Transcript Summary: Confinement, PWI, divertors, SOL, and Innovative

20th IAEA Fusion Energy Conference,
Vilamoura, Portugal, 1-6 November 2004
Summary:
Confinement, Plasma-wall
Interaction, and Innovative
Confinement Concepts
Hiro. Ninomiya
JAERI, Japan
Statistics of EX and IC
EX (Magnetic Confinement Experiments) 178
EX-C (Confinement)
~93
EX-D (Plasma-wall Interaction)
22
IC (Innovative Confinement Concept)
22
OV: 28, TH: 92, IT: 28, IF: 19, FT: 69, SE: 5
Total: 441
Outline
1. Tokamak Regimes Extended towards ITER
2. Scenario Optimization
3. Global Confinement Physics
4. Transport Physics
5. Plasma-wall Interaction
6. Innovative Confinement Concepts
1. Tokamak Regimes Extended
towards ITER
Long Pulse Operation
1.1 Long Pulse Operation: high b & G sustained>> tR
High b & AT (self regulating) regime > tR
Particle control > tw
0.1
tE
tp*
tR
tW
1
time scale (s)
*JT-60U: extended high-b duration =13tR
sustained bN
4
G=0.75
G=H89PbN/q952=0.5-0.4, q95~3.4
3
2
Inductive Op.
1
0
ITER
Advanced Op.
Weak Shear
0
5
bN=2.5x16.5s
10
15
bN= 2.3x22.3s
20
sustained duration (s)
*DIII-D: 9.5s ITER baseline scenario
~ 9tR, <b>=4%, G~0.55
*JET: 20s reversed shear
25
30
10
1.2 Long Pulse Operation: Excellent Heat Removal
TRIAM-1M
10000
Plasma duration (s)
JET:
20s RS, 326MJ
JT-60U:
30s ELMy-H, 350MJ
LHD:
2min, 115MJ
HT-7:
4min, Tlimiter still rising
TORE-SUPRA:
6min, 1GJ
TRIAM-1M:
5 hrs, No wall saturation
100 GJ
10 GJ
100 MJ
1GJ
ITER
1000
HT−７
TORE SUPRA
LHD
100
JT-60
JET
10
10 kW
100 kW
1MW
10 MW
Injected Power
100 MW
2. Scenario Optimization & Extrapolation
ITER Baseline Scenario
Long Sustainment: DIIID
Integrated exhaust scenario (Ar + pellet) : AUG, (Ar or N):JET
Steady-state / Hybrid Scenarios
Full CD approaches : JT-60U, DIII-D, JET
WS Long Sustainment: NTM-stabilization: JT-60U, DIII-D, JET,
AUG
High Integrated Performance: JT-60U, JET, DIII-D, AUG
High Density & High Radiation: DIII-D, JET, JT-60U
Extension of Improved Regimes
H-mode with small / no ELMs
Core Improvement eITB without central heating
etc.
2.1 ITER Baseline Operation
Increased confidence in reaching the ITER performance
Integrated Exhaust Scenario
DIII-D: Long sustainment
G~0.55x 9tR
AUG:divertor temperature control by
Ar + ELM control by pellet
AUG
W conc.
divertor temperature
divertor density
JET:impurity seeding (Ar or N)
JET
2.2 Steady-state / Hybrid Scenarios:
Full Non-inductive approaches successful
JT-60U (bootstrap+NBCD)
fCD>90%
WS: fBS~45%, 2.8 tR
DIII-D
(bootstrap+NBCD+ECCD)
fCD ~100%, bN<3.5,~1tR
q(r)> ~1.5, q=2 at small  P
RS: fBS ~75%, 2.8tR
100
2000
2004 NEW
f >90%
1996
80
NI
f
BS
(%)
SSTR
60
ITER
Steady
State
f >90%
NI
40
1998
20
0
2002
1998
0
2
4
6
Duration (sec)
8
High b H
p
RS
Double symbol:
10 ~Full CD
High BS Full CD without
inductive current control
2.2 Steady-state / Hybrid Scenarios:
Improved Integrated Performance & ITER access
b
JT-60U
N
r*~0.006
n*~0.06
JET
ITER_SS(I)
E44104_8.3s
HH y2
2.56
1.3
fBS
2.4
1
0.5
0.5
0.53
q95~4.5
fCD 1
1
0.54
0.7
0.77
DIII-D
0.83 ne/nGW
fuel purity
0.56
Prad/Pheat
AUG
good probability
for achieving high
fusion gain in ITER
at reduced current
(~13MA) with a
pulse length
longer than 2000s.
2.2 Steady-state / Hybrid Scenarios： Extended
to High Density & High Radiation
DIII-D
JET
LHCD+Pellet +NBI
=ITB, Ti~Te, ne0nG,
low Rotation
JT-60: ne/nGW>1,
ne(0)/nGW ~1.5
Ne , Ar, D-pellet
2
RS
HHy2
1.5
1
frad
0.5
q95=3.2
1
0.8
0.6
q95=4.5
0.4
0.6
0.7
0.8
0.9
1
ne/nGW
1.1
1.2
2.3 Extension of Improved Regimes
H-mode Improvements
Small - no ELM: AUG, C-Mod, DIII-D,JET, JFT-2M,JT-60U
Low-A MAST: high beta DB, CNTR-NB
NSTX: parametric dependence
of confinement established
Helical: CHS, Heliotron-J, Tohoku-Heliac
Core Improvement
Electron ITB without central fueling:
TCV, TJ-II
ITB with rotation: MAST
Pellet Enhanced Performance : FTU
2.3 Extension of Improved Regimes(2)
Mirror
HANBIT: A stable high density mode found at w<Wci.
GOL-3: Complete multimirror :Te~Ti~2keV at 1021/m3
GAMMA-10: ion-confining potentioal up to 2.1kV
GOL-3
HANBIT
GAMMA-10
neTe+niTi, 1021keV/m3
1.6
1.4
2004
1.2
full corrugation, 1.×
51021m-3, D
improved heating,PL5871,2.08m
1
2002
0.8
full corrugation, 0.×
81021m-3, D
PL5221,3.57m
0.6
2001
4 m corrugat ed ends, 0.3
×
1021 m-3 , D
PL4710,2.08m
dia_records
0.4
1997
0.2
21m-3 , H
unif orm f ield,×
100.9
PL2285,3.73m
0
0
0.2
0.4
time, ms
0.6
0.8
3. Global Confinement Physics
3.1 Scaling Studies of Global Confinement
•JET and DIII-D: b scan with fixed
r* and n* in ELMy H-mode show b
independent (electrostatic)
energy transport
International stellarator database
has been extended and new gyroBohm scaling has been extracted.
•Would predict improved
confinement for high b operation.
LHD
ATF/Hel.E/CHS
W7-AS
TJ-II
Heliotron J
W7-A
tEexp (s)
10-1
10-2
10-3
10-3
10-2
10-1
frentEISS04v3 (s)
t EISS 04n 3  0.148a2.33R0.64 P0.61ne0.55B0.852/0.41
3
0.90
0.14
0.01 0.04
 t Bohmr *
b nb * a
3.2 L/H transition and its power threshold
Biased H-mode in
TCABR (R=0.615m,
r=0.18m) , ISTTOK
and TU-Heliac
(R=0.48m, r=0.07m).
MAST: factor 2
reduction of PL/H in
connected DN.
L-mode
-1
I E (A)
C-MOD: distance
between primary and
secondary separatrix
has large influence to
toroidal rotation and
L/H power threshold
PL/H (low at LSN).
-1.5
B
A
-2
-2.5
-120
H-mode
-110
-100
-90
VE (V)
TU-Heliac
NSTX: HFS gas
puffing reduces PL/H
(less momentum
drag of HFS neutral).
Heliotron J:
H-mode with edge
iota windows.
3.3 ITB
Electron ITB (eITB)
MAST: ITB with steep Te-gradient and peaked
ne profile was formed with counter-NBI
where Mf ~ 1 in core.
NSTX: eITB (+ion ITB) formed with early NBI
and fast Ip ramp (negative shear).
FTU: high density eITB. Te0 up to 5keV at
ne0>11020m-3 with LHCD+ECRH
TCV: Control of eITB with inductive CD
(negligible power variation).
TJ-II: eITB was formed at low order rational
surfaces (r<0.3) with strong positive Er
by loss of ECH superthermal electrons.
JET: ion ITB with small momentum input
and ExB shear.
ITB w. no/small momentum input
MAST
4. Transport Physics
4. Transport Physics
Highlighted topics
Topics
No.
1
2
3
4
5
Zonal flow
Reynolds stress, GAM, Zonal flow
Electron transport
Critical Te, non-linear ce~ (Te )bTea
Device/paper No.
HT-7, Extrap-T2R
JFT-2M, CHS, T-10
AUG, JET, JT-60, DIII-D, LHD. TCV
Particle transport
Tore-Supra, FTU, AUG, JET,
LHD, MAST, ET
Momentum transport
Tore-Supra, C-Mod, FTU, DIII-D,
TEXTOR
G ~ -D[cqq/q- cTTe/Te], ne* dep.
Rotation without torque
Radial electric field
Er control, Flow damping
LHD, GAMMA-10, TJ-II, HSX
ISTTOK
4.1 Zonal flow: measurement of Reynolds stress
Direct measurements of Reynolds stress reported from tokamak and RFP
Electrostatic
Reynolds stress
Zonal flow
3
Edge
Electromagnetic
Reynolds stress
GAM term
SOL
ion
1
8
2
( 10 m/s )
2
0
-1
- mVq
-2
elec dRs/dr
total dRs/dr
-3
-2
-1
0
1
2
Dr ( cm )
HT-7 (Tokamak)
3
4
Extrap-T2R (RFP)
4.1 Measurement of GAM and Low Frequency Zonal Flow
f > 30 kHz
(envelope
f > 80 kHz
(envelope)
Identification of low
frequency Zonal
flow (CHS)
Measurement of
GAM (T-10)
Fourier amplitude [a.u.]
The modulation of ne,ambient
correlates with GAM (JFT-2M).
2.0
1.5
1.0
0.5
0.0
HIBP Ion beam current
r = 0.93, Dr = 0.07
CHS
HIBP plasma
potential
0.4
0.2
0.0
0
10
20
30
Frequency [kHz]
40
50
Twin HIBP
)
~
f and
envelope
Zonal flow profile
Zonal flow (f < 1kHz)
4.2 Electron transport:
Critical Te, non-linear ce~ (Te )bTea
• Critical Te
JET, JT-60U => YES, DIII-D => NO
• Non-linearity
JET, JT-60U => YES, DIII-D => NO
JET
JT-60U
LHD => NO
LHD => YES but on Te
LHD
cetr = C Tea(Te )b
Strong Te : a~1-2.5
Weak Te : b~0
ctr~cpb
Exp. of effect of plasma shape and shear (TCV)
4.3 Burning Plasma Physics
JET: Thermal Tritium transport
•Turbulence dominates thermal
particle transport for most regimes
–Large inward vT correlates with
high DT
–Neo-classical only for : high ne
ELMy H & in ITBs.
•Dimensionless parameters scans
show:
–Gyro-Bohm particle transport (DT~ r *3)
for Inner plasma;
–Bohm particle transport (DT~ r *2) for
Outer plasma;
–when q scans are included scaling is
more like Gyro-Bohm in outer plasma
(DT~ r POL*3 ; r POL*=q x r * );
–particle transport has an inverse b and
n * dependence.
ne (1019 m-3)
Non-ITB dataset DT/BT vs density
4.3 Particle transport: dependent on 1/LT,1/Lq,ne*
• Evident turbulent pinch observed in Tore Supra and FTU.
Both the thermodiffusion (Te/Te) and curvature (q/q) pinches co-exist.
• Density peaking increases with decreasing collisionality, consistent with
quasi-linear ITG/TEM model (AUG, JET)
 could lead to
higher fusion
power in ITER
Confirmation of
extrapolation to ITER
requires further
experiments.
Concern for mpurity
accumulation
(JT-60U, JET and
AUG)
4.4 Momentum transport : Rotation without torque
• Rotation without torque is important for transport and
stability (RWM).
 More reports of rotation without torque input (C-mod, DIII-D,
TEXTOR, Tore Supra)
C-Mod: rotation changes
with USN,LSN (ICRF)
DIII-D: CTR rotation with ECH
TEXTOR: control by 3/1 DED
Tore Supra
Co-rotation ~ 80km/s
Cf. AUG; -400km/s for QH
mode with counter NBI
4.5 Radial electric field
Er control, flow damping
Combination of magnetic geometry with Er produce interesting
phenomena (Gamma-X, LHD, TJ-II, HSX, ISTTOK)
Strong
Weak
Er Shear
2
2
1
1
0
2
1
0
2
0
4
0
4
2
2
0
10
5
0
10
5
0
-5
0
0
HSX
Er Shear
Viscous flow damping
Drift Wav e
ISTTOK:bias
Turbule nce
1
5
10
15
-5
Vorticity W r
=[ nVE]z /n0
TJ-II
Turbulence suppression
Er She ar
dEr /dr
0
5
10
Radial Electric field
15
20
Radius rc (cm) Radius rc (cm)
GAMMA-10
1.4x1019m-3
2.7x1019m-3
0
LHD
Er control
0.5
0.4
0.3
0.2
r
-10
0.6
h
Turbulence suppression
E (kV/m)
10
1.3x1019m-3
Rax=3.9m
0.1
-20
0
0.0 0.2 0.4 0.6 0.8 1.0 1.2
r
5. Plasma-wall Interaction
5.1 Active Control of Edge Plasma
• Higher confinement of tE=1.2 tEISS95 due to sharp edge (large Te
gradient) with a Local Island Divertor (LID) in LHD
• Onset of 2/1 and 3/1 tearing modes by Dynamic Ergodic Divertor
(DED) and reduction of the edge poloidal rotation.
• Configuration effects (USN, DN, LSN) on particle control in DIII-D
TEXTOR
LHD
0.08
4
(s)
LID
config.
t
E
exp
w/o LID
2
e
T (keV)
3
(a)
1
w/ LID
0
3.8
4
4.2
R (m)
4.4
4.6
0
0
Helical divertor
config.
ISS95
0.08
t
(s)
E
5.2 Recycling/Wall retention
22
Particles
x10(22x 10 )
•Wall saturation in JT-60U (30s NB heating, Tvv=150, 300oC)
•No wall saturation in TRIAM (5h 16min, Tvv=30-40oC) and Tore Supra
(6min., TLimiter=120oC)
TRIAM
• Wider retention area than the area
directly interacted with plasma (JT-60U,
TRIAM, Tore Supra, JET, ASDEX-U.
8x1021 H
TEXTOR).
JT-60U
4
1-2x1022 D
3 Injected
Pumped
2 Wall
Retention
1
0
5
10
Tore Supra
8x1022 D
15
20
Time ( sec )
25
30
Phase 1
Phase 2
Tungsten Wall
• 65% of all PFC are W coated in ASDEX.
• High performance discharge with moderate W concentrations
feasible.
• W concentration is controllable with central ele. heating and pellet
triggering of ELMs
• Blisters and bubbles are formed on the surface of W irradiated
with low energy (~100 eV) H beam
W
10
8
6
4
2
0
PNBI (MW)
PICRH (MW)
_
ne (1019m)-3
8
6
4
D (51020 s)-1
Da (a.u.)
2
0
3
Te0(keV)
bN
2
H98y,2
1
0
10-5
10-6
keV)
c (2.5
W
keV)
c (1
W
2.0
2.5
Time (s)
3.0
3.5
Further experiment in large tokamaks with high power heating
Carbon Migration
•C migration toward the inner target and its main origin is
main chamber (DIII-D, JET, AUG, JT-60U)
13CH
4
injection exp.
SEM analysis
Divertor
pumping
10cm
JET
JT-60U
DIII-D
Tritium Retention
T(D)
retention
JET
JET
ASDEX
JT-60
D/C
dust
3% 0.4 - 1.0 1 kg
3% 0.4 - 1.0
<2%
7g
JT-60U
T retention much lower with
vertical target in JET:
Geometry effect?
D/C ratio and dust much lower
in JT-60: better alignment?
Higher temperature?
6. Innovative Confinement
Concepts
6 Innovative Confinement Concept
Experiments:
• SC levitated internal ring in ECH heated plasma on Mini-RT
• Measurement of axial flow shear in the ZaP flow Z-pinch
• CD by Helicity injection in the HIT-II & HIT-SI
• FRC plasmas, produced and sustained by the RMF, and for MTF
(FRX-L, TCS)
• Sequence of spheromak formation (CALTECH), supersonic rotation
with centrifugal confinement (MCX)
Mini-RT
Axial flow shear in ZaP flow Z-pinch
6 Innovative Confinement Concept
Numerical studies:
• Nonlinear evolution of MHD instability
in FRC
• Design of magnetic measurement for
3D equilibrium and model of ambipolar
plasma flow for NCSX
• Simulation of liner compression using
two fluid model
• Optimization of quasi-poloidal stellarator
Rotational mode in FRC
New Concept:
• Burning spherical tokamak by pulsed high-power heating of
magnetic reconnection
• Selective heating using LH for He ash removal
• Solenoid-free start-up for spherical torus using outer poloidal field
coils and conducting center-post
• Spherical tokamak configuration using spherical snow-plug
I am very much pleased that fusion
community has made significant
progress in confinement and plasmawall interaction research areas. These
results will greatly contribute to ITER.