PVR Magnets

1
FIRE Physics Basis(detailed version)
C. Kessel for the FIRE Team Princeton Plasma
Physics Laboratory FIRE Physics Validation
Review March 30-31, 2004 Germantown, MD
2
FIRE Description
R 2.14 m, a 0.595 m, ?x 2.0, ?x 0.7, Pfus
150 MW

• AT-Mode
• IP 4.5 MA
• BT 6.5 T
• ?N 4.2
• ? 4.7
• ?P 2.35
• ???? 0.21
• q(0) 4.0
• q95, qmin 4.0,2.7
• li(1,3) 0.52,0.45
• Te,i(0) 15 keV
• ?Te,i? 6.8 keV
• n20(0) 4.4
• n(0)/?n? 1.4
• p(0)/?p? 2.5
• n/nGr 0.85
• Zeff 2.2
• fbs 0.78
• Q 5

• H-mode
• IP 7.7 MA
• BT 10 T
• ?N 1.80
• ? 2.4
• ?P 0.85
• ???? 0.075
• q(0) lt 1.0
• q95 3.1
• li(1,3) 0.85,0.66
• Te,i(0) 15 keV
• ?Te,i? 6.7 keV
• n20(0) 5.3
• n(0)/?n? 1.15
• p(0)/?p? 2.4
• n/nGr 0.72
• Zeff 1.4
• fbs 0.2
• Q 12

VV
baffle
divertor
passive plate
plasma
port
3
FIRE Magnet Layout
Error field correction coils
PF4
PF1,2,3
TF Coil
CS3
PF5
Fe shims
CS2
CS1
Fast vertical and radial position control coil
RWM feedback coil
4
Toroidal Field Coils
16 TF coils BeCu inboard legs OFHC Cu outboard
legs Coil stress and heating limit TF pulse
length (factor of 1.18 over ?
allowable) H-mode BT 10 T and Pfus 150
MW ----gt 20 s flattop AT-mode BT 6.5 T and
Pfus 150 MW ----gt 48 s flattop Maximum TF
ripple at Ra is 0.3 0.3 ?-particle loss
H-mode 8 ?-particle loss AT-mode Expect to use
Fe shims for AT-mode
5
Poloidal Field Coils
Center Stack 1, 2UL, 3UL PF1,2,3,4,5 UL All CS
and PF coils are CuCrZr H-mode Fiducial
equilibria in discharge IM, SOD, SOH, SOB, EOB,
EOH, EOD Flexibility of PF coils 0.55 li(3)
0.85 (SOB,EOB) 0.85 li(3) 1.15
(SOH,EOH) ?ref-5 ?(Wb) ?ref5 1.5 ?N
3.0 Full operating space available within stress
(1.3 margin) and heating allowables, except at
EOB, li0.85 where ? ?ref-2 Rampup consumes
40 V-s Flattop consumes 3 V-s
PF1,2,3
PF4
CS3
CS2
CS1
PF5
6
Poloidal Field Coils
AT-mode Fiducial equilibria in discharge IM, SOD,
SOH, SOF, SOB, EOB, EOH, EOD Flexibility of PF
coils 0.35 li(3) 0.65 (SOB EOB) 2.5 ?N
5.0 7.5 ?flattop(Wb) 17.5 Full operating
space is available for Ip 5.0 MA PF coils can
provide pulse length limitation -----gt 41 s for
access to op. space at Ip 4.5 MA, and scales
with Ip, li, ?p, and ? Inductive
non-inductive rampup consumes 19-22 V-s, final
state can be optimized Plasma current 100
non-inductive in flattop
shape control feedback points
7
Poloidal Field Coils
TSC simulations Free-boundary calculations with
heating, CD, bootstrap current, energy and
current transport, impurities, PF coils,
structure and feedback systems, etc. ---gt check
of equilibrium coil currents ---gt Volt-second
consumption ---gt Feedback control of vertical
position, radial position, plasma current and
shape
8
Vertical Stability
Design passive structures to slow vertical
instability for feedback control and provide a
stability factor fs gt 1.2 Passive stabilizers
are 2.5 cm thick Cu, toroidally continuous on
upper outboard and inboard sides For most
unstable plasmas (full elongation and low
pressure), over the range 0.7 lt li(3) lt 1.1, the
stability factor is 1.3 lt fs lt 1.13 and growth
time is 43 lt tg(ms) lt 19
Passive stabilizers
Cladding (ports provide poloidal cuts)
Cladding (large number of poloidal cuts)
9
Internal Control Coils
8 OFHC Cu coils (2nd redundant coils) above and
below the midplane Fast vertical position
feedback control ?ZRMS 1 cm, 65-90 kA-turn,
50-75 V/turn ------gt 7-14 MVA (peak) Fast radial
position feedback control (antenna
coupling) Analysis not completed, assuming I and
V similar to vertical control Fast radial
feedback is coupled with slower outer PF shape
control These coils also used in startup to
tailor field null
10
Resistive Wall Mode (RWM) Coils
ICRF Port Plug
DIII-D experience Modes are detectable at the
level of 1G The C-coils can produce about 50
times this field The necessary frequency depends
on the wall time for the n1 mode (which is 5 ms
in DIII-D) and they have ??wall 3 FIRE
projection FIRE has approximately 3-4 times the
DIII-D plasma current, so we might be able to
measure down to 3-4 G If we try to guarantee at
least 20 times this value from the feedback
coils, we must produce 60-80 G at the
plasma These fields require approximately I
f(d,Z,?)Br/?o 5-6.5 kA Assume we also require
??wall 3 Required voltage would go as V
3?o(2d2Z)NI/??wall 0.25 V/turn
RWM Coil
11
Error Field Correction Coils
Static or slow dynamic Cu coils Located outside
TF and PF coils Compensating TF and PF
coil/lead/etc. misalignments and other under
field conditions These coils are NOT used for
RWM feedback Extrapolated threshold to induce
locked modes 1 ? 10-4 T (very
uncertain!) Correction coils should be capable
of reducing (m1,n1), (2,1), and (3,1) error
fields, simultaneously And provide factor of 5
reduction in net error field Br2,1net
No analysis performed
ITER Error Coils
3 distributed coils provides poloidal mode
control allowing multiple (m,n1) suppression
Recent C-Mod data shows that applied Br2,1 of
6?10-4 T removed mode-locking -----gt Important
since C-Mod does NOT have external rotation source
12
ICRF Heating and CD
Frequency range 70-115 MHz 2 strap antennas 4
ports, total power 20 MW H-mode BT 10 T,
minority He3 and 2T at 100 MHz Frequency range
allows heating at a/2 on HFS and LFS AT-mode BT
6.5 T, ion heating at minority H and 2D at 100
MHz Frequency range allows ion heating at a/2 on
HFS and LFS Electron heating/CD at 70-75 MHz CD
efficiency ?20 0.14-0.21 A/W-m2
SPRUCE analysis nHe3/ne2 PICRF11.5 MW ?100
MHz THe3(0)10.2 keV PHe360 PT10 PD2 Pe26
13
ICRF Heating and CD
Vacuum Toroidal Field Resonances
BT 10 T
BT 6.5 T
Be
Be
14
ICRF Heating and CD
Want to reduce power required to drive on-axis
current 2 strap antenna and port geometry
provides only 40 of ICRF power in good CD part
of the spectrum 4 strap antenna can provide 60
of power in good CD part of spectrum Expanding
antenna cross-section and going to 4 straps
reaches 80 in good CD part of spectrum
15
ICRF Heating and CD
AORSA full wave analysis continues including fast
alpha and Be impurity effects
75 MHz
70 MHz
Pe0.44 PT0.15 PBe0.30
Pe0.65 PT0.32 PBe0.0
?200.14
?200.17
16
Lower Hybrid Current Drive
f gt 2?fLH RF power flux is 53 MW/m2 Need 0.57 m2
per waveguide for 30 MW Each waveguide is 5.7
cm(tall) ? 0.65 cm Have 1500 waveguides
Frequency 5 GHz Spectrum n 1.8-2.5, ?n
0.3 Power of 30 MW, in 2 ports Upgrade to
baseline design H-mode Used for NTM control for
BT 10 T Used for non-inductive CD for hybrid
discharges AT-mode Used for bulk CD for BT 6.5
T CD efficiency ?20 0.16 A/W-m2 at 6.5 T
(30-50 higher from 2D FP calcs.) Used for NTM
control
17
Lower Hybrid Current Drive
ACCOME
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
improves CD efficiency, 30-50 increase, but
right now.. BT 8.5T ----gt 0.25 A/W-m2 BT
6.5T ----gt 0.16 A/W-m2 Benchmarks with ACCOME,
CURRAY and LSC 3.7 GHz, 750 kW, 1000s sources
available ITER estimate for 5 GHz, 1.0 MW, CW
sources was 1.15 euro/watt
TSC-LSC
18
Electron Cyclotron
Frequency of 170 GHz to utilize ITER RD LFS,
O-mode, fundamental FIRE has high density and
high field Cutoff of EC when ?
?pe AT-mode Lower BT 6.5 T LFS deposition
implies trapping reduction of CD, however, Ohkawa
effect provides more CD than standard EC Current
required, scaled as Ip??N2 from DIII-D and
ASDEX-U expts for (3,2) mode ----gt drive 200 kA
to suppress from saturated state ----gt requires
100 MW!
Rays are bent as they approach ? ?pe Rays are
launched with toroidal directionality for CD
??ce170 GHz
?pe?ce
19
Electron Cyclotron
r/a(qmin) 0.8 r/a(3,1) 0.87-0.93 Does (3,1)
require less current than (3,2)? Local ?, ?,
Rem effects so close to plasma edge? 170 GHz may
be adequate, but 200 GHz is better fit for FIRE
parameters
J. Decker, MIT
145?155 GHz -30o?L-10o midplane launch 10 kA
of current for 5 MW of injected power
Ro
Roa
Bt6.5 T
?
170 GHz
fce182
fce142
qmin
Bt7.5 T
?
?
(3,1)
fce210
200 GHz
fce164
Bt8.5 T
?
fce190
fce238
20
Neutral Beam Injection (Difficult)
Rtan 0 m 16 TF Coils Need 1 MeV to get 50 of
power inside a/2 Rtan 0.75 m and higher Must
go to 12 TF coils, pinwheel ports, and Fe
inserts Need gt 1 MeV to get 50 of power inside
a/2 Plasma rotation for Rtan 1.7 m Assumed
120 keV 8 MW ----gt deposited r/a gt
0.65 Dominated by j?B rotation giving v/vAlfven
0.5
Rtan 1.7 m
Rtan 0.75 m
Rtan 0 m
21
Power Handling
First wall Surface heat flux Plasma radiation,
Qmax P? Paux Volumetric heating Nuclear
heating, qmax qpeak(Z0) VV, Cladding, Tiles,
Magnets. Volumetric heating Nuclear heating,
qmax qpeak(Z0) Divertor Surface heat
flux Particle heat flux, Qmax
PSOL/Adiv(part) Radiation heat flux, Qmax
PSOL/Adiv(rad) Volumetric heating Nuclear
heating, qmax qpeak(divertor)
VV
Clad
Tile
plasma
22
Power Handling
Pulse length limitations VV nuclear heating
(stress limit), 4875 MW-s -----gt Pfus
(qVVnuclear) FW Be coating temperature, 600oC
-----gt QFW Pfus (qBenuclear) TF coil heating,
373oK -----gt BT Pfus (qCunuclear) PF Coil
heating-AT-mode, 373oK -----gt Ip, li, ?p, and ?
(not limiting) Component limitations Particle
power to outboard divertor lt 28 MW Radiated
power on (innerouter) divertor/baffle lt 6-8 MW/m2
23
Power Handling/Operating Space
FIRE H-mode Operating Space ?N limited by NTM or
ideal MHD with NTM suppression -----gt maximum
Pfus Higher radiated power in the divertor
allows more operating space, mainly at higher
?N -----gt maximum Pfus Majority of operating
space limited by TF coil flattop -----gt ?flattop
20 s High Q (15-30) operation obtained with
Low impurity content (1-2 Be) Highest H98
(1.03-1.1) Highest n/nGr (0.7-1.0) Highest
n(0)/?n? (1.25)
H98(y,2) 1.1
24
Power Handling/Operating Space
FIRE AT-mode Operating Space ?N is limited by
ideal MHD w/wo RWM feedback -----gt maximum
Pfus Higher radiated power in the divertor
allows more operating space, mainly at higher
?N -----gt maximum Pfus Majority of operating
space limited by VV nuclear heating -----gt
?flattop 20-50 s Design solutions to improve
VV nuclear heating limit, could reach PF coil
limit, function of Ip Number of current
diffusion times accessible is reduced as ?N, BT,
Q increase
H98(y,2) 2.0
25
Particle Fueling/Pumping
Require 1-2?1021 tritons/s for FIRE
H-mode ---gt 0.1-0.2 g T injected per shot (20
s) ---gt 5 of injected tritium consumed HFS
launch, limited to 125 m/s (test actually
performed at ORNL to find pellet speed
limit) LFS and VL can reach much higher
velocities VL is at major radius, therefore not
expected to provide improvement over LFS
26
Particle Fueling/Pumping
Pumpdown and vessel bakeout utilize midplane
pump, to provide minimum of 2000 l/s to reach
10-7 or less base pressure
16 cryocondensation/diffusion pumps, 8 above and
8 below midplane, every other port Backed by
turbo/drag pumps H2O pumped on 1 m long 30oK
entrance duct H and impurities pumped by
cryocondensation, liquid He He pumped by
turbo/drag pump located outside bio-shield,
viscous drag compression (200 l/s
conductance) Cooling requirement for 16
cryopumps at 200 torr-l/s and nuclear heating
(0.03 W/cm3) is 48 W, and liquid He flow rate is
64 l/hour for all 16 pumps Regeneration is done
into the turbo/drag pumps
27
Particle Fueling/Pumping
WHIST simulation of FIRE H-mode discharge
(Houlberg) Assume uniform pellet
deposition Obtains some density peaking with
sufficient pumping
V 125 m/s Parks, 2003
28
MHD Stability
Sawtooth H-mode Unstable to internal kink,
r/a(q1) 0.35 m -----gt coupling to other global
modes? Porcelli sawtooth model (?WMHD ?WKO
?Wfast), incorporated into TSC indicates effect
on fusion performance is weak Pedestal/bootstrap
broadens j profile Rapid reheat of sawtooth
volume ?-particles providing stabilization Compl
ete stabilization would require RFCD since FIRE
does not have high energy minority species The
q1 surface can be removed from the plasma by
1.2 MA off-axis CD Reduction of Ip to 6.0 MA
29
MHD Stability
Neoclassical Tearing Modes H-mode Stable or
unstable? Sawteeth and ELMs are expected to be
present and can drive NTMs Typical operating
point is at low ?N and ?P Can lower ?N further if
near threshold Lower Hybrid CD at the rational
surfaces Compass-D demonstrated LH
stabilization Analysis by Pletzer and Perkins
showed stabilization was feasible (PEST3) Lowers
Q(Pfus/Paux) EC methods require high
frequencies at FIRE field and densities ----gt
280 GHz
TSC-LSC
(3,2) surface 12.5 MW 0.65 MA n/nGr 0.4 Q
6.8
30
MHD Stability
Current profile modification
ususal NTMs
ASDEX-U
FIR-NTMs
S. Günter et al., PRL 2001
31
FIRE MHD Stability
Current profile modification
(LHCD ctr-CD in start-up phase)
JET
Despite (3,2) NTM excellent confinement
H98y1.4, ?N 3.3
32
MHD Stability
Ideal MHD Stability H-mode n1 external kink and
n8 ballooning modes H-mode Stable without a
wall/feedback Under various profile conditions ?N
3 ballooning unstable in pedestal region
depending on pedestal width and
magnitude Intermediate n peeling/ballooning modes
H-mode Unstable, primary candidate for ELMs Type
I ELMs are divertor lifetime limiting, must
access Type II, III Ploss/PLH 1.0-1.6 in
flattop, not gt 2 like many present
experiments FIRE has high triangularity (?x
0.7) in Double Null and high density Active
methods to reduce ?WELM include pellets,
impurities, ergodization,
Self-consistent ohmic/bootstrap equilibria
33
MHD Stability
Neoclassical Tearing Modes AT-mode Unstable or
Stable? q(?) gt 2 everywhere, so rational
surfaces are (3,1), (5,2), (7,3),
(7,2) r/a(qmin) 0.8 r/a(3,1)
0.87-0.93 Local ?, ?, Rem effects so close to
plasma edge? L-mode or H-mode conditions Examinin
g EC stabilization at 170 GHz LFS absorption,
Ohkawa CD dominates Scaling from (3,2) expts
indicates high power ----gt early detection
required LH using two spectra, one for bulk CD
and other for NTM suppression
34
MHD Stability

• Ideal MHD Stability AT-mode
• n 1, 2, and 3external kink and n 8 ballooning
  modes
• n 1 stable without a wall/feedback for ?N lt
  2.5-2.8
• n 2 and 3 have higher limits without a
  wall/feedback
• Ballooning stable up to ?N lt 6.0, unstable in
  pedestal region of H-mode edge plasmas.
• RWM stabilization with feedback coils, VALEN
  analysis indicates 80-90 of ideal with wall
  limit for n1
• n 1 stable with wall/feedback to ?Ns around
  5.0-6.0
• n 2 and 3 appear to have lower ?N limits in
  presence of wall, possibly blocking access to n
  1 limits
• Intermediate n peeling/ballooning modes
• Unstable under H-mode edge conditions

Bialek, Columbia Univ.
Growth Rate, /s
?N4.2
?N
35
MHD Stability
TSC-LSC Simulation Equilibria
JSOLVER Equilibria
36
MHD Stability
Other MHD Issues H-mode and AT-mode Alfven
eigenmodes and energetic particle modes Snowmass
assessment indicated stable for H-mode, although
access to shorter pulse high Pfus plasmas should
destabilize AT-mode not analyzed Error fields
from coil misalignments, etc. ----gt install Cu
window coils outside TF coil, stationary to slow
response FIRE does not have an external source
of rotation Transport, sheared rotation Resistive
instabilities, sheared to bulk rotation RWM, bulk
rotation Plasma self-rotation (C-Mod), is it
sufficient for some stabilization
37
Disruption Modeling
38
Disruption Modeling
Experimental database used to project for
FIRE Thermal quench time Ihalo/Ip ? TPF dIp/dt
rates for current quench
39
Disruption Modeling
TSC simulation of disruption Critical structures
modeled VVs (SS), passive plates (Cu), cladding
(Cu), divertor (Cu), baffle (Cu), midplane port
regions (SS) Zero-net current constraint on
divertor, baffle, midplane port regions Provide
poloidal current paths for halo/structure currents
40
Disruption Modeling
Rapidly drop pressure over 0.2 ms Use
hyper-resistivity to broaden current Plasma
temperature drops to 15-30 eV, current is shared
with halo region depending on Thalo (2-7 eV) and
halo width Ip drops at rate determined largely
by Thalo Plasma shrinks rapidly, then plasma is
converted to a circuit
Thermal quench, ?t 0.2 ms
Diamagnetic flux
Ip drop, -2.9 MA/ms
41
Disruption Modeling
TSC simulation produces for Engr.
analysis Toroidal structure currents, fields and
forces Poloidal structure current, fields, and
forces Plasma toroidal currents on a
grid Halo/poloidal plasma currents at structure
interfaces Global plasma and PF coil data
42
Disruption Mitigation
Utilize fueling technology to mitigate
electromagnetic effects of disruptions Massive
gas puff into DIII-D ----gt peak halo currents
reduced by 50 by He and D puffing, and toroidal
asymmetry reduced Ne, Ar, and CH4 pellets into
DIII-D ----gt peak halo currents reduced by 50
with Ne and Ar pellets, and toroidal asymmetry
reduced from 3 to 1.1 Cryogenic liquid jet being
developed Low Z impurity pellets (LiD) if
runaway electrons not an issue Snowmass
assessment indicated large radiated power to FW
could cause Be melting
DIII-D Massive Gas Puff System
43
FIRE Transport and Confinement
Energy Confinement Database ?E98(y,2) 0.144
M0.19 Ip0.93 BT0.15 R1.97 ?0.58 n200.41
?0.78 P-0.69 (m, MA, T, MW) ?p/?E 5 Zeff
1.2-2.2 (fBe 1-3, fAr 0-0.3) Pedestal
Database (Sugihara, 2003) Pped(Pa)
1.824?104M1/3Ip2R-2.1a-0.57?3.81(1?2)-7/3(1?)3.4
1nped-1/3(Ptot/PLH)0.144 ----gt Tped 5.24 1.3
keV ----gt ?ped?? L-H Transition PLH(MW)
2.84Meff-1BT0.82nL200.58Ra0.81 (2000) ----gt 26
MW in flattop PLH(MW) 2.58Meff-1BT0.60nL200.70R0
.83 a1.04 (2002) ----gt 18.5-25 MW in flattop DN
has less or equal PLH compared to favored SN
(Carlstrom, DIII-D NSTX MAST) H-L Transition
ELMs Ploss gt PLH although hysterisis exists in
data Type I ELMs typically require Ploss gt 1.(
)?PLH, expts typically gt 2?PLH Type II ELMs
require strong shaping, higher density, DN ---gt
reduced Pdiv, H981 Type III ELMs, near Ploss
PLH, or high density, reduced H98 Active methods
----gt pellets, gas puffing, impurity seeding,
ergodization
44
Pedestal Physics and ELMs
ELITE projections for FIRE
Pedestal physics Intermediate n
peeling/ballooning modes ----gt ballooning
destabilized by high p and low j ----gt peeling
modes destabilized by high j and low p Stronger
shaping raises pped Stability analysis
distinguishes nped and Tped through ?ped
(nped/Tped2) ---gt jBS Higher nped leads to mode
envelope narrows and lowers jBS ---gt smaller
?WELM
weak shaping
strong shaping
45
Pedestal Physics and ELMs
Type I ELM trends Reduced ?WELM/Wped with
increasing ?ped ----gt inconsistent with higher
Tped for high Q Reduced ?WELM/Wped with
increasing ?i ----gt inconsistent with higher
Tped for high Q ?WELM/Wped correlated with
?Tped/Tped as nped varied, very little change in
?Nped/Nped
Type II ELMs ASDEX-U with DN and high n ----gt
H98 1-1.2 and reduction in divertor heat flux
by 3? JET with high ? and high n ----gt mixed
Type III, no reduction in confinement and 3?
reduction in ELM power loss
Pin
JET
PELM
Wth
Prad
46
Pedestal Physics and ELMs
Active methods for ELM mitigation JET argon
seeding in Type I, frad gt 0.65, H98 1, n/nGrgt
0.7, Q div reduced by 2? Type III, frad gt 0.7,
H98 0.7-0.9, n/nGr gt 0.7 Pellets that trigger
ELMs, avoiding large infrequent Type I
ELMs Ergodization of plasma edge region, use
coils to produce high (m,n) field that perturb
only ELM region
JET
47
POPCON Operating Space vs. Parameters
T(0)/?T?, n(0)/?n?, ?p/?E, H98, fBe, fAr
H98(y,2) must be 1.1 for robust operating space
48
1.5D Integrated Simulations
Tokamak Simulation Code (TSC) Free-boundary Energ
y and current transport Density profiles
assumed GLF23 MMM core energy transport Assumed
pedestal height/location ICRF heating, data from
SPRUCE Bootstrap current, Sauter single
ion Porcelli sawtooth model Coronal equilibrium
radiation Impurities with electron density
profile PF coils and conducting
structures Feedback systems on position, shape,
current Use stored energy control Snowmass E2
simulations for FIRE Corsica, GTWHIST, Baldur,
XPTOR
49
1.5D Integrated Simulations
FIRE H-mode, GLF23
50
1.5D Integrated Simulations
51
0D Advanced Tokamak Operating Space
Scan ----gt q95, n(0)/?n?, T(0)/?T?, n/nGr, ?N,
fBe, fAr Constrain ----gt ?LH 0.16, ?FW 0.2,
PLH 30 MW, P 30 MW, IFW 0.2 MA, ILH
(1-fbs)Ip, Q Screen ----gt ?flattop(VV, TF, FW
heating), Prad(div), Ppart(div), Pauxlt Pmax
52
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
• Access to higher radiated power fractions in
  divertor enlarge operating space significantly
• Access to higher ?flattop/?j decreases at higher
  ?N, higher BT, and higher Q, since ?flattop set
  by VV nuclear heating

53
Examples of FIRE Q5 AT Operating Points That
Obtain ?flat/?J gt 3
HH lt 1.75, satisfy all power constraints,
Pdiv(rad) lt 0.5 P(SOL)
54
1.5D Integrated Simulations AT-mode
fBS0.77 Zeff2.3 q(0) 4.0 q(min) 2.75
q(95) 4.0 li 0.42, ? 4.7, ?P 2.35
Ip4.5 MA Bt6.5 T ?N4.1 t(flat)/?j3.2
I(LH)0.80 P(LH)25 MW
55
1.5D Integrated Scenarios AT-mode
t 12-41 s
56
1.5D Integrated Scenarios AT-mode
Q 5 I(bs) 3.5 MA, I(LH) 0.80 MA I(FW)
0.20 MA, t(flattop)/?j3.2
n/nGr 0.85 n(0)/ltngt 1.4 n(0) 4.4x1020 Wth
34.5 MJ tE 0.7 s H98(y,2) 1.7 Ti(0) 14
keV Te(0) 16 keV Dy(total) 19 V-s, Pa 30
MW P(LH) 25 MW P(ICRF/FW) 7 MW (up to 20 MW
ICRF used in rampup) P(rad) 15 MW Zeff 2.3
57
Perturbation of AT-mode Current Profile
5 MW perturbation to PLH Flattop time is
sufficient to examine CD control
t 12 s t 25 s
t 25 s t 41 s
58
Conclusions

• The FIRE device design provides
  sufficient/flexible/relevant operating space to
  examine burning plasma physics
• Sufficient to provide burning conditions (Q 10
  inductive and Q 5 AT, does not preclude
  ignition)
• Flexible to accommodate uncertainty and explore
  various physics regimes
• Relevant to power plant plasma physics and
  engineering design
• The subsystems on FIRE, within their operating
  limits, are suitable to examine burning plasma
  physics ----gt subject to RD in some cases
• Auxiliary heating/CD
• Particle fueling and pumping
• Divertor/baffle and FW PFCs
• Magnets
• Diagnostics

59
Conclusions

• Burning plasma conditions can be accessed and
  studied in both standard H-mode and Advanced
  Tokamak modes. The range of AT performance has
  been expanded significantly since Snowmass
• FIRE can reach 1-5 ?j, and examine current
  profile control
• Design improvement to FW tiles could extend
  flattop times further
• FIRE can reach 80-90 of ideal with wall limit,
  with RWM feedback
• FIRE can reach high IBS/IP (77 in 1.5D
  simulation)
• Identified that radiative mantle/divertor
  solutions significantly expand operating space
• FIRE will pursue Fe shims for AT operation
• The physics basis for FIREs operation is based
  on current experimental and theoretical results,
  and projections based on these continue to
  provide confidence that FIRE will achieve the
  required burning plasma performance

60
Issues/Further Work

• Magnets
• Ripple reduction, design Fe shims for AT mode
• Continue equilibrium analysis
• Complete plasma breakdown and early startup
• Complete internal control coil analysis
• RWM coil design/integration into port plugs, time
  dependent analysis
• Error field control coil design
• Heating and CD
• Continue ICRF antenna design, disruption loads,
  neutron/surface heating
• Engineering of 4 strap expanded antenna option
• More detailed design of LH launcher, disruption
  loads, neutron/surface heating
• Complete 2D FP/expanded LH calculations for FIRE
  specific cases
• Continue examination of EC/OKCD for NTM
  suppression in AT mode
• Pursue dynamic simulations/PEST3 analysis of LH
  NTM stabilization for both H-mode and AT-mode

61
Issues/Further Work

• Power Handling
• Pulse length limitations from VV nuclear heating,
  design improvements
• FW tile design, material choices, impacts on
  magnetics
• Continue divertor analysis, UEDGE and neutrals
  analysis for integrated heat load, pumping,and
  core He concentration solutions
• Continue examination of ITPA ELM results and
  projections, encourage DN strong triangularity
  experiments
• DN up-down imbalance, implications for divertor
  design (lots of work on DII-D)
• Disruption mitigation strategies, experiments
• Particle Handling
• Continue pellet and gas fueling analysis in high
  density regime of FIRE
• Neutrals analysis for pumping
• Be behavior as FW material and intrinsic impurity
• Impurity injection, core behavior, and
  controllability
• Particle control techniques puff and pump,
  density feedback control, auxiliary heating to
  pump out core, etc.
• Wall behavior, no inner divertor pumping, what
  are impacts?

62
Issues/Further Work

• MHD Stability
• LH stabilization of NTMs, analysis and
  experiments (JET, JT-60U and C-Mod)
• Examine plasmas that appear not to be affected by
  NTMs (current profile)
• Early (before they are saturated) stabilization
  of NTMs with EC/OKCD
• Continue to develop RWM feedback scheme in
  absense of rotation
• Identify impact of n2,3 modes on wall/feedback
  stabilized plasmas
• Examine impact of no external rotation source on
  transport, resistive and ideal modes
• Alfven eigenmodes/energetic particle modes, onset
  and accessibility in FIRE
• Plasma Transport and Confinement
• Continue core turbulence development for H-mode,
  ITPA
• Establish AT mode transport features, ITB onset,
  ITPA
• Pedestal physics and projections, and ELM
  regimes, ITPA
• Impact of DN and strong shaping on operating
  regimes, Type II ELMs
• Improvements to global energy confinement
  scaling, single device trends
• Expand integrated modeling of burning plasmas