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

JSOLVER Equilibria

TSC-LSC Simulation Equilibria

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 Simulations

FIRE H-mode, GLF23

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-mode

t = 12-41 s

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