Simulation and Analysis of the Hybrid Operating Mode in ITER

1
Simulation and Analysis of the Hybrid Operating
Mode in ITER

• C. Kessel, R. Budny, and K. Indireshkumar
• Princeton Plasma Physics Laboratory
• Symposium On Fusion Engineering
• Knoxville, TN
• September 26-29, 2005

2
So What is a Hybrid Operating Mode in ITER?
Reference H-mode Ip 15 MA BT 5.3 T R 6.2
m a 2.0 m Vloop 0.09 q95 3 ?N
1.8 H98(y,2) 1.0 q(0) 1.0(rsaw 1 m) Q
10 Tflattop 500 s
Hybrid Mode Ip 12 MA BT 5.3 T R 6.2 m a
2.0 m Vloop 0.025-0.04 q95 4 ?N
3.0 H98(y,2) 1.5 q(0) 1.0(rsaw small) Q
5-10 Tflattop 3000 s
Steady State (AT) Mode Ip 9 MA BT 5.3 T R
6.35 m a 1.85 m Vloop 0.0 q95 4 ?N
3-4.5 H98(y,2) 1.6 q(0) gt 1.5-2.0 Q
5 Tflattop 8
3
Hybrid Scenario in ITER

• Plasma parameter ranges
• ?E 1.0-1.5 ? ?E98(y,2)
• ?NNTM lt ?N lt ?Nno 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 fNIoff-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)

4
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, NTMs

5
0D Systems Analysis Identifies Device Constraints
for Scenario Simulations

• ITERs Primary Device Limitations That Affect
  Scenarios
• Fusion power vs pulse length ----gt heat
  rejection system
• 350 MW for 3000 s
• 500 MW for 400 s
• 700 MW for 150 s ----gt (maximum Pfus cryoplant
  limits)
• Divertor conducted heat load, maximum gt 20 MW/m2,
  nominal 5-10 MW/m2 ----gt 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 ----gt 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) ----gt
  cryoplant for SC coils
• Limited to about 25

What device upgrades are required for advanced
operating modes, and are they major or minor
upgrades?
6
0D Operating Space Analysis
IP 12 MA BT 5.3 T R 6.2 m A 3.1 ?95
1.75 ?95 0.5 ?P/?E 5 ??total 300
V-s ??breakdown 10 V-s li 0.80 CE
0.45 ?NBCD 0.3 x 1020 A/W-m2 PCD 33 MW ?T
1.75, Ta/To 0.1 ?n 0.075, na/no 0.3 fBe
2.0 1.5 ?N 3.0 0.4 n/nGr 1.0 3.0 Q
12.0 0.0 fC 2.0 0.0 fAr 0.2
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 (benchmar
ked with TSC) T(r) (To - Ta)1-(r/a)2?T
Ta Same for density profile Etc.
Input parameters
Scanned parameters
7
ITER Hybrid Systems Analysis
Fusion power pulse length limitation
significantly reduces accessible fluence values,
and changes dependence on density
8
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
9
ITER Hybrid Systems Analysis
Increasing the power radiated in the divertor can
recover operating space at lower conducted peak
heat flux
10
ITER Hybrid Systems Analysis
Large Operating Space Scan 1.05 n(0)/?n?
1.25 1.5 T(0)/?T? 2.5 11.0 IP (MA)
13.0 1.5 ?N 3.0 0.4 n/nGr 1.0 3.0 Q
12.0 1 fBe 3 0 fC 2 0 fAr
0.2 Other input fixed at previous values
11
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

12
Systems Analysis Results

• 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/(tflattdwell) 25
• For Hybrid operating modes in ITER to provide
  significant fluence the cryoplant must be upgraded

13
Pursuing 1.5D Integrated Modeling of ITER with
TSC/TRANSP Combination
Plasma geometry T, n profiles q profile

• 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

• 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

TSC evolution treated like an experiment
Accurate source profiles fed back to TSC
both codes have models for bootstrap current,
radiation, sawteeth, ripple loss, pellet fueling,
impurities, etc.
14
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) gt 1, keep Pheat lt 10 MW
• Control on plasma stored energy, PICRF in
  controller, PNB not in controller since it is
  supplying NICD

15
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
NINB Heating/CD
ICRF Heating
Wth 300 MJ Wth 350 MJ
INB 2.1 MA INB 1.8 MA
16
Source Modeling in TRANSP
Each NB source is 16 MW, although modulation
could provide finer power injection
ITERs NBs are large single source beams
Minority heating with ICRF shows very centralized
absorption slightly off axis
Plasma rotation produced by NBs is much lower
than present devices
17
ITER Hybrid at ?N 3 Produces 475 MW of Fusion
Power
IP 12 MA BT 5.3 T INI 7.8 MA ?N
2.96 n/nGr 0.93 n20(0) 0.93 Wth 450 MJ H98
1.6 Tped 9.5 keV ??rampup 150 V-s
Vloop 0.025 V Q 9.43 P? 100 MW Paux 53
MW Prad 28 MW Zeff 2.25 q(0) lt 1,
0.93 r(q1) 0.45 m li(1) 0.78 Te,i(0) 30 keV
Available tflattopV-s gt 4000 s
18
Shape control points
19
High Tped Required to Get ?N 3 with GLF23 Core
Transport
ITER expected to have Low vrot ( 1/10
vrotDIII-D) Ti Te Low n(0)/ltngt Present Expts
have High vrot Ti gt Te n(0)/ltngt gt 1.25 Direct
extrapolation from present Expts to ITER may be
optimistic
20
Density Peaking Makes Energy Transport Worse with
GLF23 Core Transport
GLF23 predicts higher thermal diffusivities for
more peaked density case
Flat n(?) Peaked n(?)
21
Efforts to Benchmark GLF23 Transport in DIII-D
104276 Hybrid Discharge
t 1.5 s
t 5.0 s
TSC free-boundary, discharge simulation DIII-D
104276 data PF coil currents Te,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
L-mode, i-ITB
H-mode
22
TSC Simulation Benchmark of DIII-D 104276
Discharge
Profiles from TSC and TVTS and CER data at t 5 s
23
TSC Simulation Benchmark of DIII-D 104276
Discharge
24
Using JSOLVER/BALMSC/PEST2, to Analyze Ideal
MHD Stability of ITER Hybrid
?Ns up to 3.2 are stable to n1 w/o conducting
wall
Hybrid discharges operate in a ?N window ?NNTM lt
?N lt ?Nn1(no wall) Hybrid discharges have fNI
40, from NBCD on-axis and BS off-axis Hybrid
discharges prefer q(0) gt 1 or small sawtooth
amplitude or possibly small r(q1)
Examine Porcelli sawtooth model in 1.5D
simulations to determine the sawtooth response to
small r(q1), and local dq/dr and dp/dr
25
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 0.95 High
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
26
Present Hybrid experiments have characteristics
that give them high performance Strong plasma
rotation Ti gt Te Some level of density
peaking However, 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 n1 external kink modes without a wall, but
they may be unstable to sawteeth, work is
continuing