1 JT-60SA Eurat om 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 Eurat om JT-60SA Eurat om Prospective performances in JT-60SA towards the ITER and DEMO relevant plasmas O1A-A-360 24th SOFT Conference Sep. 2006, Warsaw, Poland
26
Embed
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,
This document is posted to help you gain knowledge. Please leave a comment to let me know what you think about it! Share it to your friends and learn new things together.
Transcript
1
JT-60SA Euratom
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. Takase1), K. Tsuchiya, D. Campbell2), S. Clement3), J. J. Cordier4), J. Pamela5), F. Romanelli6), and C. Sborchia7)
JT-60SA Euratom
JT-60SA Euratom
Prospective performances in JT-60SA towards the ITER and DEMO relevant plasmas
O1A-A-36024th SOFT Conference
Sep. 2006, Warsaw, Poland
2
JT-60SA EuratomOUTLINE
• Mission and Concept
• Plasma Performance
• Engineering Design
• Time Schedule
• Summary
3
JT-60SA Euratom Mission and Concept
• Mission and Concept
• Plasma Performance
• Engineering Design
• Time Schedule
• Summary
4
JT-60SA Euratom
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 EuratomMission 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, q95 =3.0
Support to ITER
divertor structure : TBDhigh-, shape for high-beta operation Time (s)
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
Plasma Current Ip(MA) 3.5 / 5.5
Toroidal Field Bt (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, A 3.10 / 2.64
Shape Parameter, S 4.0 / 6.7
Safety Factor q95 3.0 / 3.77
Flattop Duration 100 s (8 hours)
Heating & CD power 41 MW x 100 s
N-NBI 34 MW
ECRH 7 MW
PFC wall load 10 MW/m2
Neutron (year) 4 x 1021
D2 main plasma + D2 beam injection
Basic Machine Parameters of JT-60SA
ITER similarhigh-S for DEMO
Gravity Support
NBIPort
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
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
EC110 GHz 3 MW
140 GHz 4 MW
total 41 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
Ip
co
Ip
N-NB(co)
EC
Tangential P-NB (ctr)
Perpendicular P-NB
Tangential P-NB (co)
Remote HandlingSystem
Resonance layer of EC with two-frequency system
8
JT-60SA Euratom
Mission and Concept
• Plasma Performance
• Engineering Design
• Time Schedule
• Summary
Plasma Performance
9
JT-60SA Euratom
• 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
Feasibility for current drive scenario like an ITER hybrid operation
Hybrid operation up to 3.7MA for 100s will be available.
Cur
rent
dis
trib
utio
n (M
A)
Plasma current (MA)
4.0
3.0
2.0
1.0
0.03.23.0 3.4 3.6 3.8 4.0 4.2
beam driven
bootstrap
ohmic
70
80
90
100
110
120
130
140
150
0.114
0.116
0.118
0.120
0.122
0.124
0.126
0.128
0.130
3.0 3.2 3.4 3.6 3.8 4.0
Loop
vol
tage
(V
)
Ava
ilabl
e fla
ttop
(s)
Plasma current (MA)
FlattopVl
ACCOME-code analysis ITER similar configuration fGW=0.85, HHy2=1.3, q95=3.1, Pin=41MW
11
JT-60SA Euratom
High- full non-inductive current drive scenario
• 2.4 MA full current drive with A = 2.65, N = 4.4, fGW = 0.86, fBS = 0.70 and HH98y2 = 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).
0
0.2
0.4
0.6
0.8
1
0
0.2
0.4
0.6
0.8
1
2 2.2 2.4 2.6 2.8 3 3.2
f GW
f BS
Ip [MA]
fGW
fBS
-0.20
0.20.40.60.8
1
0 0.2 0.4 0.6 0.8 1
j [M
A/m
2 ]
total
BS
BDEC
OH
02468
10
0
2
4
6
8
n e [10
19 m
-3],
Ti ,
Te [
keV
]
qTe Ti
ne
q
12
JT-60SA Euratom
Access for breakeven and high- plasma with ITER and DEMO relevant parameters
A=2.6, DN, q95~3.5, HH98y2=1.5
2.5
3.0
3.5
4.0
4.5
5.0
5.5
6.0
0.0 0.5 1.0 1.5 2.0
N
QDT
eq
3MA1.5T
3.5MA1.8T
4MA2T
4.5MA2.3T 5MA
2.5T
2.5MA1.25T
25 MWn/nGW=0.8
40 MWn/nGW=0.8
5.5MA2.8T
Accessibility for high QDT and high N is enhanced with increased heating power.
0.00 0.02 0.04 0.06 0.08 0.10 0.12
Normalized collision frequency e*
41MW, HH98y2=1.3
ITER (Steady state)
DEMO (J05)
3MA, fGW=0.56
25MW, HH98y2=1.50.010
0.008
0.006
0.004
0.002
0.000
Nor
mal
ized
Lar
mor
rad
ius
i*
2.4MA, fGW=0.86
Non-dimensional parameters with ITER and DEMO relevant region are expected.
A~2.6, ~1.8, q95~5.5, N~4 (2.4MA, fGW=0.86)
break-even class plasma
TFTR
ITER
JT-60
DIII-D
FTU
LHD
C-Mod JET
Ti(0) (K)
JT-60SA
KSTAR
Self-ignitionCondition
1021
1020
1019EAST
DEMO
Break-evenCondition
108 109107
nD(0
) E
(s
ec/m
3)
collisionless /small normalized Larmor radius
13
JT-60SA Euratom
SIp
aBT
q95 A-11+2(1+22)
4 5 6 7 8
2.5
3.0
3.5
Shape parameter S
Asp
ect
rat
io
A
Divertor pumping(m3/s)≥100 <100
Double nullSingle null
ITER
Flexibility in aspect ratio and plasma shape for high- plasma accessibility
*M. R. Wade, et al., Phys. Plasmas 8 (2001) 2208.
JT-60SA
S=2.0-2.2S=3.1-3.6
JT-60 ASDEX-U JET DIII-D
6
5
4
3
2
Nor
mal
ized
bet
a
N
2 3 4 5 6 7Shape parameter S
DIII-D Experiment *
ITERJT-60
Target of JT-60SAN: 3.5~5.5
S=2-8S=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
Achievable N depends very much on the location of sector coil