# Presentation: 045929

## 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**

## 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) > 1.5-2.0
- Q = 5
- Tflattop = ∞

## Hybrid Scenario in ITER

**Plasma parameter ranges**
- E ≈ 1.0-1.5  E98(y,2)
- NNTM < N < 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)

## 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, f****NI****, 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 P****fus**** 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 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%

**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 ≤ 1.25
- 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

## 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

- 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 geometry**
**T, n profiles**
**q 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 Z****eff**** 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 P****heat ****< 10 MW**
** ****Control on plasma stored energy, P****ICRF**** in controller, P****NB**** 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

**W****th**** = 300 MJ**
**W****th**** = 350 MJ**

**I****NB**** = 2.1 MA**
**I****NB**** = 1.8 MA**

**ICRF Heating**

**NINB Heating/CD**

## Source Modeling in TRANSP

**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

**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/n****Gr**** = 0.93**
**n****20****(0) = 0.93**
- Wth = 450 MJ
**H****98**** = 1.6**
**T****ped**** = 9.5 keV**
- ∆rampup = 150 V-s

- Vloop = 0.025 V
- Q = 9.43
**P******** = 100 MW**
**P****aux**** = 53 MW**
**P****rad**** = 28 MW**
- Zeff = 2.25
**q(0) < 1, ≈ 0.93**
**r(q=1) = 0.45 m**
- li(1) = 0.78
- Te,i(0) = 30 keV

**Available t****flattop****V-s**** > 4000 s**

## Shape control points

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

**High T****ped**** Required to Get ********N**** ≈ 3 with GLF23 Core Transport **

**ITER expected to have**
**Low v****rot**** (≈ 1/10 v****rot****DIII-D****)**
**T****i**** ≈ T****e**
**Low n(0)/<n>**
**Present Expts have**
**High v****rot**
**T****i**** > T****e**
**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 

**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

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

**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 T****e**** and T****i**
**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**

## TSC Simulation Benchmark of DIII-D 104276 Discharge

- 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

**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
******N****NTM**** < ********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 T****ped**** ≈ 10 keV (making power to divertor too high)**
**High n/n****Gr**** ≈ 0.95**
**High H****98(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 rotation**
**T****i**** > T****e**
**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 n=1 external kink modes without a wall**, but they may be unstable to sawteeth, work is continuing