fire physics basis (detailed version) c. kessel for the fire team princeton plasma physics...
Post on 19-Dec-2015
221 views
TRANSCRIPT
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
AES, ANL, Boeing, Columbia U., CTD, GA, GIT, LLNL, INEEL, MIT, ORNL, PPPL, SNL, SRS, UCLA, UCSD, UIIC, UWisc
FIRE Collaboration http://fire.pppl.gov
FIRE Description
H-modeIP = 7.7 MABT = 10 TN = 1.80 = 2.4%P = 0.85 = 0.075%q(0) < 1.0q95 ≈ 3.1li(1,3) = 0.85,0.66Te,i(0) = 15 keVTe,i = 6.7 keVn20(0) = 5.3n(0)/n = 1.15p(0)/p = 2.4n/nGr = 0.72Zeff = 1.4fbs = 0.2Q = 12burn = 20 s
R = 2.14 m, a = 0.595 m, x = 2.0, x = 0.7, Pfus = 150 MWAT-ModeIP = 4.5 MABT = 6.5 TN = 4.2 = 4.7%P = 2.35 = 0.21%q(0) ≈ 4.0q95, qmin ≈ 4.0,2.7li(1,3) = 0.52,0.45Te,i(0) = 15 keVTe,i = 6.8 keVn20(0) = 4.4n(0)/n = 1.4p(0)/p = 2.5n/nGr = 0.85Zeff = 2.2fbs = 0.78Q = 5 burn = 32 s
portplasma
divertorbaffle
passiveplate
VV
FIRE Magnet Layout
TF Coil
CS1
CS2
CS3
PF1,2,3PF4
PF5
Error field correction coils
Fast vertical and radial position control coil RWM feedback coil
Fe shims
Toroidal Field Coils16 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
----> 20 s flattop
AT-mode
BT = 6.5 T and Pfus = 150 MW
----> 48 s flattop
Maximum TF ripple at R+a is 0.3%
0.3% -particle loss H-mode
8% -particle loss AT-mode
Expect to use Fe shims for AT-mode
Poloidal Field CoilsCenter Stack 1, 2U&L, 3U&LPF1,2,3,4,5 U&LAll CS and PF coils are CuCrZr
H-modeFiducial equilibria in discharge
IM, SOD, SOH, SOB, EOB, EOH, EOD
Flexibility of PF coils0.55 ≤ li(3) ≤ 0.85 (SOB,EOB)0.85 ≤ li(3) ≤ 1.15 (SOH,EOH)ref-5 ≤ (Wb) ≤ ref+51.5 ≤ N ≤ 3.0
Full operating space available within stress (1.3 margin) and heating allowables, except at EOB, li=0.85 where ≤ ref-2
Rampup consumes ≈ 40 V-sFlattop consumes ≈ 3 V-s
CS1
CS2CS3
PF1,2,3 PF4
PF5
Poloidal Field CoilsAT-modeFiducial equilibria in dischargeIM, SOD, SOH, SOF, SOB, EOB, EOH, EOD
Flexibility of PF coils0.35 ≤ li(3) ≤ 0.65 (SOB & EOB)2.5 ≤ N ≤ 5.07.5 ≤ flattop(Wb) ≤ 17.5
Full operating space is available for Ip ≤ 5.0 MA
PF coils can provide pulse length limitation -----> 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
Poloidal Field CoilsTSC simulations
Free-boundary calculations with heating, CD, bootstrap current, energy and current transport, impurities, PF coils, structure and feedback systems, etc.
---> check of equilibrium coil currents---> Volt-second consumption---> Feedback control of vertical position, radial position, plasma current and shape
Vertical StabilityDesign passive structures to slow vertical instability for feedback control and provide a stability factor fs > 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 < li(3) < 1.1, the stability factor is 1.3 < fs < 1.13 and growth time is 43 < g(ms) < 19
Passive stabilizers Cladding
(ports provide poloidal cuts)
Cladding(large number of poloidal cuts)
Internal Control Coils
8 OFHC Cu coils (2nd redundant coils) above and below the midplane
Fast vertical position feedback controlZRMS = 1 cm, 65-90 kA-turn, 50-75 V/turn ------> 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
Resistive Wall Mode (RWM) Coils
Current strap, grounded at each end
Faraday shield(one side only)
Port flange
ICRF Port Plug
RWM Coil
DIII-D experienceModes are detectable at the level of 1GThe C-coils can produce about 50 times this fieldThe necessary frequency depends on the wall time for the n=1 mode (which is 5 ms in DIII-D) and they have wall ≈ 3
FIRE projectionFIRE has approximately 3-4 times the DIII-D plasma current, so we might be able to measure down to 3-4 GIf we try to guarantee at least 20 times this value from the feedback coils, we must produce 60-80 G at the plasmaThese fields require approximately I = f(d,Z,)Br/o = 5-6.5 kA
Assume we also require wall ≈ 3Required voltage would go as V ≈ 3o(2d+2Z)NI/wall ≈ 0.25 V/turn
Error Field Correction CoilsStatic 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 (m=1,n=1), (2,1), and (3,1) error fields, simultaneously
And provide factor of ≈ 5 reduction in net error field Br2,1
net
3 distributed coils provides poloidal mode control allowing multiple (m,n=1) suppression
Recent C-Mod data shows that applied Br2,1 of 610-4 T removed mode-locking -----> Important since C-Mod does NOT have external rotation source
No analysis performed
ITER Error Coils
ICRF Heating and CD
1250
160 300Current straps
Faraday shield
Coax feeds
First wall Ant front-side 7/30/03
Adjustablecapacitorstructure
Shielding
.32
.71
2.56
.65
FIRE Antenna Plan
4/24/01
Dimensions in m
1999 version of Vacuum vessel
Frequency range 70-115 MHz
2 strap antennas
4 ports, total power 20 MW
H-modeBT = 10 T, minority He3 and 2T at 100 MHzFrequency range allows heating at a/2 on HFS and LFS
AT-modeBT = 6.5 T, ion heating at minority H and 2D at 100 MHzFrequency 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 analysisnHe3/ne=2%PICRF=11.5 MW=100 MHzTHe3(0)=10.2 keVPHe3=60%PT=10%PD=2%Pe=26%
ICRF Heating and CD
BT = 10 T BT = 6.5 T
Vacuum Toroidal Field Resonances
H, 2D, 3T 2 H
D
He3, 2T
2Be
2 He3 3 D
50.0
60.0
70.0
80.0
90.0
100.0
110.0
120.0
130.0
140.0
150.0
1.40 1.60 1.80 2.00 2.20 2.40 2.60 2.80 3.00
Major radius (m)
Frequency (MHz)
B0 = 10.00 TR0-a R0+a
Freq. Range
Be
H, 2D, 3T
2 H
D
He3, 2T
2Be
2 He3
3 D
50.0
60.0
70.0
80.0
90.0
100.0
110.0
120.0
130.0
140.0
150.0
1.40 1.60 1.80 2.00 2.20 2.40 2.60 2.80 3.00
Major radius (m)
Frequency (MHz)
B0 = 6.50 TR0-a R0+a
Freq. Range
Be
ICRF Heating and CD
Module A
Module B
.27
.23
10 8 6 4 2 0 2 4 6 8 10
1
2
3
nz
Launched
spectrum
"Good CD"
region
.32
.71
2.56
.65
FIRE Antenna Plan
4/24/01
Dimensions in m
1999 version of Vacuum vessel
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
ICRF Heating and CDAORSA full wave analysis continues including fast alpha and Be impurity effects
75 MHz 70 MHzPe=0.44PT=0.15PBe=0.30
Pe=0.65PT=0.32PBe=0.0
20=0.14 20=0.17
Lower Hybrid Current Drive
Port outline
Each block =8 waveguides+ cooling
1250
720
Array of 768 waveguides in midplane port
536
EC Launcher
EC Launcher
Blow-up of one square
60
67
Frequency 5 GHz
Spectrum n|| ≈ 1.8-2.5, n|| = 0.3
Power of 30 MW, in 2 ports
Upgrade to baseline design
H-modeUsed for NTM control for BT = 10 TUsed for non-inductive CD for hybrid discharges
AT-modeUsed for bulk CD for BT = 6.5 TCD efficiency 20 ≈ 0.16 A/W-m2 at 6.5 T (30-50% higher from 2D FP calcs.)Used for NTM control
f > 2fLH
RF power flux is 53 MW/m2
Need 0.57 m2 per waveguide for 30 MWEach waveguide is 5.7 cm(tall) 0.65 cmHave ≈ 1500 waveguides
Lower Hybrid Current Drive
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 ----> 0.25 A/W-m2BT = 6.5T ----> 0.16 A/W-m2
Benchmarks with ACCOME, CURRAY and LSC
3.7 GHz, 750 kW, 1000s sources availableITER estimate for 5 GHz, 1.0 MW, CW sources was 1.15 euro/watt
ACCOME
TSC-LSC
Electron Cyclotron
=ce=170 GHz
pe=ce
Rays are bent as they approach = pe
Rays are launched with toroidal directionality for CD
Frequency of 170 GHz to utilize ITER R&D
LFS, O-mode, fundamental
FIRE has high density and high fieldCutoff of EC when = pe
AT-modeLower BT = 6.5 T
LFS deposition implies trapping reduction of CD, however, Ohkawa effect provides more CD than standard EC
Current required, scaled as IpN2 from
DIII-D and ASDEX-U expts for (3,2) mode----> drive 200 kA to suppress from saturated state----> requires 100 MW!
Electron Cyclotron
Bt=6.5 T
Bt=7.5 T
Bt=8.5 T
Ro Ro+a
fce=182 fce=142
fce=210 fce=164
fce=190fce=238
170 GHz
200 GHz
qmin(3,1)
J. Decker, MIT
145≤≤155 GHz-30o≤L≤-10o
midplane launch10 kA of current for 5 MW of injected power
r/a(qmin) ≈ 0.8r/a(3,1) ≈ 0.87-0.93Does (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
Neutral Beam Injection (Difficult)
.32
.71
2.56
.65
FIRE Antenna Plan
4/24/01
Dimensions in m
1999 version of Vacuum vessel
Rtan = 0 m
Rtan = 0.75 m
Rtan = 1.7 m
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 > 1 MeV to get 50% of power inside a/2
Plasma rotation for Rtan = 1.7 m
Assumed 120 keV & 8 MW ----> deposited r/a > 0.65
Dominated by jB rotation giving v/vAlfven ≤ 0.5%
Power Handling
First wallSurface heat flux
Plasma radiation, Qmax = P+ Paux
Volumetric heatingNuclear heating, qmax = qpeak(Z=0)
VV, Cladding, Tiles, Magnets….Volumetric heating
Nuclear heating, qmax = qpeak(Z=0)
DivertorSurface heat flux
Particle heat flux, Qmax = PSOL/Adiv(part)Radiation heat flux, Qmax = PSOL/Adiv(rad)
Volumetric heatingNuclear heating, qmax = qpeak(divertor)
plasm
a
VV Clad Tile
Power HandlingPulse length limitations
VV nuclear heating (stress limit), 4875 MW-s -----> Pfus (qVV
nuclear)
FW Be coating temperature, 600oC -----> QFW & Pfus (qBe
nuclear)
TF coil heating, 373oK -----> BT & Pfus (qCu
nuclear)
PF Coil heating-AT-mode, 373oK -----> Ip, li, p, and (not limiting)
Component limitations
Particle power to outboard divertor < 28 MW
Radiated power on (inner&outer) divertor/baffle < 6-8 MW/m2
Power Handling/Operating SpaceFIRE H-mode Operating Space
N limited by NTM or ideal MHD with NTM suppression-----> maximum Pfus
Higher radiated power in the divertor allows more operating space, mainly at higher N
-----> maximum Pfus
Majority of operating space limited by TF coil flattop-----> 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
Power Handling/Operating SpaceFIRE AT-mode Operating Space
N is limited by ideal MHD w/wo RWM feedback -----> maximum Pfus
Higher radiated power in the divertor allows more operating space, mainly at higher N
-----> maximum Pfus
Majority of operating space limited by VV nuclear heating-----> 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
Particle Fueling/Pumping
Gas Fueling System Pellet Fueling System Remarks
Design fueling rate 200 torr-l/s for 20 s 200 torr-l/s for 20 s Pumping cap. 200 torr-l/s
Operational fuel rate 100-175 torr-l/s 100-125 torr-l/s Isotopic fueling
Fuel isotope D(95-99%), T,H(5-1%) T(40-99%), D(60-1%) D-rich edge, T-rich core
Impurity fuel rate 25 torr-l/s Prefer gas
Impurity species He3, Ne, Ar, others
Rapid shutdown Massive gas puff Killer pellet or liquid jet Disruption/VDE
Pellet sizes 3-4 mm
Require ≈ 1-21021 tritons/s for FIRE H-mode---> 0.1-0.2 g T injected per shot (20 s)---> 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
Particle Fueling/Pumping16 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
Pumpdown and vessel bakeout utilize midplane pump, to provide minimum of 2000 l/s to reach 10-7 or less base pressure
Particle Fueling/Pumping
V = 125 m/sParks, 2003
WHIST simulation of FIRE H-mode discharge (Houlberg)
Assume uniform pellet depositionObtains some density peaking with sufficient pumping
MHD StabilitySawtooth H-mode
Unstable to internal kink, r/a(q=1) ≈ 0.35 m-----> 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 profileRapid reheat of sawtooth volume
-particles providing stabilization
Complete stabilization would require RFCD since FIRE does not have high energy minority species
The q=1 surface can be removed from the plasma by
1.2 MA off-axis CDReduction of Ip to 6.0 MA
MHD StabilityNeoclassical Tearing Modes H-mode
Stable or unstable?
Sawteeth and ELM’s are expected to be present and can drive NTM’s
Typical operating point is at low N and P
Can lower N further if near threshold
Lower Hybrid CD at the rational surfacesCompass-D demonstrated LH stabilizationAnalysis by Pletzer and Perkins showed stabilization was feasible (PEST3)Lowers Q(=Pfus/Paux)
EC methods require high frequencies at FIRE field and densities ----> 280 GHz
TSC-LSC
(3,2) surface
12.5 MW
0.65 MA
n/nGr = 0.4
Q = 6.8
MHD Stability
FIR-NTMsususal NTMs
S. Günter et al., PRL 2001
0.030.012
ASDEX-U
Current profile modification
FIRE MHD Stability
Despite (3,2) NTM excellent confinement: H98y=1.4, N = 3.3
(LHCD ctr-CD in start-up phase)
JET
Current profile modification
MHD StabilityIdeal MHD Stability H-mode
n=1 external kink and n=∞ 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 ELM’s
Type I ELM’s are divertor lifetime limiting, must access Type II, III
Ploss/PLH ≈ 1.0-1.6 in flattop, not > 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
MHD StabilityNeoclassical Tearing Modes AT-mode
Unstable or Stable?
q() > 2 everywhere, so rational surfaces are (3,1), (5,2), (7,3), (7,2)…
r/a(qmin) ≈ 0.8r/a(3,1) ≈ 0.87-0.93Local *, *, Rem effects so close to plasma edge? L-mode or H-mode conditions
Examining EC stabilization at 170 GHzLFS absorption, Ohkawa CD dominatesScaling from (3,2) expts indicates high power ----> early detection required
LH using two spectra, one for bulk CD and other for NTM suppression
MHD StabilityIdeal MHD Stability AT-mode
n= 1, 2, and 3…external kink and n = ∞ ballooning modes
n = 1 stable without a wall/feedback for N < 2.5-2.8n = 2 and 3 have higher limits without a wall/feedback
Ballooning stable up to N < 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 n=1
n = 1 stable with wall/feedback to N’s 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
Gro
wth
Rat
e, /s
N
N=4.2
Bialek, Columbia Univ.
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. ----> 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
Disruption ModelingComments
frequency 10-30% 30% plasma development, 10% for repetitive operation
number 300 (900) 300 at maximum Wmag and Wth
thermal energy 35 MJ
themal quench duration 0.2 ms (0.1-0.5) single or multi-step quench
fraction of Wth to divertor 80-100% (or ver low for mitigation) conduction to targets, 2-1 asymmetry
fraction of Wth to first wall/baffle < 30% (or nearly 100% for mitigation) radiation
inside to outside divertor split 5 to 1 need data on DN
poloidal localization 3x typical SOL width(1x to 10x) this may not be reliable
magnetic energy 25 MJ
current quench 6 ms (2-600 ms)
Ip decay rate 1 MA/ms typical, 3 MA/ms maximum
fraction of Wmag to FW by radiation 80-100% peaking factor of 2
fraction of Wmag to FW by conduction 0-20%
VDE frequency 10% of disruptions uncertain due to up-down symmetry
halo current fraction, Ihalo/Ip 0.4 (0-0.5)
halo current toroidal peaking 2 (1.2-4)
(Ihalo/Ip) TPF 0.5 typical maximum experimental data boundary is 0.75
runaway electron current 50% of Ip Highly uncertain, high density in FIRE makes this much weaker
localization of runaway deposition < 1 m2 PFC and wall alignment
Disruption ModelingExperimental database used to project for FIRE
Thermal quench timeIhalo/Ip TPFdIp/dt rates for current quench
Disruption ModelingTSC simulation of disruption
Critical structures modeled; VV’s (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
Disruption Modeling
Thermal quench, t = 0.2 ms
Ip drop, -2.9 MA/ms
Diamagnetic flux
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
Disruption ModelingTSC 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
Disruption MitigationUtilize fueling technology to mitigate electromagnetic effects of disruptions
Massive gas puff into DIII-D ----> peak halo currents reduced by 50% by He and D puffing, and toroidal asymmetry reduced
Ne, Ar, and CH4 pellets into DIII-D ----> 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 MassiveGas Puff System
FIRE Transport and ConfinementEnergy Confinement Database
E98(y,2) = 0.144 M0.19 Ip0.93 BT
0.15 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) = 1.824104M1/3Ip2R-2.1a-0.573.81(1+2)-7/3(1+)3.41nped
-1/3(Ptot/PLH)0.144
----> Tped = 5.24 ± 1.3 keV----> ped??
L-H TransitionPLH(MW) = 2.84Meff
-1BT0.82nL20
0.58Ra0.81 (2000) ----> 26 MW in flattopPLH(MW) = 2.58Meff
-1BT0.60nL20
0.70R0.83 a1.04 (2002) ----> 18.5-25 MW in flattop
DN has less or equal PLH compared to favored SN (Carlstrom, DIII-D; NSTX; MAST)
H-L Transition & ELM’s Ploss > PLH although hysterisis exists in dataType I ELM’s typically require Ploss > 1.( )PLH, expts typically > 2PLH
Type II ELM’s require strong shaping, higher density, DN ---> reduced Pdiv, H98=1Type III ELM’s, near Ploss ≈ PLH, or high density, reduced H98
Active methods ----> pellets, gas puffing, impurity seeding, ergodization
Pedestal Physics and ELM’sPedestal physics
Intermediate n peeling/ballooning modes----> ballooning destabilized by high p’ and low j----> peeling modes destabilized by high j and low p’
Stronger shaping raises pped
Stability analysis distinguishes nped and Tped through *
ped (nped/Tped2) ---> jBS
Higher nped leads to mode envelope narrows and lowers jBS ---> smaller WELM
weak shaping strong shaping
ELITE projections for FIRE
Pedestal Physics and ELM’sType I ELM trends
Reduced WELM/Wped with increasing *ped ----> inconsistent with higher Tped for
high Q
Reduced WELM/Wped with increasing ||i ----> inconsistent with higher Tped for
high Q
WELM/Wped correlated with Tped/Tped as nped varied, very little change in Nped/Nped
Type II ELM’sASDEX-U with DN and high n ----> H98 = 1-1.2 and reduction in divertor heat flux by 3
JET with high and high n ----> mixed Type I+II, no reduction in confinement and 3 reduction in ELM power loss
Pin
Wth
Prad
PELM
JET
Pedestal Physics and ELM’s
Active methods for ELM mitigation
JET argon seeding in Type I, frad > 0.65, H98 ≈ 1, n/nGr> 0.7, Q div reduced by 2Type III, frad > 0.7, H98 ≈ 0.7-0.9, n/nGr > 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
POPCON Operating Space vs. ParametersT(0)/T, n(0)/n, p
*/E, H98, fBe, fAr
H98(y,2) must be ≥ 1.1 for robust operating space
1.5D Integrated Simulations
Tokamak Simulation Code (TSC)
Free-boundaryEnergy and current transportDensity profiles assumedGLF23 & MMM core energy transportAssumed pedestal height/locationICRF heating, data from SPRUCEBootstrap current, Sauter single ionPorcelli sawtooth modelCoronal equilibrium radiationImpurities with electron density profilePF coils and conducting structuresFeedback systems on position, shape, currentUse stored energy control
Snowmass E2 simulations for FIRECorsica, GTWHIST, Baldur, XPTOR
1.5D Integrated SimulationsFIRE Q Paux(MW) Tped(keV)
TSC
GLF
10.3 13.5 4.5
10.0 7.5 3.8
10.0 10.0 4.1
10.0 12.5 4.4
10.0 15.0 4.7
10.0 20.0 5.4
Baldur
MMM
4.5 30.0 2.5*
7.0 10.0 2.5*
XPTOR/12
GLF
5.0 20.0 3.0
10.0 20.0 4.0
15.0 20.0 5.0
Corsica
GLF
4.0 12.5 2.5
6.0 12.5 4.0
10.0 12.5 5.0
0D Advanced Tokamak Operating SpaceScan ----> q95, n(0)/n, T(0)/T, n/nGr, N, fBe, fAr
Constrain ----> LH = 0.16, FW = 0.2, PLH ≤ 30 MW, P ≤ 30 MW, IFW = 0.2 MA,ILH = (1-fbs)Ip, QScreen ----> flattop(VV, TF, FW heating), Prad(div), Ppart(div), Paux< Pmax
Observations from 0D Analysis for Burning Plasma AT
• In order to provide reasonable fusion gain Q≥5, can’t 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 > 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
Examples of FIRE Q=5 AT Operating Points That Obtain flat/J > 3
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
HH < 1.75, satisfy all power constraints, Pdiv(rad) < 0.5 P(SOL)
1.5D Integrated Simulations AT-modeIp=4.5 MA Bt=6.5 T N=4.1 t(flat)/j=3.2 I(LH)=0.80P(LH)=25 MW
fBS=0.77 Zeff=2.3 q(0) =4.0 q(min) = 2.75 q(95) = 4.0 li = 0.42, = 4.7%, P = 2.35
1.5D Integrated Scenarios AT-moden/nGr = 0.85
n(0)/<n> = 1.4
n(0) = 4.4x10^20
Wth = 34.5 MJ
E = 0.7 s
H98(y,2) = 1.7
Ti(0) = 14 keV
Te(0) = 16 keV
(total) = 19 V-s,
P = 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
Q = 5
I(bs) = 3.5 MA, I(LH) = 0.80 MA
I(FW) = 0.20 MA, t(flattop)/j=3.2
Perturbation of AT-mode Current Profile
5 MW perturbation to PLH
Flattop time is sufficient to examine CD control
t = 12 st = 25 s
t = 25 st = 41 s
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 ----> subject to R&D in some cases– Auxiliary heating/CD– Particle fueling and pumping– Divertor/baffle and FW PFC’s– Magnets– Diagnostics
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 FIRE’s 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
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
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?
Issues/Further Work
• MHD Stability– LH stabilization of NTM’s, analysis and experiments (JET, JT-60U and C-Mod)
– Examine plasmas that appear not to be affected by NTM’s (current profile)
– Early (before they are saturated) stabilization of NTM’s with EC/OKCD
– Continue to develop RWM feedback scheme in absense of rotation
– Identify impact of n=2,3 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