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NATIONAL FUSION FACILITYS A N D I E G O
DIII–D QTYUIOP045-99/TST
THE DIII–D ADVANCED TOKAMAK PROGRAM
byT.S. Taylor
Presented atWorkshop on Physics Requirements
for Advanced TokamaksSan Diego, California
March 9–11, 1999
NATIONAL FUSION FACILITYDIII–D 031–99/TCS
OUTLINE
Introduction— What is an advanced tokamak
— DIII–D AT program goals
Steady state and high βN
Principal AT scenarios
DIII–D AT program elements— ITB physics
— Profile control [J(r)]
— Stability limits and pressure profile
— Edge stability
— Neoclassical tearing modes
— Wall stabilization
DIII–D — A NATIONAL FUSION RESEARCH PROGRAM
Collaborations with 50 institutions — 300 users1999 DIII–D National Physics Team (93 FTE)
General Atomics
LLNL
Visitors
ORNL
PPPLOther
Universities
045–99NATIONAL FUSION FACILITYDIII–D
AlaskaAlbertaCal TechChalmers U.Columbia U.Georgia TechHampton U.Helsinki U.Johns Hopkins U.LehighMITMoscow State U.RPIU. MarylandU.TexasU.TorontoU.WalesU.WashingtonU.WisconsinUC BerkeleyUC IrvineUCLAUCSD
ASIPP (China)Cadarache (France)CCFM (Canada)Culham (England)FOM (Netherlands)Frascati (Italy)Ioffe (Russia)IPP (Germany)JAERI (Japan)JET (EC)KAIST (Korea)Keldysh Inst. (Russia)KFA (Germany)Kurchatov (Russia)Lausanne (Switzerland)NIFS (Japan)Troitsk (Russia)SINICA (China)SWIP (China)Southwestern Inst. (China)Tsukuba U. (Japan)
ANLINELLANLLLNLORNLPNLPPPLSNLASNLL
CompXCPI (Varian)GAGycomOrincon
NATIONAL LABS UNIVERSITIES INTERNATIONAL LABS
INDUSTR
NATIONAL FUSION FACILITYDIII–D 031–99/TCS
WHAT IS AN ADVANCED TOKAMAK
"Improvement of the tokamak concept towards higher performance andsteady-state operation through profile modification and control, plasmashape, and MHD stabilization"
Goal: concept improvement— High performance
3333 High β3333 High τE
— Steady state
How, techniques— Profile control
— Shape
— MHD stabilization
500
0
– 500
"SMART": R = 4.8 m,$rel = 0.59, I = 10/7 MA,
Qss = 27
500
0
– 500
– 1000
Z (c
m)
ITER CDA: R = 6.0 m,$rel = 1.0, I = 22/19 MA,
Qss = 4
IMPROVED CONFINEMENT AND STABILITY LIMITSLEAD TO MORE COMPACT POWER PLANT DESIGNS
142–96
β ↑ 50% β ~ 4.5 I/aB
τ ↑ 80% τ ~ 3.6 τITER–89P
(Perkins, LLNL, 1993)
QTYUIOP
ADVANCEDTOKAMAK
CONVENTIONALTOKAMAK
QTYUIOP
ADVANCED TOKAMAK APPROACH RELIESON MODIFICATION OF PROFILES
"Conventional Tokamak" Advanced Physics Approach
Global parameters (0–D) Internal profiles (1–D, 2–D)
Confinement: 1 < H < 2 2 < H < 4Increase Ip JJJJ Shape, δ, κ, ε, . . .JJJJ Larger size, or JJJJ Plasma rotation → sheared (E×B)JJJJ Large B JJJJ Current density
JJJJ Pressure po / < p >, βp JJJJ Density profile control, edge control
Stability: 2 I/aB ≤ β ≤ 3 I/aB 3 I/aB ≤ β ≤ 6 I/aBJJJJ Limited power/A JJJJ ShapeJJJJ Increase Ip, lower q JJJJ Plasma rotation, wall stabilization
JJJJ Current density, lllli, qo, < J(a) > JJJJ Pressure po / < p >, p′(a)
Current Inductive RF "smart" current drive bootstrapDrive: JJJJ Pulsed reactor JJJJ Steady-state
Disruptions: High Ip → low q Lower Ip — higher q → lower disruptivity→ increased risk Disruption avoidance – "maintain" stable profilesof disruptions Disruption control — local heating and
current drive, passive and active mode control
⇒ Advanced Tokamak achieves same performance at lower Ip
045–99
reduced force
NATIONAL FUSION FACILITYDIII–D QTYUIOP
045–99
THE MISSION OF THE DIII–D PROGRAM IS
To establish scientific basis for the optimization of the tokamak approach tofusion energy production
— The DIII–D Program's primary focus is the Advanced Tokamak thatseeks to find the ultimate potential of the tokamak as a magneticconfinement system
Strategy:— Demonstrate improvements
separately, then simultaneously
— Develop solid scientific understandingand predictive capability3333 Diagnostics3333 Theory/modeling3333 Require strong coupling of theory
and experiment
— Develop control scenarios and tools basedon scientific understanding
— Increase performance and duration
Performance
GOALβN H
τDUR/τE
⇒ Steady State
QTYUIOP112–98/TST
THE DIII–D ADVANCED TOKAMAK PROGRAM SEEKS TOEXPLORE THE ULTIMATE POTENTIAL OF THE TOKAMAK
Performance High confinement enhancement factor, H 4 High β, βN 6
Duration, Steady State Non-inductively driven, Eφ(ρ) 0
— High bootstrap fraction
Heat Removal and Particle Control High volume recombination and
radiated power fraction Low core impurity content
SimultaneouslyIntegrated
Approach: Active Control of Profiles Plasma shape Current density profile
⇒ Localized current drive Pressure profile
⇒ Localized heating for ITB⇒ ΩE×B, poloidal rotation with RF
Rotation profile, E×B Density profile Radiation, impurity, profile Recombination, fueling source
016–98 QTYUIOP
ADVANCED TOKAMAK PROGRAMDEMANDS NEW RESEARCH PARADIGM
Details of profiles are important— q(ρ); P′(ρ); Ω(ρ) or ωE×B; P(a), J(a); turbulence, n(ρ), T(ρ)— Geometry is important, plasma shape, divertor shape; interaction with profiles— Inherently 1 and 2 dimensional— Requires new profile diagnostics— Requires new analysis approaches, new analysis tools
Self-consistency and simultaneity increase complexity— Strong, nonlinear interaction of many elements
3333 Current profile → transport → pressure profile → current profile3333 Heating deposition → pressure profile → ωE×B → transport → pressure profile →
current profile3333 Strong coupling between pressure profile, plasma shape, wall stabilization and stability limit3333 Divertor → SOL → pedestal → edge stability and core transport
— Synergism among different physics must be fully explored— Demands more integrated experimental research program— Requires new analysis code capabilities — more “integrated” codes
Fundamental physics theories are required to understand, lead the experiments, andprovide predictive capability— Strong coupling between theory, modeling, and experiment
⇒ Increased challenge ⇒ exciting
QTYUIOPBR FEAC-20
THE CHALLENGE OF SELF-CONSISTENT PROFILES
MHDStability
Shape
ECCD and FWCDCurrent Drive
q ProfilePressureProfile
Bootstrap
Magneticshear
E×B shea
r
BeamPower
TransportRotation, ΩωE×B, γmax
Magnetic
shear
Shafranov
shift
InductiveShaping
045–99
LocalizedHeating ECH
NATIONAL FUSION FACILITYS A N D I E G O
DIII–D
FIGURES OF MERIT
045–99NATIONAL FUSION FACILITY
S A N D I E G O
DIII–D
Confinement enhancement
Normalized beta
Bootstrap fraction
Fusion gain
Steady state Q
Bootstrap alignment (Politzer)
Duration
H = τE/τEITER–89P
βN = (I/aB)BT
fBS ∝ ξ √A q βN
PfusPloss
∝ nT τ ∝ βN HI1 B3 ∝ βN H q2
PfusPCD
∝ γcur εeff βN B
3
n q (1 – ξ √A qβN)
falign = 1 – ∫ dv
∫ dv
neT abs (J – JBS)
neT abs (J)
τDUR/τE ; τDUR/τCR
QTYUIOP112–98 TT/df
A COMPACT STEADY-STATE TOKAMAK REQUIRES OPERATION AT HIGH βN
βN = 3.5
Large Bootstrap Fraction (Steady State)
0.0 0.5 1.0 2.0
10
εβp
β
ε 1 + κ2
Equi
libriu
m L
imit
q* = 4
LargePower
Density
βN = 5
• High power density⇒ high βT
• Steady state ⇒ high IB/Ip
⇒ high βp
• High βT + high βp ⇒ high βN
2 5
01.5
Advanced Tokamak
Curr
ent L
imit
Ideal MHD
Pressure Limit
β β κ βT Np ∝ +
1
2
22
β βN T aB= ( )I
StableUnsta
ble (Exte
rnal Kink)
Unstable (Internal Kink)p
r
0.5 0.7 0.9 1.10.0
2.0
4.0
6.0
βN
li (INTERNAL INDUCTANCE)
NATIONAL FUSION FACILITYS A N D I E G O
DIII–D
MODELING INDICATES PROFILE OPTIMIZATIONAS A WAY TO INCREASE THE BETA LIMIT
Beta limit sensitive to— Current profile— Pressure profile
Inductively driven currentconstrains allowable currentprofiles (and possiblypressure profiles)
Strong heating, fueling,current drive alters theconstraints
Implies two time scales— Pressure profile
relaxation → τE— Current profile
relaxation
045–99
(J.R. Ferron, Phys. Fluids B, 1990)
NATIONAL FUSION FACILITYDIII–D 031–99/TCS
HIGH PERFORMANCE DURATION IS A RELEVANTFIGURE OF MERIT
Profiles impact the stability limit
— Pressure profile
— Current density profile
⇒ Two obvious time scales of interest to develop physics basis
— Pressure profile relaxation, τE3333 Required to demonstraate self-consistency between transport,
pressure profile, and stability limit
— Current profile relaxation, τCR
τCR ~ 1. 4a2 K
Te3 2
Zeff3333 Required to demonstraate self-consistency between pressure profile,
current density profile, and stability limit
NATIONAL FUSION FACILITYS A N D I E G O
DIII–D 045–99
THE DIII–D AT PROGRAM BALANCE — FOCUS
The DIII–D program aims to develop the best possible operational scenario forfusion energy production— Many opportunities to make improvements— Many complex interdependencies ⇒ many possible advanced tokamak solutions
A key to our approach to research.Maintain research attitude that is open to— Evolving and improving knowledge base— Innovations— New discoveries⇒ Device flexibility⇒ Diagnostic capabilities⇒ Control capabilities
BUT,
Focus sufficiently to test specific scenarios— Indicated by the theory— Identified potential for high performance + steady state— Consistent with limited resources
031-99 tcs/rsNATIONAL FUSION FACILITYDIII–D
Plasma Parameters DIII–D Plasma ShapeIp = 1.6 mega-amperesB = 2 Teslaq95 = 4.4Ti(0) = 10 keVTe(0) = 6 keVne = 6.5×1019 m–3
τE = 0.21 seconds
0
4
8 βNH98hy
0
1
2 H98hy
1.0 1.5Conventional Tokamak (ITER–EDA)
2.0 2.5 3.0 3.5Time (s)
0
2
4
Plasma Pressure
Plasma Energy Confinement Improvement
Advanced Performance Product
β (%)96686
96686
AT duration limited by neoclassical tearing mode
βN = βT/(I/AB)
H98y = τE/τEITER98y
Recent Progress
0
2
4
6
8
10
0.1 1 10 100
Perf
orm
ance
, βN
H9 8
hy
Duration, ∆t/τE
1998
1997
Conventional Tokamak(ITER–EDA)
(Max QDD)
≈ H89P /1.7
Advanced Tokamak(ARIES–RS)
DIII–DJT–60UAUG
DIII–D 1998 RESEARCH HAS EXTENDEDTOKAMAK PERFORMANCE AND DURATION
NATIONAL FUSION FACILITYDIII–D 031–99/TCS
WE HAVE IDENTIFIED FOUR ADVANCED TOKAMAK SCENARIOSTHAT HAVE POTENTIAL FOR IMPROVED PERFORMANCE
Negative central magnetic shear
High internal inductance
Radiative improved modes
— Consistent with high lllli
GyroBohm scaling of ELMing H–mode to improved confinement
— ρ* scaling to ignition gives H > 3
— Requires improvement of the β-limit (NTM)
045–99
STEADY STATE CONSIDERATIONS ALSOLEAD TO TWO “NATURAL” CURRENT PROFILES
QTYUIOP
High current drive
J ⇒ high li
⇒ NCS
ρ
High bootstrap
Qss = PfusPCD
∝γcur Pfus
nRICD∝
γcur PfusnRI (1 – JBS)
QTYUIOP045–99
IMPROVED PERFORMANCE CONSIDERATIONSLEAD TO THE IDENTIFICATION OF TWO AT SCENARIOS
BASED ON THE CURRENT PROFILE
γLS HS
LS
HS
unstable
High n ballooning stability diagram
Magnetic Shear
Both low magnetic shear (LS) and high shear (HS) are favorable for:
Growth rate of trapped particle modes
Reduced turbulence andReduced transport Higher Beta
Motivated by theoretical considerations
— High lllli — high magnetic shear (HS) in the outer region of the plasma— Negative central magnetic shear → low or negative shear (LS) in the plasma core
α
Magnetic Shear1
0
HSLS
0
LS
HS
QTYUIOP124–97
lllli
10
2
00 1 3
β N (%
–T
–m/M
A)
Ballooning limit (Calc.)Experiment
4lllli
8
6
4
12
5
1.5 2.0 2.51.0
2
1
3
4
3.0li (target)
βN
BT = 1.5T, Ip = 0.4 – 0.5 MA, PNBabs = 12– 16MW,p (0)/<p> = 1.7 – 2.2
βN
(a)
(c)
JT-60U
DIII–D
βN INCREASES WITH INTERNAL INDUCTANCE,lllli
20.0
0.5
1.0
1.5
2.0
2.5
3.0
0
1
2
3
li (magnetics)
β N (
diam
agne
tic)
1.1 - 1.9 MA2.0 - 2.7 MA
D DT Ip
[Ferron, Phys. Fluids B5, 2534, (1993)]
(b)
[Sabbagh, IAEA, Montreal, F1-CN-64/AP2-17]
[Kamada, Nucl. Fusion 34, 1603, (1994)]
QTYUIOP
ENERGY CONFINEMENT INCREASES WITHPEAKING OF CURRENT DENSITY PROFILE
τ TH/
τ JET
/DIII
–D
2.0
1.5
1.0
0.5
0.00.5 1.0 1.5 2.0 2.5
Internal Inductance (li)
H–modeκ ∼ 1.7
L–mode κ ∼ 1.7
Current RampL–mode, κ ∼ 1.2
κ Ramp
H–mode, κ ∼ 2.1
124–97
τJET/DIII–D = 0.106 PL–0.46 I1.03 R1.48
Confinement improvement with li observed in many experiments
— TFTR [Zarnstorff, Phys. FluidB 3, 2338, (1991)]— JET [Christiansen, Proc. 19th EPS Conf., Innsbruck, Austria, Vol. I, p. 13. (1992)]— Tore Supra [Hoang, ibid, p. 27]— JT–60U [Kamada,Nucl. Fusion 34 1605, (1994)]
TFTR has obtained high fusion power in high li discharges (Sabbagh, IAEA, Montreal, F1-CN-64/AP2-17):
— PFus = 8.7 MW
— lllli = 1.4
— Ip /BT = 2.0 MA, 4.8 T
Ferron, Phys. Fluids B5, 2534 (1993)
145–97 TT QTYUIOP
INCREASING H AND βN WITH lllli SUGGESTAN ATTRACTIVE ADVANCED TOKAMAK SCENARIO
Advantages
Ease of central current drive
High βN, high H observed on manyexperiments
No power threshold Compatibility with ELMing H–mode,
radiative I–mode
Challenge: Self-Consistent High β, High lllli Scenario
0.0 0.4 0.80.0
1.0
2.0
ψ 1/2
<J>
(MA/
m2 )
Total current
Bootstrapcurrent
Limitation
Alignment of bootstrap current:
High lllli
⇒ Low edge J, high Sn⇒ Reduced edge transport⇒ High edge p′⇒ High edge bootstrap
High edge J
qo = 1.05
lllli = 1.2
q95 ~ 8 qo = 0.55
IBS/Ip ~ 50%–60% ⇒ ⇒ ⇒ lllli = 1.4–1.6
βN ~ 4 with sawtooth
H ~ 2–3 stabilization
NATIONAL FUSION FACILITYS A N D I E G O
DIII–D QTYUIOP298–98
NEGATIVE CENTRAL MAGNETIC SHEAR (NCS) SCENARIO IS BEINGPURSUED BY DIII–D AS STEADY-STATE HIGH PERFORMANCE
ADVANCED TOKAMAK (AT) SCENARIO Reduced current drive requirements
with aligned bootstrap is predicted— Ozeki et al., Nucl. Fusion 33, 1025 (1993)— Kessel, Phys. Rev. Lett. 72, 1212 (1994)— Manickam, Phys. Plasma 1, 1601 (1994)— Turnbull, Phys. Rev. Lett. 74, 718 (1995)
Improved performance observed experimentally— Reduced core transport observed
in a number of experiments— Highest performance in DIII–D is in NCS
with H–mode edge
Great progress in understanding the iontransport in internal transport barriers— ωE×B > γLIN— χ ~ χNeo 0 < ρ < 1
Challenges remain in understandingelectron thermal transport
Reverse Magnetic Shear
Current profile control and transport barriercontrol is needed to increase duration
Broad pressure profiles and wall stabilizationare needed for improved stability
016–98 QTYUIOP
SELF-CONSISTENT SIMULATIONS INDICATE DESIRED CURRENTDENSITY PROFILE CAN BE OBTAINED WITH OFF-AXIS ECCD
— Extended duration —
βN = 4, H = 2.8
Ip = 1.0 MA, ne = 3.2 × 1019
PEC = 2.3 MW, IEC = 0.15 MA
150
0.0ρ
100
50
00.5 1.0
J(A/
cm2 )
Jtot JEC + JNBJboot
— Fully penetrated profiles —
βN = 5.7, H = 3.6
Ip = 1.6 MA, ne = 5.7 × 1019
PEC = 7 MW, IEC = 0.32 MA
NATIONAL FUSION FACILITYDIII–D 031–99/TCS
KEY CHALLENGE FOR NCS SCENARIO
⇒ Consistency of resulting pressure profile in discharges with transportbarriers with stability at high beta
Steep pressure gradients at ITB or at the boundary can lead toinstability at low beta
⇒ Develop scenarios that have "naturally" favorable profiles
— Requires fundamental understanding of ITB, and stabilityboundaries
— Magnetic shear has an input
or
⇒ Develop transport barrier control techniques
NCSNCS
OPTIMIZEDSCENARIO
(SUSTAINED)
WALL STABILIZATIONFEEDBACK
PHYSICS UNDERSTANDING DRIVES DIII–D AT RESEARCH PLAN1999
PHYSICSPRINCIPLES
PLASMACONTROL
INTEGRATEDPHYSICS
2000 2001 2002
WALL STABILIZATIONPRINCIPLES
NTM PHYSICS NTM CONTROL
NCS DEVELOP
EDGE STABILITYPHYSICS
COUNTER NBI RF CDCOUNTER
ADVANCED TOKAMAK
ITB PHYSICS
OPTIMAL EDGE ANDDIVERTORδ, SN/DN
TOOLAPPLICATION
OPTIMALMODE SPECTRUM
AT DIVERTORNEUTRAL, FLOW
CONTROL
HIGH liSCENARIO
DEVELOPMENT
INTERMEDIATESCENARIO
(EXISTENCE)
HIGH li(EXISTENCE)
045–99NATIONAL FUSION FACILITYDIII–D
NA
TIO
NA
L F
USIO
N F
AC
ILIT
YS
AN
D
IE
GO
DIII–D
031-99 KHB/jy
STEEPEST CORE G
RADIENTS FORM
IN SHOTS
WITH NEG
ATIVE MAG
NETIC SHEAR
20
Shot 959891500 m
s10
Ti (keV)Vϕ/R (103s–1)ne (1019m–3)
5066
44
22
00
0.00.2
0.40.6
ρ0.8
1.0
q
200
100 15
INTERNAL TRANSPORT BARRIER (ITB) HAS BEEN SUSTAINEDFOR ~5 s IN DIII–D L–MODE DISCHARGES
NATIONAL FUSION FACILITYS A N D I E G O
DIII–D011–99/RDS
Fully penetrated current, profile with weak central magnetic shear and q > 1
Tem
pera
ture
(keV
)
0
5
10
0.0 0.2 0.4 0.6 0.8 1.00
2
4
6
8
Ti
Te
q
ρ
I p (M
A)T i
(keV
)
0
5
10
0
1
0
24
Pb (M
W)
0 1000 2000 4000 5000Time (ms)
3000
R (m) 1.834 1.930 1.993 2.043 2.123 2.200
ITB
BES shows suppression of turbulence ρ < 0.4∼ Key to long-pulse ITB is sustainment of the current profile
IAEA-F-CN-69 EX5/6 Synakowski
MICROWAVE ELECTRON CYCLOTRON HEATINGPROVIDES LOCALIZED CURRENT DRIVE
031-99 tcs/rsNATIONAL FUSION FACILITYDIII–D
J ECCD
(A c
m–2
)
0.0 0.2 0.4 0.6 0.8 1.0
IECCD = 92 kA
–200
20
40
60
80
100
ρ
ρ
0.0 0.2 0.4 0.6 0.8 1.0
0–2–4
2468
10
J ECC
D (A
cm
–2)
IECCD = 35 kA Second Harmonic Resonance
SteerableAntenna
ρ = 0.15
ρ = 0.5
NATIONAL FUSION FACILITYDIII–D 031–99/TCS
NCS DISCHARGES CAN BE FURTHER CATEGORIZEDBY THE EDGE CONDITION
L–Mode Edge ELMing H–Mode Edge ELM-Free H–Mode Edge
Advantages Low p′ edge Large SOL Neoclassical
transport in core Good beam
penetration
Steady-state edge ELMs purge impurities Cryopump can control
density
Highest βNH product Neoclassical χi over
entire radius Better bootstrap
alingment High performance even
with Ti ~ Te
PresentLimitations/Status
H ~ 2.3, βN ~ 2.3 Low beta limit Misaligned
bootstrap current Small core volume Control of qmin
H ~ 2.4, βN ~ 2.8 Difficult to establish ITB Type I ELM perturbations
are too large Neoclassical tearing
modes
Little control over edgedensity, p′ and Jbs
⇒ peeling modes Carbon impurity
accumulation at edge Pump ineffective for
density control
ResearchEffort/Directions
Expand radius ofqmin
Control core p′ usingRF
Use ECH to slowcurrent diffusion
Reduce edge 2nd stabilityaccess and ELM size byvarying shape(squareness)
Sustain ELMing H–modeduring Ip ramp foradditional profile control
Change edgecollisionality to reduceJbs
Trigger ELMs prior to X-event (Global MHD event)using pellets or other?
THE PATH TO THE NCS–AT GOAL LEADS THROUGHMANY STABILITY ISSUES
031-99 rds/jyNATIONAL FUSION FACILITYDIII–D
PressureProfile
BroadHigher H
Larger ITBBetter Bootstrap
Alignment
H–mode EdgeBroader P (r)
NCS–ATGoal
Edge Stability(T1, T5)
Neoclassical Tearing(T3)
Wall Stabilization(T4)
L–mode EdgeMore Peaked P (r)
Kink Modeβ–limit
PeakedLower H
Narrow ITB
Neoclassical Tearing
(T3)Stable? Low ∇Pin s > 0 Region
031-99 RDS/jyNATIONAL FUSION FACILITYS A N D I E G O
DIII–D
EXPERIMENTAL β LIMITS CONSISTENT WITHCALCULATED DEPENDENCE ON po/⟨p⟩
TFTR high po/⟨p⟩ ~ 6.0 (ERS–mode):βN < 2
— Limited by fast n = 1 disruption~
DIII–D high po/⟨p⟩ ~ 6.0 (L–mode):βN < 2.5
— Limited by fast n = 1 disruption~
DIII–D low po/⟨p⟩ ~ 1.5 (H–mode):βN < 4
— No disruptionlimited by ELM-like activity fromfinite edge pressure gradients
~
0 2 4 6 8 10
Resistive
Ideal
Unstable
H–modeL–mode
P(O) / ⟨P⟩
β N (%
-m-T
/MA)
HIGH BETA REQUIRES BOTH BROAD PRESSURE AND STRONG SHAPING
QTYUIOP124–97
Calculated ideal n = 1 stability limit, wall at r/a = 1.5
Fixed q profile: q0 = 3.9, qmin = 2.1, q95 = 5.1
2 3 4 5 61
3
5
βNδ = 0.7κ = 1.8
βN = β (I/aB)–1 β∗ = <p2>1/2 2 µo/B2
δ = 0κ = 1.0
p0 /⟨p⟩
10
8
6
4
2
02 4 63 5
β*
κ = 1.8 δ = 0.7
κ = 1.0 δ = 0
p0 /⟨p⟩[A. Turnbull, IAEA 1996]
THE PATH TO THE NCS–AT GOAL LEADS THROUGHMANY STABILITY ISSUES
031-99 rds/jyNATIONAL FUSION FACILITYDIII–D
PressureProfile
BroadHigher H
Larger ITBBetter Bootstrap
Alignment
H–mode EdgeBroader P (r)
NCS–ATGoal
Edge Stability(T1, T5)
Neoclassical Tearing(T3)
Wall Stabilization(T4)
L–mode EdgeMore Peaked P (r)
Kink Modeβ–limit
PeakedLower H
Narrow ITB
Neoclassical Tearing
(T3)Stable? Low ∇Pin s > 0 Region
ENERGY CONFINEMENT TIME INCREASES WITHTHE PEDESTAL PRESSURE IN DIII–D ELMing H–MODES
NATIONAL FUSION FACILITYS A N D I E G O
DIII–D248–98
QTYUIOP
"Stiff" transport models predictτE increasing with PPED
For fixed shape H ∝ βPED1/2
ITER SHAPE(fixed)δ = 0.24κ = 1.75ε = 0.34
H-I
TE
R93
H
0 2 4 60
0.4
0.8
1.2
1.6
Type I ELMsType III ELMsL–mode
ITER Shape, q = 3.2, I = 1.5 MA95 P
PPED (kPa)elec
m = 4m = 5
m = 6m = 7
q = 1 q = 4/3 5/3 2
PressurePedestal
m = 8m = 9
Discharge #75121
n = 3VH–Mode
0.0 0.2 0.4 0.6 0.8 1.0ρ
0.0 0.2 0.4 0.6 0.8 1.0ρ
1.0
0.0
0.5
m = 4m = 5
m = 6
m = 7
x m
1.0
0.0
0.5
x m
q = 4/3 q = 2
m = 8
n = 3
Discharge #92001
GATO CALCULATIONS
ELM
WIDTH OF THE EDGE MODE IS LARGER WITHA LARGER PRESSURE PEDESTAL
031-99 RDS/jyNATIONAL FUSION FACILITYDIII–D
VH–mode (βN = 3): termination H–mode (βN = 2): ELM PressurePedestal
confinement.
EMERGING EDGE STABILITY PICTURE SUGGESTS NEED TO REGULATETHE EDGE BOOTSTRAP CURRENT AND/OR THE EDGE PRESSURE
GRADIENT TO EXTEND THE DURATION OF AT MODES (T1)
Need ELMs to provide density and impurity control
Pressure gradient and bootstrapcurrent work in a positive feedbackloop until second stability limitis reached
But, most of our high performance discharges are terminated by an ELM that couplescouples tolow–n modes, causing a drop in plasma confinement
Need to regulate the edge ∇P and or JBS either through changes in transport or stability
Techniques: shaping, impurity radiation, fueling, edge ECH
Reducing the edge current should, in general, be the favorable path since this willreduce the kink-like (low–n) character of the edge instability
Reference Equilibrium:κ = 1.8, δ = 0.3, A = 170/65, ψped = 0.98,ψwid = 0.0125, q95 = 3.5, and qaxis = 1.1
031–99 RDS/jyNATIONAL FUSION FACILITY
S A N D I E G O
DIII–D
Unstable
Un
Stable
Stable Equil. p'
0.9 0.95 1.0
p'
ψ~
SHAPING: HIGH AND LOW SQUARENESS ELIMINATES SECONDSTABLE ACCESS AND HAS A LARGE IMPACT ON ELMs
031–99/RDS/jyNATIONAL FUSION FACILITYS A N D I E G O
DIII–D
1
0
0
4
8
1.9
Edge Te (keV)
βN = 1.8
UnstableUnstable
βN = 1.9
Edge Te (keV)
Dα (a.u.) Dα (a.u.)
Time (s) Time (s)2.0
0.8ψ
Nψ
N
1.00.9 0.8 1.00.9
2.5 2.6
IAEA-F1-CN-69/EX8/1 Lao
No second regime accesssmall ELMs
Second regime accesslarge ELMs
THE PATH TO THE NCS–AT GOAL LEADS THROUGHMANY STABILITY ISSUES
031-99 rds/jyNATIONAL FUSION FACILITYDIII–D
PressureProfile
BroadHigher H
Larger ITBBetter Bootstrap
Alignment
H–mode EdgeBroader P (r)
NCS–ATGoal
Edge Stability(T1, T5)
Neoclassical Tearing(T3)
Wall Stabilization(T4)
L–mode EdgeMore Peaked P (r)
Kink Modeβ–limit
PeakedLower H
Narrow ITB
Neoclassical Tearing
(T3)Stable? Low ∇Pin s > 0 Region
NA
TIO
NA
L F
USIO
N F
AC
ILIT
YS
AN
D
IE
GO
DIII–D
FUTURE WO
RK: RADIALLY LOCALIZED O
FF–AXIS ECCDFO
R SUPPRESSION O
F NEOCLASSICAL TEARING
MO
DES
MO
TIVATION – Expected beta lim
it for ITER-like discharge
—Island sustained by “m
issing” bootstrap current in O–PT
G
OAL – Replace “m
issing” bootstrap current (Hegna, Callen, Zohm)
—Suppress island size or m
ake vanish
3
1.0
2960 ms
before 3/2 island
3120 ms
with 3/2 island
(and 2/2 kink?)
# 86176
0.80.6
0.40.2
q ≈ 2/2q ≈ 3/2
ρ0.0
0.1
ε1/2 ∇P/Bθ (MA/m2)
0.0
0.2
0.3
ECHf = 110.0 G
Hzfacet ang = 19.0 degtilt ang = 52.0 deg
2τRr
0 1
12
–1
34
56
SaturatedIsland
Modified Rutherford Eqn
(unmodulated,
δFW
HM = 3 cm,
localized at q = 3/2)
W (cm
) Iaux = 0
Iaux = 30 kA
dwdt
011-99 EJS/jy
THE PATH TO THE NCS–AT GOAL LEADS THROUGHMANY STABILITY ISSUES
031-99 rds/jyNATIONAL FUSION FACILITYDIII–D
PressureProfile
BroadHigher H
Larger ITBBetter Bootstrap
Alignment
H–mode EdgeBroader P (r)
NCS–ATGoal
Edge Stability(T1, T5)
Neoclassical Tearing(T3)
Wall Stabilization(T4)
L–mode EdgeMore Peaked P (r)
Kink Modeβ–limit
PeakedLower H
Narrow ITB
Neoclassical Tearing
(T3)Stable? Low ∇Pin s > 0 Region
0.0 0.2 0.4 0.6 0.8 1.0
RPlasma /Rwall
0
5
10
15 = 1.5
βN
p0/<p> = 4.8
p0/<p> = 2.4
Wall atInfinity
βN< 3.5
R PlasmaRwall
0.2 0.4 0.6 0.8 1.00.0
J
JBS
JFTJ
(MA/m2)
2.0
0.0
1.0
ρ
f ~70%BS βN =5.5
p0/<p> = 4.8
p0/<p> = 2.4
031-99 RDS/jyNATIONAL FUSION FACILITY
S A N D I E G O
DIII–D
WALL STABILIZATION IS CRUCIALFOR ADVANCED TOKAMAK OPERATION
AT operation with
— High normalized beta, βN = β/I/aB— Large bootstrap current fraction— Good bootstrap current alignment
Requires plasmas with
— Broad pressure profile— Broad current profile
Such plasmas have
— Low βN stability limit to n = 1external kink without a wall
— Significantly higher limit with aconducting wall
NATIONAL FUSION FACILITYS A N D I E G O
DIII–D038–99/EJS
IDEAL KINK MODE STABILIZED BY ROTATION AND RESISTIVE WALLABOVE NO-WALL βN LIMIT FOR > 30 τwall
Wall stabilization sustained with βN up to 1.4 × βN Plasma rotation slows as βN exceeds the no-wall limit
Resistive wall mode grows when rotation drops below a critical value
0
80
60
40
20
3
2
1
PNB = 10 MW
βN No Wall Limit
Plasma ToroidalRotation
δBr (n = 1)
Discharge 92544
1.51.1 1.3 1.41.2Time (s) RWM Onset
qmin ~ 2
q = 3
2015
5
0
10kHz
gaus
s
100
–100
0
–200
60 Hz
1.465 1.470 1.475 1.480 1.485Time (s)
Outboard dBθ /dt (T/s)
10–1
Toro
idal
Ang
le
no-wall
RESISTIVE WALL MODE (RWM) IS SUPPRESSED (~30 ms)BY OPEN-LOOP ACTIVE CONTROL
NATIONAL FUSION FACILITYS A N D I E G O
DIII–D248–98
QTYUIOP
Near stationary RWM is reproducibly obtained
IAEA F1-CN-69/EXP3/10 Strait
Feed-forward static n = 1 field is preprogrammed at RWM onset, with phase opposing the mode
kHz
This result is encouraging for active feedback
o
x
C-coil
o
x
o
x
PlannedC-coil
extension
8
4
01.20 1.25 1.30
Plasma rotationat q = 3
Time (s)
2
1
βN
96625
96633
Onset of RWM
kA
5.0
2.5
C–coil current
2.0
keV
3.0 Te at R = 2.1 m
2.5
Gaus
s
4.0
2.0
δBr
VALEN 3D FEEDBACK CONTROL MODEL PREDICTS IMPROVED β LIMIT IN DIII–D
• Existing 6 coil set can increase RWM stability limit to β ~ 3.4N
βN
Gro
wth
Rat
e (s
)
–1
1 2 3 4 51
100
104
106
ideal wallβ limit
no wallβ limit
ideal kink
resistivewall mode
No feedback6 coil feedback
031-99 RDS/jy
NATIONAL FUSION FACILITYS A N D I E G O
DIII–D
N
• Extended 18 coil set can increase RWM stability limit to β ~ 3.8(not optimized design)
18 coil feedback
• Gain in fusion powerat fixed bootstrapfraction = = 2.6! (3.8
3.0 )4
NCSNCS
OPTIMIZEDSCENARIO
(SUSTAINED)
WALL STABILIZATIONFEEDBACK
PHYSICS UNDERSTANDING DRIVES DIII–D AT RESEARCH PLAN1999
PHYSICSPRINCIPLES
PLASMACONTROL
INTEGRATEDPHYSICS
2000 2001 2002
WALL STABILIZATIONPRINCIPLES
NTM PHYSICS NTM CONTROL
NCS DEVELOP
EDGE STABILITYPHYSICS
COUNTER NBI RF CDCOUNTER
ADVANCED TOKAMAK
ITB PHYSICS
OPTIMAL EDGE ANDDIVERTORδ, SN/DN
TOOLAPPLICATION
OPTIMALMODE SPECTRUM
AT DIVERTORNEUTRAL, FLOW
CONTROL
HIGH liSCENARIO
DEVELOPMENT
INTERMEDIATESCENARIO
(EXISTENCE)
HIGH li(EXISTENCE)
045–99NATIONAL FUSION FACILITYDIII–D
1999 2000 2001 2002 2003
Inside launch pellet Inside Divertor Pump and Baffle
18 Coil Feedback6 Coil Feedback
3 MW ECH
Fueling and Edge Control
CurrentDrive
Resistive WallMode Control
Operation
10 MW ECH
10-s pulseLower divertor
Liquid jet
Counter NBI
• Verify off-axis ECCD
• βN = 1.4βN(no wall)
• Increase NTM onset threshold (βN)
• χi ~ neoclassical
• βN*H98 ≥ 6 (1 s); ≥ 3.5 (25 τE)
6 MW ECH 10 MW ECH
Progress Checkpoint
Simultaneously Demonstrate High: βN, H98, fBS, radiative divertorAT Research Program
Extend duration, performance, reactor similarity
Upgrade Options
DIII–D ADVANCED TOKAMAK 5–YEAR RESEARCH PLAN
031-99 tcs/rsNATIONAL FUSION FACILITYDIII–D
Hardware upgrades with new diagnostics in 1999–2000 supports a two-phase AT physics development and integration plan
An initial test of AT integration with a progress checkpoint in 2001will evaluate upgrade options to extend AT integration
Physics IntegrationPhysics Development
NATIONAL FUSION FACILITYS A N D I E G O
DIII–D 045–99
DIII–D AT PROGRAM:REMAINING CHALLENGES AND OPPORTUNITIES
Understand transport barrier dynamics; broaden pressure profiles
— Develop ITB control as needed
Implement methods to sustain hollow current profiles and high bootstrap fractions
Deepen the physics understanding of neoclassical tearing modes; avoid or stabilize
Confirm our edge stability physics picture; find a compromise
Understand the physics of wall stabilization; implement feedback
WHAT DO WE HAVE TO GAIN?
An understanding of the ultimate potential of the tokamak as a magneticconfinement system
Greatly increased fusion power output
Much improved prospects for steady-state