Simulation and Analysis of the Hybrid Operating Mode in

ITER

C. Kessel, R. Budny, and K. IndireshkumarPrinceton Plasma Physics Laboratory

Symposium On Fusion Engineering

Knoxville, TN

September 26-29, 2005

So What is a Hybrid Operating Mode in ITER?

Reference H-mode

Ip = 15 MABT = 5.3 TR = 6.2 ma = 2.0 m

Vloop = 0.09q95 = 3N = 1.8H98(y,2) = 1.0q(0) ≤ 1.0(rsaw ≈ 1 m)

Q = 10Tflattop = 500 s

Hybrid Mode

Ip = 12 MABT = 5.3 TR = 6.2 ma = 2.0 m

Vloop = 0.025-0.04q95 = 4N = 3.0H98(y,2) = 1.5q(0) ≥ 1.0(rsaw small)

Q = 5-10Tflattop = 3000 s

Steady State (AT) Mode

Ip = 9 MABT = 5.3 TR = 6.35 ma = 1.85 m

Vloop = 0.0q95 = 4N = 3-4.5H98(y,2) = 1.6q(0) > 1.5-2.0

Q = 5Tflattop = ∞

Hybrid Scenario in ITER

• Plasma parameter ranges E ≈ 1.0-1.5 E

98(y,2)

NNTM < N < N

no wall (≈ 3)– fNI ≈ 50%– IP ≈ 12 MA– n/nGr varied CD determined from TRANSP,

or other analysis– Impurities defined to provide

acceptable divertor heat loading

• Operating Modes– NNBI + ICRF– NNBI + ICRF + LH– NNBI + ICRF + EC

• Prefer to avoid (or minimize) the sawtooth, q(0) ≥ 1.0– Maximize fNI

off-axis (IBS, ILH, IECCD)

• Maximize neutron fluence– Nwall tflattop

– tflattop is minimum of tV-s or tnuc-

heat

• Remain within installed power limitations– NNBI at 1.0 MeV, 33 MW– ICRF at about 52 MHz, 20

MW– EC at 170 GHz, 20 MW– LH at 5 GHz, 30+ MW

(UPGRADE)

Integrated Modeling of ITER Hybrid Burning Plasma Scenarios

• 0D systems analysis to identify operating space within engineering contraints

• 1.5D discharge simulations– Energy transport (GLF23)– Heating/CD– Free-boundary equilibrium evolution/feedback control

– Other control; stored energy, fNI, etc.

• Energy transport experimental verification

• Ideal MHD analysis

• Offline heating/CD source analysis

• Offline gyrokinetic transport simulations (Budny)

• Fast particle effects and MHD (Gorelenkov)

• Particle transport/impurity transport

• Integrated SOL/divertor modeling

• Non-ideal MHD, NTM’s

0D Systems Analysis Identifies Device Constraints for Scenario Simulations

• ITER’s Primary Device Limitations That Affect Scenarios– Fusion power vs pulse length ----> heat rejection system

• 350 MW for 3000 s• 500 MW for 400 s

• 700 MW for 150 s ----> (maximum Pfus cryoplant limits)

– Divertor conducted heat load, maximum > 20 MW/m2, nominal 5-10 MW/m2 ----> allowable divertor heat load

• Radiation from plasma core and edge, PSOL = (1 - fcorerad) Pinput

• Radiation in divertor and around Xpt, Pcond = (1 - fdivrad) PSOL

• Radiation distribution in divertor channel, impurities, transients

– Volt-second capability ----> PF coil current limits• Approximately 260-280 V-s

– First wall surface heat load limit (not limiting for normal operation)

– Duty cycle, tflattop/(tflattop + tdwell) ----> cryoplant for SC coils

• Limited to about 25%What device upgrades are required for advanced operating modes, and are they major or minor upgrades?

0D Operating Space Analysis

Energy balance

Particle balance, P*/E and quasi-neutrality

Bosch-Hale fusion reactivity

Post-Jensen coronal equilibrium

Albajar cyclotron radiation model

Hirshman-Neilson flux requirement(benchmarked with TSC)

T(r) = (To - Ta)[1-(r/a)2]T + Ta

Same for density profile

Etc.

IP = 12 MABT = 5.3 TR = 6.2 mA = 3.195 = 1.7595 = 0.5P*/E = 5∆total = 300 V-s∆breakdown = 10 V-sli = 0.80CE = 0.45NBCD = 0.3 x 1020 A/W-m2

PCD = 33 MWT = 1.75, Ta/To = 0.1n = 0.075, na/no = 0.3fBe = 2.0%

1.5 ≤ N ≤ 3.00.4 ≤ n/nGr ≤ 1.03.0 ≤ Q ≤ 12.00.0 ≤ fC ≤ 2.0%0.0 ≤ fAr ≤ 0.2%

Input parameters

Scanned parameters

ITER Hybrid Systems Analysis

Fusion power pulse length limitation significantly reduces accessible fluence values, and changes dependence on density

ITER Hybrid Systems Analysis

Operating space shows strong dependence on allowable conducted peak heat flux on divertor, which must be low enough to accommodate radiation flux and transients

ITER Hybrid Systems Analysis

Increasing the power radiated in the divertor can recover operating space at lower conducted peak heat flux

ITER Hybrid Systems Analysis

Large Operating Space Scan

1.05 ≤ n(0)/n ≤ 1.251.5 ≤ T(0)/T ≤ 2.511.0 ≤ IP (MA) ≤ 13.01.5 ≤ N ≤ 3.00.4 ≤ n/nGr ≤ 1.03.0 ≤ Q ≤ 12.01% ≤ fBe ≤ 3%0% ≤ fC ≤ 2%0% ≤ fAr ≤ 0.2%

Other input fixed at previous values

Results

• Fusion power pulse length limitation is a significant factor in determining Hybrid operating space– Hybrid operating modes on present tokamaks operate in N

window, close to N ≈ 3– Existing pulse length vs fusion power limits indicate optimum N to

maximize neutron fluence is about 2.0 (Pfusion ≈ 325 MW)– For ITER to operate close to N ≈ 3, Pfusion ≈ 500 MW, the pulse

length would be severely limited by heat rejection system– Hybrid operating modes in ITER require upgrades to heat rejection

system

• Volt-seconds capability of PF coils appears to be enough to

offer few thousand second flattops– Depending on precise value of Vloop

• First wall surface heat load limits do not appear to be limiting during normal operation due to large FW surface area

• Divertor heat load limits is second most significant factor for Hybrid operating space– Core/edge radiated power (bremsstrahlung, cyclotron, line)– Conducted power– Power radiated in divertor region – Transient conducted power

• Operating space shows that existing ITER design can provide reasonable fluence levels within a discharge– HOWEVER time between discharges is constrained– Appears that cryoplant limitation sets tflat/(tflat+tdwell) ≈ 25%– For Hybrid operating modes in ITER to provide significant fluence

the cryoplant must be upgraded

Systems Analysis Results

Pursuing 1.5D Integrated Modeling of ITER with TSC/TRANSP Combination

•TRANSP

•Interpretive

•Fixed boundary Eq. Solvers

•Monte Carlo NB and heating

•SPRUCE/TORIC/CURRAY for ICRF

•TORAY for EC

•LSC for LH

•Fluxes and transport from local conservation; particles, energy, momentum

•Fast ions

•Neutrals

Plasma geometryT, n profilesq profile

Accurate source profiles fed back to TSC

•TSC•Predictive•Free-boundary/structures/PF coils/feedback control systems•T, n, j transport with model or data coefficients (, , D, v)•LSC for LH•Assumed source deposition for NB, EC, and ICRF: typically use off-line analysis to derive these

both codes have models for bootstrap current, radiation, sawteeth, ripple loss, pellet fueling, impurities, etc.

TSC evolution treated like an experiment

1.5D ITER Hybrid Simulations Integrate Transport, Heating/CD, and Equilibrium

• Density evolution prescribed, magnitude and profile

• 2% Be + 2% C + 0.12% Ar for high Zeff cases

• GLF23 thermal diffusivities, no rotation stabilization, and with rotation stabilization (plasma rotation from TRANSP assuming = i)

• Prescribed pedestal height and location amended to GLF23 thermal diffusivities

• Control plasma current, radial position, vertical position and shape

• Plasma grown from limited starting point on outboard limiter, early heating required to keep q(0) > 1, keep Pheat < 10 MW

• Control on plasma stored energy, PICRF in controller, PNB not in controller since it is supplying NICD

Using TRANSP Monte Carlo NB and SPRUCE Full Wave/FP ICRF Analysis to

Model ITER Hybrid Sources

IP = 12 MA, PNB = 33 MW, PICRF = 20 MW

Wth = 300 MJWth = 350 MJINB = 2.1 MAINB = 1.8 MA

ICRF HeatingNINB Heating/CD

Source Modeling in TRANSP

ITER’s NBs are large single source beams

Plasma rotation produced by NBs is much lower than present devices

Minority heating with ICRF shows very centralized absorption slightly off axis

Each NB source is 16 MW, although modulation could provide finer power injection

ITER Hybrid at N ≈ 3 Produces 475 MW of Fusion Power

IP = 12 MABT = 5.3 TINI = 7.8 MAN = 2.96n/nGr = 0.93n20(0) = 0.93Wth = 450 MJH98 = 1.6Tped = 9.5 keV∆rampup = 150 V-s

Vloop = 0.025 VQ = 9.43P = 100 MWPaux = 53 MWPrad = 28 MWZeff = 2.25q(0) < 1, ≈ 0.93r(q=1) = 0.45 mli(1) = 0.78Te,i(0) = 30 keV

Available tflattopV-s > 4000 s

Shape control points

High Tped Required to Get N ≈ 3 with GLF23 Core Transport

ITER expected to haveLow vrot (≈ 1/10 vrot

DIII-D)Ti ≈ Te

Low n(0)/<n>

Present Expts haveHigh vrot

Ti > Te

n(0)/<n> > 1.25

Direct extrapolation from present Expts to ITER may be optimistic

Density Peaking Makes Energy Transport Worse with GLF23 Core Transport

GLF23 predicts higher thermal diffusivities for more peaked density case

Flat n()

Peaked n()

Efforts to Benchmark GLF23 Transport in DIII-D 104276 Hybrid Discharge

TSC free-boundary, discharge simulation

DIII-D 104276 dataPF coil currentsTe,i(), n(), v()NB data TRANSP

Use n() directly

TSC derives e, I to reproduce Te and Ti

Turn on GLF23 in place of expt thermal diffusivities

Test GLF23 w/o ExB and w EXB shear stabilization

t = 1.5 s t = 5.0 s

L-mode, i-ITB

H-mode

Profiles from TSC and TVTS and CER data at t = 5 s

TSC Simulation Benchmark of DIII-D

104276 Discharge

TSC Simulation Benchmark of DIII-D 104276 Discharge

Using JSOLVER/BALMSC/PEST2, … to Analyze Ideal MHD Stability of ITER Hybrid

Hybrid discharges operate in a N window

NNTM < N < N

n=1(no wall)

Hybrid discharges have fNI ≥ 40%, from NBCD on-axis and BS off-axis

Hybrid discharges prefer q(0) > 1 or small sawtooth amplitude or possibly small r(q=1)

Examine Porcelli sawtooth model in 1.5D simulations to determine the sawtooth response to small r(q=1), and local dq/dr and dp/dr

N’s up to 3.2 are stable to n=1 w/o conducting wall

ITER Hybrid Scenario Requires High Tped, High n/nGr, and High E

Systems Analysis shows that upgrades to the heat rejection and cryoplant systems will be necessary to achieve long pulses and significant neutron fluence in the Hybrid operating mode

1.5 Discharge Evolution calculations, with GLF23 core transport model, indicate that the Hybrid will require

High Tped ≈ 10 keV (making power to divertor too high)High n/nGr ≈ 0.95High H98(y,2) ≈ 1.6

to reach its operating space of N ≈ 3

Including plasma rotation, determined by TRANSP, does not improve the energy confinement significantly

Present Hybrid experiments have characteristics that give them high performance

Strong plasma rotationTi > Te

Some level of density peakingHowever, these features will be missing in ITER, so we must project with caution to the ITER Hybrid

Verification of the GLF23 core transport model shows reasonable agreement with experimental Hybrid discharges, work is continuing

Ideal MHD calculation indicate the ITER Hybrid discharge simulation cases are stable to n=1 external kink modes without a wall, but they may be unstable to sawteeth, work is continuing