current P-NBCD:~3.3MWP-NBCD:~3.3MW • momentum P-NBH :~6.7MW Full non-inductive CD P-NBH :~6.7MW Ip ECCD : ~1.7MW ECCD : ~1.7MW self-generation I Total : ~15MW of current self-generation 1 beam Total : ~15MW of electric filed Dt t diff Ibs "flow" structure * plasma rotation 0 structure 4 6 8 10 12 new equilibrium field TIME (s) Self-Organization relaxation and cascade phenomena in wave number space Full non-inductive CD with a reversed Ideal and non-ideal Electrostatic and macro-scale electromagnetic shear plasma in JT-60U (~2.7 s) MHD fluctuation micro-scale fluctuation N *

/a Self-organiztion 15 of novel structure New research area #2 High performance 3 MA Full CD Time Dependent TOPICS with Multi-Beam 1D Fokker-Planck NBCD code 3 Ip 1013 2 IBS JT-60 Break-even I Reversed Shear OH Q equ=1.25 1 DT JET (MA) INB JT-60 DIII-D(US) High bp

0 RS Q equ~0.5, 0.8s 20 DT Ti(0)=45keV (DT) TFTR 10 Ws (DT)

(MJ) JT-60 1012 High bp Full CD 0 RS Full CD 20 P-NB-perp. 108 109 10 P-NB-para. Ti (0) (deg.) (MW) 0 N-NB Simulation result: 2 Ip=3MA, Rp=2.85m,ap=0.85m, 20 3 t E=0.54s, bN=2.6, nd(0)t ETi(0)=5x10 keVs/m , 1 HHy2 Ti(0)=19.3keV, fbs=0.6, fbd=0.42,HHy2=1.4 0 0 10 20 30 40 50 Time (sec) 16 High Pressure towards Compact Reactor needs high bN and Bt

The economically viable fusion reactor requires average plasma pressure 2

of ~1MPa(10atm.) to produce high density (Pf~

) . High bN and high Bt are required to reach such high pressure plasma.

I B p t ×

(MPa )= 0.004 N (MA T/m) ap

1 6 SSTR CREST =3.5 5 Steady-state High- H-mode N ITER JT-60SC p reactor High- mode =2 p N 4 0.1 SSTR High N JT-60U H-mode 3 Steady-state ITER

2 Inductive 0.01

1 Existing tokamak operation region I B

(MPa )= 0.004 p t(MA ×T/m) N a 0 p 0 2 4 6 8 10 12 0.001 0.4 1 10 80 I B /a (MAT/m) p t p 17 New research area #3 Long sustainment of high bN plasma in JT-60SC Effective shaping, profile control, and feedback MHD stabilization are required for long sustainment of the high beta plasma. 6 IDEAL MHD LIMIT WITH WALL JT-60U JET DIII-D CREST( N=5.5) 1MA/1.4T q95=5.1 DIII-D ASDEX-U 5 • Region above Ideal MHD limit Sustainment of b limited by N 1.5MA/2.1T • High beta only achieved transiently heating time, etc n /n =0.8 4 2MA/2.8T e GW • Onset of RWM HH=1.5 SSTR ( =3.5) 3MA/3.5T q95=3.9 N

3 IDEAL MHD LIMIT W/0 WALL ITER SS ( N=2.6-3.5) JT-60U q =3.2 2 4MA/3.8T 95 ITER Ind. ( N=1.8) ASDEX-U JET Total heating DIII-D 1 50 • Region below Ideal MHD limit 44 MW power (MW) 30MW 15 MW - for positive or weak negative shear 0 0 • Limited by heatng time, etc 0 20 40 60 80 100 • Limited by onset of NTM time (s) - for negative shear : JT-60U • Onset of ideal or resistive modes 3 25 N d-coil time limit~4.5 s (@ max. current) due to current profile change E32218 PNB 0 0 ( 1.1 MA/2.1T, d =0.4) 6 time (s) 11 95

E30006 2 N d-coil current limit: low-d @ high I 25 p PNB 0 0 (1.5MA/3.6T, d =0.25) 5 s, lower b 4 time (s) 10 95 N 18 Machine Optimization for High b study in JT-60SC

A flux surface for a SOL width of 3 cm at the midplain to be closed for heat and particle control at divertor CL Vertical position control coil (Copper) An equilibrium with Horizontal position control coil (Copper) 4 MA and bN=3 Ferritic steel plate for TF ripple reduction (1% to 0.4%)

RWM Control coil for n=1,2 modes 10kA

5kA 10kA -20kA -5kA 10kA

-10kA 5kA Horizontal port 20kA -10kA -5kA

10kA

-10kA -5kA 5kA

20kA 10kA -10kA

-10kA -5kA 5kA

Divertor material sample installation device Conductive baffle plates for vertical stability and ideal MHD stability(Ex. RAF) High b compatible divertor Vertical target divertor Cryo-pump for divertor pumping 19 New research area #4 Divertor concept with a wide SOL at high beta • Based on JT-60U results, a wide SOL width at midplane is taken into account for heat particle control at divertor: ~1 cm for heat flux, ~3-4 cm for particle flux Divertor Concept - High and consistent divertor - Vertical target with forced water cooling 2 Tsurf. ~ 1000 deg C at10MW/m - Metal target (or Metal coated CFC) - Strong pumping (two cryo pump) k95=1.83 5.0x1022 /s (~6 times) (100 m3/s at 1Pa) d95=0.36 Characteristics of JT-60SC divertor bp =1.28 - Almost 100% recycling - Characteristic times ; Saturation of wall absorption >10 s Satulation of divertor plate temp. ~10 s

2000 Indirect cooling for JT-60U 1500 Direct cooling for JT-60SC 1000

500 10 MW/m2 0 0 20 40 60 80 100 Time (s) Strong pumping separately from inner and outer 20 Study of Type II (Grassy) ELM with high

• In JT-60U, Grassy ELMy H-mode with

Full CD and HHy2~1.2 is achieved. • The parameter region of the appearance of Grassy ELM is clarified.

• > ~ 0.4 & q95 > ~ 5, bp> ~ 1.6

0.6 Grassy 95~0.36

x 0.5 ( x~0.52) 0.4 0.3 Giant 0.2

triangularity 0.1 0 3 4 5 6 7 q95 • In JT-60SC, it is necessary to find a

way to lower q95 with ~0.35 to have Grassy ELM. 21 New research area #5 Technological Research Subjects

• First wall technology - Development of metal divertor material and first wall in terms of high heat and particle flux from the plasma, erosion, redeposition, dust, high-Z material - Divertor material sample installation device • High heat and particle flux (10 MW/m2, 1022~23 m-2s-1), wall material test • Tritium retention • Reduced Activation Ferritic test (as of JFT-2M AMTEX program) - Elucidate the issue on plasma in the use of ferritic steel for in-vessel component - Clarify effects of strong magnetized material on the plasma behaviors such as plasma build-up, mode locking, positional stability - Application to the toroidal field ripple reduction (1% to 0.4% at the plasma edge)

22 JT-60 modification incorporates many advanced tokamak elements for their integration.

DT burning Ultra-long pulse(TRIAM-1M) (JET,TFTR)

Q~1, Low r *,n* High performance Full CD (JET,JT-60U) (JT-60U)

JT-60 modification High bootstrap (JT-60U) High beta (DIII-D)

N-NBI (JT-60U) Strong Shaping (DIII-D)

ITB (JT-60U) RWM stabilization (DIII-D) Long pulse/forced cooling High Field Pellet (AUG) (Tore Supra)

Advanced divertor (AUG) RAF plasma test (JFT-2M) TM stabilization (JFT-2M) Ergodic divertor(Tore Supra) NTM stabilization (AUG) High field (C-Mod,FTU,TRIAM-1M) 23 Summary

- JT-60 will become powerful advanced tokamak with proposed modification.

- JT-60 will become an integrated test bed of advanced tokamak for ITER and DEMO.

[1] Long sustainment of Q~1, low r *,n* plasma [2] Long sustainment of high performance full CD plasma (N-NBCD+1st harmonic ECCD for current drive) [3] Long sustainment of high beta (strong shaping, stabilizing NTM &RWM) [4] High d and b compatible divertor [5] Technology test such as RAF plasma test , forced cooled divertor target 24