Advanced Tokamak Plasmas and Their Control

1
Advanced Tokamak Plasmas and Their Control

• C. Kessel
• Princeton Plasma Physics Laboratory
• Columbia University, 4/4/03

2
Power Plant Studies Show the Potential Benefits
of Advanced Tokamaks

• Simultaneous achievement of
• Steady state
• High ? ----gt high fusion power density
• Large bootstrap (self-driven) current ----gt low
  recirculating power
• Good energy confinement consistent with the high
  ? and high fBS ----gt high fusion gain
• This combination drives down the machine size and
  cost of electricity (COE)
• High potential benefits of Advanced Tokamak
  operation make AT research on any Burning Plasma
  Experiment mandatory (Snowmass 1999)
• Present tokamak experiments pursuing AT plasma
  physics worldwide

3
ARIES-AT Power Plant Design Provides a Goal for
AT Research
Ip12.8 MA, BT5.9 T, R5.20 m, a1.30 m,
?X2.20, ?X0.9
?N5.4 ---gt can we stabilize n1-4 with feedback?
Do we need to? I?P/Ip0.91, ICD1.25 MA ---gt can
the current profile be dominated by bootstrap
current, and can it be controlled? Pressure
profile optimized to maximize ?N, and T and n
chosen to maximize bootstrap current with ITB in
location of qmin ---gt can transport be controlled
to provide this? Plasma edge must be consistent
with the divertor, CD, power handling ---gt what
can be produced and controlled?
4
ARIES-AT Physics Basis

• n1 RWM feedback control with coils do we need
  to stabilize ngt1? Use higher order coils for
  higher n
• No plasma rotation source
• 37 MW LHCD and 5 MW (25 MW capable) ICRF/FW for
  external current drive/heating
• HHFW and NBI (120 keV) also shown capable of
  providing CD
• NTM stability are (5,2) and (3,1) unstable? LH
  current profile modification (?) at (5,2)
• 90bootstrap current fraction
• Strong plasma shaping
• Double null
• Tungsten divertors allow high heat flux

• Vertical and kink passive stability tungsten
  structures in blanket, feedback coils behind
  shield
• Transport assumed roughly agreed with GLF23 new
  versions of GLF23 are now available
• Very low ripple (0.02)
• n/nGreenwald 1
• Plasma edge and divertor solution balancing of
  radiating mantle and radiating divertor, with Ar
  impurity
• High field side pellet launch allows fueling to
  core, and ?P/?E10 allows sufficiently low
  dilution

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ARIES-AT Layout
6
Next Step Devices Must Provide Basis for Tokamak
Power Plant Regime
ITER, KSTAR, JT-60U have super-conducting
coils ITER is DT Shielding required JT-60SC and
KSTAR are DD
FIRE has Cu coils FIRE is DT Minimal shielding
AT
FIRE
KSTAR
Inductive
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Objectives of FIRE

• Develop the experimental/theoretical basis for
  burning plasma physics
• Q 10 ELMy H-mode for ?burn gt 2 ? ?cr
• Q 5 Advanced Tokamak for ?burn gt 1-5 ? ?cr
• Adopt as many features as possible of projected
  Power Plant designs
• Only address technological issues required for
  successful device operation
• Fueling, pumping, power handling, plasma control,
  neutronics, materials, remote handling, and
  safety
• Utilize the compact high-field Cu coil approach
  to keep the device cost at 1 B

8
Fusion Ignition Research Experiment
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FIREs Efforts to Self-Consistently Simulate
Advanced Tokamak Modes
0-D Systems Analysis Determine viable operating
point global parameters that satisfy constraints
Plasma Equilibrium and Ideal MHD
Stability Determine self-consistent stable
plasma configurations to serve as
targets Heating/Current Drive Determine current
drive efficiencies and deposition
profiles Transport(GLF23 and pellet fueling
models to be used in TSC) Determine plasma
density and temperature profiles consistent with
heating/fueling and plasma confinement Dynamic
Evolution Simulations Demonstrate
self-consistent startup/formation and control
including transport, current drive, and
equilibrium Edge/SOL/Divertor Find
self-consistent solutions connecting the core
plasma with the divertor
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0D Analysis Includes

• Power balance energy confinement time scaling
• Fusion cross-section from Bosch-Hale formulation
• Particle balance self-consistent helium content
  (input ?He/?E), quasi-neutrality, input impurity
  fractions
• Radiation bremsstrahlung, cyclotron (new Albajar
  formulation), line (coronal equilibrium)
• Current diffusion time, flux consumption
  (Hirshman-Nielson) with neoclassical resistivity
• Bootstrap current (equilibrium fits) and external
  CD input CD efficiency
• Fast alpha beta contribution
• Parabolic or parabolicpedestal profiles
• Post-processor used for database screening

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0D Power/Particle Balance Identifies Operating
Space for FIRE AT

• Heating/CD Powers
• ICRF/FW, 30 MW
• LHCD, 30 MW
• Using CD efficiencies
• ?(FW)0.20 A/W-m2
• ?(LH)0.16 A/W-m2
• P(FW) and P(LH) determined at r/a0 and r/a0.75
• I(FW)0.2 MA
• I(LH)Ip(1-fbs)
• Scanning Bt, q95, n(0)/ltngt, T(0)/ltTgt, n/nGr, ?N,
  fBe, fAr

• Q5
• Constraints
• ?(flattop)/?(CR) determined by VV nuclear heat
  (4875 MW-s) or TF coil (20s at 10T, 50s at 6.5T)
• P(LH) and P(FW) max installed powers
• P(LH)P(FW) Paux
• Q(first wall) lt 1.0 MW/m2 with peaking of 2.0
• P(SOL)-Pdiv(rad) lt 28 MW
• Qdiv(rad) lt 8 MW/m2

Generate large database and then screen for
viable points
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FIREs Q5 AT Operating Space
Access to higher tflat/?j decreases at higher ?N,
higher Bt, and higher Q, since tflat is set by VV
nuclear heating Access to higher radiated power
fractions in the divertor enlarges operating
space significantly
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Observations from 0D Analysis for Burning Plasma
AT

• In order to provide reasonable fusion gain Q5,
  cant operate at low density to maximize CD
  efficiency
• Density profile peaking is beneficial (pellets or
  ITB), since broad densities increase required H98
  and PCD
• Access to high density relative to Greenwald
  density, in combination with high bootstrap
  current fraction gives the lowest required H98
• H98 1.4 are required to access ?flattop/?curr
  diff gt 3, however, the ELMy H-mode scaling law is
  known to have a ? degradation that is not
  observed on individual experiments
• Radiative core/divertor solutions are a critical
  area for the viability of burning AT experiments
  due to high P?PCD, suggesting impurity control
  techniques

14
FIREs AT Operating Space
Q 5-10 accessible ?N 2.5-4.5 accessible fbs
50-90 accessible tflat/tj 1-5
accessible If we can access.. H98(y,2)
1.2-2.0 Pdiv(rad) 0.5-1.0 P(SOL) Zeff
1.5-2.3 n/nGr 0.6-1.0 n(0)/ltngt 1.5-2.0
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Examples of Q5 AT Points That Obtain ?flat/?J gt 3
HH lt 1.75, satisfy all power constraints,
Pdiv(rad) lt 0.5 P(SOL)
?n ?n? ?T ?T? BT q95 Ip HH fGr fBS Pcd P? zeff fBe fAr t/?
0.5 2.60 1.5 8.17 6.5 4.25 4.25 1.71 0.8 0.80 27.5 27.8 2.08 1 .3 3.58
0.5 2.93 2.0 7.28 6.5 4.25 4.25 1.57 0.9 0.80 30.9 31.4 1.77 1 .2 3.95
0.75 3.10 1.5 7.83 6.5 3.75 4.82 1.46 0.9 0.80 33.1 36.5 1.89 2 .2 3.07
0.75 2.91 1.0 7.71 6.5 4.00 4.52 1.62 0.9 0.85 24.7 28.6 1.77 1 .2 3.52
0.75 3.23 1.5 7.00 6.5 4.00 4.52 1.54 1.0 0.85 27.5 32.0 2.08 1 .3 4.40
0.75 2.44 1.5 8.90 6.5 4.25 4.25 1.74 0.8 0.91 16.0 28.0 2.20 2 .3 3.65
1.00 3.49 1.0 7.35 6.5 3.50 5.16 1.36 1.0 0.83 32.6 38.6 1.77 1 .2 3.00
1.00 3.26 1.0 7.60 6.5 3.75 4.82 1.54 1.0 0.89 23.9 30.1 2.01 3 .2 4.00
1.00 2.44 1.5 9.59 6.5 4.00 4.52 1.65 0.8 0.95 13.6 31.5 2.32 3 .3 3.29
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Dynamic Simulations of FIRE AT Discharges with
TSC-LSC

• Free-boundary time-dependent simulation 2D MHD
  equations, Maxwells equations, and 1D transport
  equations for particles, energy, and current,
  coupled thru boundary conditions to the PF coils
• Physics models
• Transport coefficients
• Heating/fueling deposition for alphas, NBI, ICRF,
  etc.
• Current drive and bootstrap current
• Sawteeth
• Radiation
• Impurity transport
• Feedback control systems
• High-n ballooning
• LSC is a lower hybrid ray-tracing code

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TSC Model
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TSC-LSC Simulation of Q5 FIRE AT Discharge
Ip 4.5 MA, Bt 6.5 T, ?N 4.1, H981.7,
n/nGr 0.85, n(0)/?n? 1.45 ? 4.7, ?p 2.35,
?flattop/?curr diff 3.5, Zeff 2.2, q(0)
4.0, qmin 2.7, q95 4.0
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TSC-LSC Simulation of Q5 AT Burning Plasma
During flattop, t10-41s
li(3)0.42
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TSC-LSC Simulation of Q5 AT Burning Plasma
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MHD in FIRE AT Plasmas and its Control

• n8 ballooning modes ---gt limit pressure
  locally, not observed experimentally since very
  localized, self-limiting by adjusting profile to
  be marginally stable
• n0 vertical instability ---gt slowed with
  conducting structures and controlled with coils
  that provide a radial magnetic field
• n1 external kink modes (resistive wall modes)
  ---gt disruptive, slowed with conducting
  structures and can be controlled with plasma
  rotation and/or direct feedback with saddle
  coils, strong influence of error fields
• 1 lt n lt 4 external kink modes ---gt disruptive??,
  behavior similar to n1, however, more localized
  toward the plasma boundary, and may set lower
  ?-limit than n1, should be controllable like n1
  if necessary
• 4 lt n lt 20 peeling modes ---gt ballooning and kink
  mode character, localized to the plasma edge,
  associated with pressure pedestal and associated
  bootstrap current, and considered primary
  candidate for ELMs, plasma shaping has
  significant influence
• Neo-classical tearing modes ---gt non-disruptive
  but reduce achievable ? in long pulse discharges,
  controllable with current driven at island or by
  modifying current profile to increase ?

22
Updating FIRE AT Equilibrium Targets Based on
TSC-LSC Equilibrium
TSC-LSC equilibrium Ip4.5 MA Bt6.5 T q(0)3.5,
qmin2.8 ?N4.2, ?4.9, ?p2.3 li(1)0.55,
li(3)0.42 p(0)/?p?2.45 n(0)/?n?1.4 Stable
n? Stable n1,2,3 with no wall
vV/Vo
23
Stabilization of n1 RWM is a High Priority on
FIRE
Feedback stabilization analysis with VALEN shows
strong improvement in ?, taking advantage of
DIII-D experience, most recent analysis indicates
?N(n1) can reach 4.2
What is impact of n2??
24
(No Transcript)
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Stabilization of n1 RWM on DIII-D
Experiments on DIII-D have verified plasma
rotation stabilization by reducing the error
fields (amplified by RWMs) that slow the plasma
down, and VALEN analysis shows that better
sensors and in-vessel feedback coils strongly
improve ?N
26
Theoretical Results for n1 RWM Stabilization
from MARS and VALEN
VALEN shows that feedback can work with detailed
structure and coil model
MARS shows that feedback can work with simple
structure and coil model
HBT-EB
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How Do n2-4 Manifest Themselves if They Are
Linearly Ideal Unstable
Shape study on DIII-D AT plasmas
n2 and n3 would not allow access to the n1
?-limit These modes appear too broad to be
peeling modes This feature is common from wall
stabilized ideal MHD analysis Are these modes
triggering tearing modes that subsequently become
NTMs?? ---gt DIII-D
wall at 1.5a
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Neo-Classical Tearing Modes for FIRE AT Modes
Target Bt6.5-7 T for NTM control, to utilize 170
GHz from ITER RD Must remain on LFS for
resonance and use O-mode, due to high Bt ECCD
efficiency?? (trapping)
Can we avoid NTMs with j(?) and qgt2.0 or do we
need to suppress them??
Ro
Roa
Bt6.5 T
?
fce182
fce142
170 GHz
Ro
Roa
Bt7.5 T
?
?
fce210
fce164
200 GHz
Ro
Roa
Bt8.5 T
?
fce190
fce238
Can we rely on OKCD to suppress NTMs far
off-axis on LFS versus ECCD ?? (enhanced Ohkawa
affect at plasma edge)
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J. Decker, APS 2002,MIT
OKCD allows LFS EC deposition, with similar A/W
as ECCD on HFS
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Comments on ECCD in FIRE

• ASDEX-U shows that 3/2 island is suppressed for
  about 1 MW of power with IECCD/Ip 1.6, giving
  0.013 A/W
• Ip0.8 MA and ?N2.5
• DIII-D shows that 3/2 island is suppressed for
  about 1.2-1.8 MW with jEC/jBS 1.2-2.0
• Ip1.0-1.2 MA, ?N2.0-2.5
• OKCD analysis of Alcator-CMOD gives about 0.0056
  A/W
• FIREs current requirement should be about 15
  times higher than ASDEX-U (scaled by Ip and ?N2)
• Need about 200 kA, which would require about 35
  MW?? Early detection reduces power alot according
  to ITER
• Do we need less current for 5/2 or 3/1, do we
  need to suppress them??
• Is 170 GHz really the cliff in EC technology??

MIT, short pulse results
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FIRE EC Geometry
?ce ?
n(0)4.5?1020 f pe9?vn
Rays are bent as they ? approaches ?pe
EC launcher
Rays must be launched with toroidal
directionality for CD
?pe gt ? cutoff for 170 GHz
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Neo-Classical Tearing Mode Stabilization on
DIII-D, ASDEX-U and JT-60U
Actively stabilize NTMs ---gt must spatially
track island ECCD LHCD (Compass-D) Passively
avoid NTMs q gt 2? J(?) that is stable?
DIII-D
Required IECCD scales as Ip and ?N2
ASDEX-U
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Heating and Current Drive for FIRE AT Plasmas and
its Control

• ICRF ion heating on and off-axis
• ICRF/FW for electron heating and on-axis CD
• LH for off-axis CD and electron heating
• EC for NTM control off-axis deposition (no
  analysis yet)
• NBI?? presently being examined (no AT analysis
  yet)
• High energy needed for ELMy H-mode, not practical
• ATs have slightly lower density, more density
  peaking, and off-axis deposition is desirable
  ---gt prefer conventional energies 120 keV
• Heating and Current Drive directly affect
  Transport

34
ICRF Ion Heating
80-120 MHz, 2 strap antennas, 4 ports, 20 MW (10
MW upgrade) He3 minority, 2T, 2D, H minority
accessible resonances at center and off-axis
(C-Mod ITB) ----gt full wave analysis gives 75
power on ions
35
ICRF/FW Viable for FIRE On-Axis CD
Calculations assume same ICRF ion heating system
frequency range, approximately 40 of power
absorbed on ions, can provide required AT on-axis
current of 0.3-0.4 MA with 20 MW (2 strap
antennas)
PICES (ORNL) and CURRAY(UCSD) analysis f
110-115 MHz n 2.0 n(0) 5x1020 /m3 T(0)
14 keV 40 power in good part of spectrum (2
strap) ----gt 0.02-0.03 A/W CD efficiency with 4
strap antennas is 50 higher Operating at lower
frequency to avoid ion resonances, vph/vth??
E. Jaeger, ORNL
36
Benchmarks for LHCD Between LSC and ACCOME
(Bonoli)
Trapped electron effects reduce CD
efficiency Reverse power/current reduces forward
CD Less than 1.0 MW is absorbed by
alphas Recent modeling with CQL and ACCOME/LH19
will improve CD efficiency, but right
now.. Bt8.5T ----gt 0.25 A/W-m2 Bt6.5T ----gt
0.16 A/W-m2 FIRE has increased the LH power from
20 to 30 MW
f4.6 GHz n 2.0 ?n0.3
37
Energy and Particle Transport in FIRE AT Plasmas
and Its Control

• Significant reductions in particle and energy
  transport have been achieved at the plasma edge
  (ETB) and in the core (ITB)
• Most present tokamaks have NBI, which provides
  sheared rotation and strongly stabilizes
  micro-instabilities
• Negative magnetic shear and Shafranov shift are
  also found to stabilize microinstabilites
• Heating, current drive, rotation, pellets, and
  impurities are found to influence transport
• It appears that transport barriers can be made to
  leak a little to avoid excessive particle
  buildup
• Lots of other observations ---gt C-Mod ITBs with
  off-axis ICRF, JET ITBs triggered by qmin
  passing rational surfaces,
• The control of transport (pressure profile
  control) is critical to achieving high bootstrap
  current fractions, that remain MHD stable
• The transport barrier may be an ideal method for
  controlling the pressure profile ---gt by turning
  the ITB on and off, with a given frequency, a
  desirable pressure profile could be produced

38
Transport Has Multiple Scales and Multiple
Stabilization Mechanisms
39
GLF23 AT Predictive Modeling Is Improving
Kinsey
Ti, Te
V?
GLF23, ver. 1.61
model
DIII-D
expt
40
FIRE Pellet Launch Geometry
41
HFS Launch V125 m/s, set by ORNL pellet tube
geometry Vertical and LFS launch access higher
velocities
42
HFS Pellet Launch and Density Peaking ---gt Needs
Strong Pumping
Simulation by W. Houlberg, ORNL, WHIST
FIRE reference discharge with uniform pellet
deposition, achieves n(0)/ltngt 1.25
P. T. Lang, J. Nuc. Mater., 2001, on ASDEX and
JET L. R. Baylor, Phys. Plasmas, 2000, on DIII-D
43
FIRE Uses Cryo-Pumping Coupled to Turbopumps
44
FIREs Divertor Must Handle Attached(25 MW/m2)
and Detached(5 MW/m2) Operation
D. Dreimeyer, M. Ulrickson
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Other Issues for FIRE AT Plasmas

• Alpha particle losses from ripple, aggravated by
  high safety factor and low Ip and Bt
• TAEs are also driven more easily at high safety
  factor (not analized)
• PF Coil operational flexibility for AT modes in
  FIRE

46
TF Ripple and Alpha Particle Losses
TF ripple very low in FIRE ?(max) 0.3
(outboard midplane) Alpha particle collisionless
collisional losses 0.3 for reference ELMy
H-mode For AT plasmas alpha losses range from
2-8 depending on Ip and Bt ----gt are Fe inserts
required for AT operation??? Optimize for Bt6.5T
47
Fe Shims for Ripple Reduction for AT Modes in FIRE
TF Coil
Fe Shims
Outer VV
Inner VV
48
PF Coil Capability for AT Modes
AT modes have flattops ranging from 16-50 s

• Advanced tokamak plasmas
• Range of current profiles 0.35 lt li(3) lt 0.55
• Range of pressures 2.50 lt ?N lt 5.0
• Range of flattop flux states chosen to minimize
  heating and depends on flattop time (determined
  by Pfusion)
• Ip limited to 5.5 MA
• Lower li operating space led to redesign of
  divertor coils
• PF1 and PF2 changed to 3 coils and total
  cross-section enlarged
• Presently examining magnet stresses and heating
  for AT scenarios

49
AT Physics Capability on FIRE
Control
Strong plasma shaping and control Pellet
injection, divertor pumping, impurity
injection FWCD (electron heating/CD) on-axis,
ICRF ion heating on/off-axis LHCD (electron
heating/CD) off-axis ECCD (LFS, electron
heating) off-axis, MHD control RWM MHD feedback
control NBI ?? (need to examine for AT
parameters!!) t(flattop)/t(curr diff)
1-5 Diagnostics
MHD J Profile P-profile Rotation
50
Ongoing Work to Establish Advanced Tokamak Regime
for FIRE

• Establish PF Coil operating limits
• Revisit Equilibrium/Stability Analysis
• Use recent GLF23 update in AT scenarios
• LHCD efficiency updates
• EC with FIREs parameters
• Orbit calculations of lost alphas for scenario
  plasmas, Fe shim requirements
• RWM coil design in port plugs and RF ports
• Determine possible impact of n2 RWM on access to
  high ?N
• Examine NBI for FIRE AT parameters