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

1
Title
24th SOFT Conference Sep. 2006, Warsaw, Poland
O1A-A-360
JT-60SA
Prospective performances in JT-60SA towards 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
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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
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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
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Basic Machine Parameters of JT-60SA
high-S for DEMO
ITER similar
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, k95 1.7 / 1. 83
Triangularity, d95 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
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Heating Current Drive Equipement
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
N-NB (500 keV) co (2u) 10 MW
P-NB (85 keV) co (2u) 4 MW
P-NB (85 keV) ctr (2u) 4 MW
P-NB (85 keV) perp (8u) 16 MW
EC 110 GHz 3 MW
EC 140 GHz 4 MW
total 41 MW
P3-B-336 Y. Ikeda, et al.
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Plasma Performance

Mission and Concept Plasma Performance
Engineering Design Time Schedule Summary

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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
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Feasibility for current drive scenario like an
ITER hybrid operation
ACCOME-code analysis ITER similar
configuration fGW0.85, HHy21.3, q953.1,
Pin41MW
Hybrid operation up to 3.7MA for 100s will be
available.
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High-b full non-inductive current drive scenario
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 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).
fGW
fBS
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Access for breakeven and high-b plasma with ITER
and DEMO relevant parameters
Accessibility for high QDT and high bN is
enhanced with increased heating power.
Non-dimensional parameters with ITER and DEMO
relevant region are expected.
A2.6, DN, q953.5, HH98y21.5
A2.6, k1.8, q955.5, bN4 (2.4MA, fGW0.86)
0.010 0.008 0.006 0.004 0.002 0.000
25MW, HH98y21.5
2.4MA, fGW0.86
3MA, fGW0.56
Normalized Larmor radius ri
41MW, HH98y21.3
ITER (Steady state)
DEMO (J05)
0.00 0.02 0.04 0.06 0.08 0.10 0.12
Normalized collision frequency ne
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Flexibility in aspect ratio and plasma shape for
high-b plasma accessibility
Flexibility in S and A is extended, which
enhances the research capability for high-b
plasma operation.
Shape parameter
Ip
q95 µ A-11?2(12?2)
S
aBT
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Controllability for resistive wall mode (RWM)
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
bN3.8 inside stabiliser plates
bN5.6 Sector coils are located on the port
entrance in the present design (Analysis
ongoing)
Outside
Inside
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Heat particle control with semi-closed divertor
Divertor plasma simulation with SOLDOR/NEUT2D code
Qtotal12 MW, Gion 1 x1022s-1, Gpuff 0.5
x1022s-1 , Spump 50 m3/s, ?e?i1 m2/s, D0.3
m2/s , Cimp1
- ??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.
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Engineering Design
?
Mission and Concept Plasma Performance
Engineering Design Time Schedule Summary
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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 RD of shielding
material Boron doped resin etc. Shielding
analysis 2D/3D code
Divertor Target design Heat removal
Particle pumping Cooling system
Remote Handling System
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Superconducting Coils
CS
TF
EF
conductor
P1-E-328 K. Tsuchiya, et al P1-E-286 K.
Kizu, et al.
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Vacuum vessel
consists of 18 sections
VV has a double-wall structure.
cylindrical toroidally, polygonalpoloidally
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140mm
weight 300 ton without in-vessel components
one turn resistance 15µ?
Shielding water
(Boronic acid Water)
baking temp. 200C (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
Birds-eye view of vacuum vessel
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Plasma facing components
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 RD.
Exchange with full metal plasma facing
components will be decided after experimental and
computational analyses.
Width 10deg, Weight lt500kg
Crank support for allowing large thermal expansion
Heat sink for bolted armor
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.
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Radiation Shield
? DD neutron emission rate
P3-J-302 A. M. Sukegawa, et al.
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Time Schedule
Mission and Concept Plasma Performance
Engineering Design Time Schedule Summary
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Time Schedule
Schedule of construction and operation agreed in
JA-EU WG Construction 7 years exploitation 3
years

1. Completeion of Conceptual Design with the
   collaboration of JA and EU design teams
2. Detailed Design and Starts of Construction

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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.
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Acknowledgement
Thank you for your attention.
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