Transcript Document

Advanced Tokamak Research
on Alcator C-Mod:
Towards Profile Control and Long Pulses
Amanda Hubbard,
MIT Plasma Science and Fusion Center
Thanks for input from all of the C-Mod team,
especially C. Fiore, D. Ernst, P. Bonoli, R. Granetz,
R. Parker, E. Marmar
Columbia Plasma Physics Colloquium
November 21st, 2003
Advanced Tokamak Research on C-Mod
• Introduction
– What do we mean by “Advanced Tokamak”?
– Why do we want it?
– Overview of C-Mod tokamak, AT tools.
• Transport and Profile Control;
– recent ITB results and near-term plans.
• Lower Hybrid Current Drive
– Hardware and predicted current drive.
• Modelling of AT target scenarios,
– MHD stability limits.
“Advanced Tokamak”
•
Attractiveness of the tokamak as an ultimate fusion reactor
increases if it is steady state and produces power at lower cost.
•
Steady state implies non-inductive current drive.
•
Lower cost implies much of the current drive is self-generated, ie.
High bootstrap fraction.
(external drive is expensive, in $ and Watts!).
Also want high confinement, and high b for lower cost (all are
inter-related).
•
•
Research aimed at demonstrating these features and optimizing
the tokamak configuration is a key part of the US fusion
development plan.
Will they be achievable in ITER? In CTF and/or DEMO?
Alcator C-Mod Tokamak
Alcator C-Mod Plasma
•
•
•
•
R=0.68 m
a=0.21 m
B<8T
Ip < 1.7 MA
(now 2 MA cap)
• PRF < 6 MW
• ne~1020-1021 m-3
• T ~1-6 keV
Profile Control Tools on C-Mod
“The crucial distinguishing feature of an Advanced Tokamak over a
conventional tokamak is …the use of active control of the
current or shear profile, and of the pressure profile or transport
characteristics” (AT Workshop, GA, 1999)
Tools available or under development:
• Current profile:
– Lower Hybrid Current Drive. (2004. Upgrade 2005-6).
4 MW, 4.6 GHz, 2 launchers with independent phasing, N//.
– Mode Conversion Current Drive. (on or off axis, tests started in 2003)
– Bootstrap current drive via pressure profile control.
•
Density profile.
– Control of core transport, peaking.
– Cryopump planned to control the edge source. (2005)
– D2 and Lithium pellet injectors.
•
Temperature Profiles
•
– 8 MW ICRH, 40-80 MHz, 2 independently variable deposition locations.
– 4 MW LHCD.
– Control of core transport via RF deposition, magnetic shear.
Shear Flow - MC flow drive; preliminary expts started in 2003.
C-Mod is in a unique and Burning Plasma
relevant regime
•
•
•
•
Most AT expts have Ti > Te, te-i > tE and use NBI for core fuelling and
rotation drive in Internal Transport Barriers.
Reactor scenarios have
te-i << tE (Te >Ti), no core fuelling, mainly RF heating and CD.
Due to high density of C-Mod, electrons and ions are almost always
strongly coupled.
C-Mod only uses RF heating and fuelling, no core fuelling or rotation
drive. [Interestingly, intrinsic rotation is routinely seen.]
– These features are particularly relevant for core transport barrier
studies.
•
Reactor relevant divertor flux; was up to 0.5 GW/m2 with 3-4 MW
ICRH.
•
C-Mod can test feasibility of advanced scenarios with all these
features simultaneously.
Normalized pulse length exceeds that
of all other present divertor tokamaks
Assumed:
Zeff=2,
Te=6 keV (ITER
19 keV).
•
•
•
Allows us to be sure that current is fully diffused for most of the
discharge; not always the case even in fully non-inductive discharges.
Enables the study of transport and stability limits in steady conditions.
C-Mod has already run 3 s pulses, has 5 s TF capability; some times
on plot are machine upgrades.
Main goals of the AT physics
program on C-Mod
1. Demonstrate and model current profile control using LH and
ICRF waves, at high densities (>1020 m-3).
2. Understanding, control and sustainment of Internal Transport
Barriers, with coupled ions and electrons, te-i << tE (Te~Ti ) and
without momentum input (RF only).
3. Achieve full non-inductive current drive (70% bootstrap) and
extend pulse length to near steady state (5 sec, 4-6 tCR)
- divertor power handling and wall particle issues.
4. Attain and optimize no-wall b limits, with bn of at least 3, and
explore means of achieving higher values
AT program integrates all physics areas (RF, transport, divertor, MHD),
involves the whole C-Mod team.
This talk will focus on 2 and 1, touch on 3, 4.
Control of transport, temperature
and density profiles:
Internal Transport Barriers
‘Transport barrier’ is a local region
of much reduced c, D
nT
Q
• With constant heat flux Q,
reduced c is seen by steeper
‘Internal Transport
gradients.
Barrier’
(in ci, ce and/or D)
• Can occur in either edge or
core.
– Edge barrier, or
‘pedestal’ in T, n first seen
Edge Transport
Barrier
on ASDEX in 1982.
Height
‘pedestal’
– Called ‘High Confinement’
Width D
‘H-mode’.
r
• Core ‘internal’ barriers
(‘ITB’s) first produced with
pellets (on Alcator C), now
‘insulating layers’
many different methods,
names.
Internal Transport Barriers are produced
and controlled by varying heating profile
• ITB’s routinely are triggered
on C-Mod by off-axis
ICRH, at r/a ~0.5. (high or
low-field side)
• These core barriers always
co-exist with edge pedestal
(EDA H-mode.)
i.e. Double-Barrier Regime
• Reversed shear not needed
(most discharges are
sawtoothing).
C. Fiore, APS Invited 2003
Off-axis ICRF heating at
r/a=0.5 on the low field side
ITBs also arise in Ohmic EDA H-mode
plasmas
 Once an EDA H-mode is
established in an ohmic plasma
and maintained for several
energy confinement times, an
ITB will arise and persist until a
terminating event, usually an H
to L back transition.
 Characteristics of the ITBs in
ohmic plasmas are
indistinguishable from those in
off-axis ICRF heated discharge.
EDA ohmic H-mode, no ICRF during
ITB development
Density profiles from Thomson scattering show
peaking from ITB in good agreement with visible
bremsstrahlung data (used for higher res’n)
The ratio of the vb data to the
density from Thomson
scattering is sqrt(Zeff)
Off-axis ICRF heating at r/a=0.5 on
the low field side
Barrier location depends on B,
not on heating location
• Barriers have been triggered
with both 70 and 80 MHz
IRCH, on high or low field
side.
• Same threshold in deposition
location (rdep/a > 0.5) in all
cases.
• Also a (weaker, less clear)
dependence of the barrier
location on q95.
• This year, plan to use 50
MHz ICRF, 3.2 T.
Will ITB get even wider?
Temperature increases in ITB, but does
not peak significantly
Central Te rises as ITB peaks
Barrier foot location
Profile changes seen in the
density are not reflected in the
electron temperature profiles.
Operation at higher magnetic fields
allows measurement of Te from ECE
emission; this is cut off during ITBs
obtained at lower field.
The power deposition profile is hollow when
ITBs develop, in both ohmic and ICRH cases
Inside ITB location, net P(r)
is quite similar in the two
cases.
This may be the common
feature behind the barrier
phenomenon.
ceff reaches neoclassical value as the ITB
evolves
Although Te is not strongly
peaked, heat transport is
clearly reduced.
Off-axis ICRF heating at
r/a=0.5 on the high field
side
EDA Ohmic H-mode, no
ICRF during ITB
development
Sawtooth Perturbation Propagation is Modified
•Pulse is significantly delayed
compared to EDA case between
rtangent= 9.8 cm and 11.0 cm.
chp Profile Appears to have a Narrow Region of
Reduced Transport, near the barrier foot
3 Te 1   hp Te 
ne

rne c

2
t
r r 
r 
Te (0,t)  Te ( 2 * rmix ,t)
Te (a,t )  0
• Barrier region appears to be
narrower than 2 cm.
• Indicates electron thermal
barrier.
On-axis heating controls the level of
core transport
• Control of ITB strength is key to its
usefulness in AT (or other)
regimes.
• Typical problem with ITBs, on all
experiments, is that particle,
impurity, and/or heat confinement
are too good.
• One of two problems usually
arises:
•
– Pressure gradient rises,
exceeds MHD limits; core
collapse and/or disruption.
– Core Zeff and radiation rise
•
continuously, get radiative
collapse.
These effects can be avoided
on C-Mod using a
combination of on, off-axis
heating (ICRF at 70, 80 MHz).
ITBs sustained in near-steady
state (> 10 tE), limited only by
discharge length.
Density rise can be controlled by central
ICRF heating after the ITB is established
Nearly identical Ohmic H-mode ITBs
were induced for 5 shots. Increasing
amounts of central rf power were
added after the ITB formation.
Central rf power > 0.2 MW
suppresses the peaking of the
density. Highest density reached
decreases with incremental
power.
Similar results were obtained
with RF-induced ITBs.
On-axis heating also controls
impurities
• Zeff does tend to peak
somewhat during ITB
formation.
• Can be kept below 2, and
steady, with central ICRH.
• Prad also stops rising.
Off-axis ICRF heating at r/a=0.5 on
the low field side
The toroidal rotation decreases as the ITB
forms; causal role unclear
The co-going central toroidal
rotation (from argon impurity)
declines, reverses as ITB
develops and the density rises.
Vf profile becomes hollow.
When central ICRF is added
the rotation increases and
becomes positive (co-going)
again
Physical Mechanism for ITB
formation, control
• The peaking, transport suppression with hollow
power deposition was unexpected.
• Some similarity with ASDEX upgrade results, though
the profile break-in-slope, transport bifurcation seem
stronger on C-Mod.
• Recent microstability analysis with GS2 is giving
insight into the physical mechanisms of barrier
formation and stabilization.
• Collaborative effort involving Martha Redi (PPPL),
and Darin Ernst (MIT theory).
At the ITB onset time, long wavelength drift modes (ki = 0.1
to 0.8) are not strongly growing at or to the inside of the ITB
location
TEM: not strongly growing at
& within ITB
ITG: stable at & within ITB,
unstable outside ITB
Once D is reduced, Ware
pinch is sufficient to peak ne.
Calculation is at the onset
of the ITB during off-axis
ICRF heating, before the
density becomes peaked.
M. Redi, EPS 03, APS 03
With on-axis heating (no ITB), sawteeth are larger
and Te more peaked than with ohmic or off-axis RF
EXB shearing is comparable to growth rates at
onset; lower in core after ITB develops.
ITB forming
ne rising
On-axis RF
max: maximum linear
growth rate from linear
GS2 simulation at r/a=0.4
(Ernst, APS invited 2003)
w EXB: dominated by the
pressure gradient term late
in the ITB phase;
Profiles from this late
phase, with on-axis RF, on
next slide.
Replace w Ernst slide 7
On-axis heating increases Te, GTEM
• Non-linear GS2 runs show
increase in diffusivity is due to
TEM. DTEM~ T 3/2.
– Collisionality, LT effects are
less important in our regime.
• Higher particle flux GTEM
balances density increase due
to Ware pinch.
– Leads to a new particle
equilibrium, dependent on T.
• Modelling has shown a new
non-linear upshift of critical a/Ln
for TEM.
Ernst, APS 2003,
paper submitted to PoP.
Te
Control of core transport:
Future Research
•
•
•
•
•
•
•
Study of ITBs generated by optimized heating deposition is interesting
for physics understanding, development and validation of models in
regime of interest to ITER.
– Coupled ions and electrons, Ti~ Te, low Vtor, high ne , ,peaked
without core fuelling.
Possible application as a core transport control tool.
Obviously, a regime which depends on low central heating is not
directly applicable in a burning plasma heated by alphas.
In future, we intend to create and study ITB’s with flat or reversed
shear, as on other tokamaks; expect easier access and better
compatibility with central heating, T peaking.
Current profile control with Lower Hybrid Current Drive.
Also, will test effect of flow drive with mode converted ICRF
Aim to expand barriers, increase bootstrap current for use in Advanced
Tokamak scenarios.
Control of current profile for long
pulses:
Lower Hybrid Current Drive
C-Mod Lower Hybrid Current Drive system
• LHCD is a highly efficient
means of driving localized
current.
2004
2005-6
Frequency 4.6 GHz
4.6 GHz
• Used on Alcator C, compatible
with high densities, fields.
Source
Power
3 MW
4 MW
• New C-Mod system is a
collaboration between MIT,
PPPL, nearly complete.
Antenna
1 grille
(4x24
guides)
2 grilles
(4x24
guides
each)
Main aims are
1) current profile control, j(r) well off-axis.
2) extend pulse length to > 5 secs ( few tCR).
“Grill” launchers designed for well controlled
spectrum
•
Each antenna will have flexible N//, variable over range 2-4.
•
Variable between or during discharges using phase shifters.
•
2 launchers can have different spectra.
•
Allows us to tailor spectrum for desired wave accessibility
(depending on n(r), B), and to control deposition and current
drive profiles, including CD far off axis.
The LH Antenna Hardware
Coupler (with vacuum windows)
Forward waveguide assembly
Rear waveguide assembly and
splitter network
Mainly PPPL design, fabrication
LH Sources
• 16 klystrons available from
Alc. C, PBX-M experiments.
• 4.6 GHz, 250 kHz each.
• 12 klystrons for Phase I are
installed, tested in C-Mod
cell.
• Have been used for high
power testing of launcher,
waveguide components.
• Somewhat ‘overpowered’ for
1 launcher; expect to couple
~2 MW for short pulses, 1.5
MW for long pulses.
• Second launcher (16
klystrons) will be needed
to couple 3 MW for l5 sec
pulses.
• Will also allow compound
spectra for better j(r) control.
Near term plans: j(r) control
•
•
•
AT thrust now developing target
2003 Example: H-mode.
plasmas for good LHCD efficiency.
3.5 MW ICRF, 5.4 T, 0.6 MA, H89=2.1
Want high Te, med ne (1-2x1020m-3).
Modelling CD with ACCOME
ne
2x1020
(may underestimate ILH), also CQL3D,
m-3
LSC
ILH=131 kA
IBS=243 kA
62% non-ind
5 keV
Te
Longer term: Integrated Advanced
Tokamak Goals
•
A successful AT demonstration must combine all of the
control tools and physics/technology areas.
– Eg. LHCD and high bootstrap and high b and long-pulse
divertor.
– Integration and parameter optimization are important.
For example, tradeoffs necessary in Ip, density,
confinement in designing LH targets.
•
Five-year goal is:
– Fully non-inductive current drive of 0.85 MA, from
LHCD plus bootstrap current,
– bN=3.0 (or higher), for
– 5 second pulse length (~6 tCR at 5 keV).
– Core transport barrier with H89P > 2.5.
•
Expect continuous progress in performance and scientific
understanding along the way.
Integrated Scenario Modelling.
Modelling is critical to assess wave damping and CD, confinement
and stability in various regimes, and guide discharge development
toward more optimal scenarios.
Currently available models include:
– ACCOME: LH, ICRH and bootstrap Ip. Consistent MHD equilibria.
No transport; temperature, density specified.
– CQL3D (R. Harvey): Self-consistent 2-D velocity space Fokker-Planck.
(20-30% higher LHCD than ACCOME).
– TRANSP: Time dependent, predictive or simulation mode.
+ LSC for LHCD. Limitations in reverse N//, 1 poloidal launch point.
– TORIC: Full wave field solver for ICRH, mode conversion (and LH).
– PEST-2, Keldysh/KINX code, MARS: MHD stability analysis
– GS2: Microstability analysis
Many Plans for new and improved AT-related modelling in RF,
TRANSPORT and MHD areas
Collaborations with PPPL, Cadarache, Lodestar, ORNL, CompX and
others (hopefully including Columbia!)
Example of an AT target scenario
meeting performance target.
•
One of many optimized
scenarios modelled with
ACCOME.
– Ip=860 kA, non-inductive.
– ILH=240 kA
– IBS=600 kA (70%)
– bN=2.9
•
Double transport barrier
•
•
•
•
•
•
BT=4 T
ICRH: 5 MW
LHCD: 3 MW, N//0=3
ne(0)= 1.8e20 m-3
Te(0)=6.5 keV (H=2.5)
Scenarios without barrier, or
only an ITB, have similar
performance.
P. Bonoli, Nucl. Fus. 20(6) 2000.
MHD Stability of non-inductive plasmas
•
•
•
•
•
•
Present plasmas are kink
stable.
Expect core MHD stability to be
more important for C-Mod as
power, b raised.
Ideal no-wall limit bn~3.
– With optimized p(r), j(r).
– Strong shaping.
With no wall, bn,crit~3.7, set by
n=1 external kink.
With rwall = 1.2 a, bn,crit~4.5-5.,
set by n =3 external mode.
Lower bn,crit (~2.1-2.5) found
with lower  and/or stronger
profile peaking;
We expect to encounter and
study kink modes early in AT
program.
• Stability assessed with
PEST-II, using ACCOME
equilibrium (ITB, double
barrier).
• Optimal stability with k=1.8,
=0.7, p0/pav=3.
MHD Stability: research plans.
•
•
•
•
New antennas for active core MHD
spectroscopy can measure linear growth
rates.
– Plan to feedback on power, profiles
optimize shape to avoid limit.
Study ELM, core MHD interaction.
Try stabilization of NTMs using LHCD and/or
MCCD.
To further increase bN, which should be
possible given high confinement and planned
input power, need to assess what it would
take to provide a stabilizing shell.
– Effect of our current wall, including ICRH, LH
antennas.
– How large a fraction of toroidal circumference
would be necessary?
– Can feedback coils for RMW stabilization be
incorporated?
TRANSP simulation of full non-inductive CD
with ITB, L-mode edge, with LHCD.
•
•
•
Input ne() with barrier at r/a=0.5.
Input c() taken from analysis of C-Mod ITB experiments.
Te, current profiles evolve in time.
PICRF=3 MW
PLH=1.65 MW
N//=2.75
J. Liptac,
APS 2002
Other AT control tools, upgrades
• Mode Conversion of ICRH waves, to IBW and ICW.
– Exciting new measurements of MC waves. [Invited talk by Y.
Lin, APS2003]
– Potential application for current drive, on or off-axis, and for
flow drive. Can we generate flow shear to control transport?
– Preliminary expts in summer 2003 give indications of flow
drive; more planned soon!
• Density control
– Very important for LH accessibility, efficiency, profile.
– Adding new cryopump in 2005.
• Divertor upgrades to handle long pulses, higher heat loads.
No time to present details today
Insert Yijun slides
Overview of 5-year schedule for
Advanced Tokamak Program
Conclusions
•
Advanced Tokamak research is an increasingly important part of the CMod program.
– Focuses on unique features of RF control of current, transport and
pressure profiles in high density regime, for t >> tCR,
•
Internal transport barriers are routinely produced and controlled by
optimizing heating profiles - without momentum input or reversed shear,
and with Te ~Ti.
Modeling is leading to physical understanding.
– ITG growth rate is low at onset time, neoclassical pinch peaks ne.
– Density peaking further stabilizes turbulence, reduces D, cto ~
neoclassical.
– TEM grows with density peaking, until it balances inward flux.
•
•
•
Initial LHCD system is nearly complete
– Installation Jan ’04, 1st experiments April 2004.
– Strong off-axis current drive is predicted with experimentally
demonstrated hot H-mode target plasmas.
Research program leads progressively to a non-inductive, steady
state, high confinement advanced tokamak demonstration in a
unique regime highly relevant to ITER and the steps beyond.