# Presentation: 045930

## Analysis and Simulations of the ITER Hybrid Scenario

- C. Kessel, R. Budny, K. Indireshkumar
- Princeton Plasma Physics Laboratory, USA
***ITPA Topical Group on Steady State Operation and Enhanced Performance***
- Centro di Cultura Scientifica - A. Volta
- Societa del Casino, Como Italy, May 2005

## Contents

- Groundrules for Hybrid Scenario
- Goals
- Plasma parameters and operating modes
- Constraints
- 0D Systems analysis of Hybrid operating points
- Brief description
- Major parameters and effects of constraints
- Large scan and operating space
- 1.5 D simulations of the Hybrid Scenario with TSC-TRANSP
- Brief TSC-TRANSP description
- TSC simulation results with GLF23 energy transport, comments
- TRANSP source modeling
- Benchmarking of GLF23 energy transport DIII-D 104276

## Hybrid Scenario in ITER

- Plasma parameter ranges
- E ≈ 1.0-1.5  E98(y,2)
- 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
- Maximize fNI, Te(0) and avoid high Pfusion
- 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)

## Hybrid Scenario in ITER

- Constraints
**Fusion power** **vs pulse length**: 350 MW -- 3000s, 500 MW -- 400 s, 700 MW -- 150s
**Installed auxiliary heating/CD power**
**Divertor heat loads**: allowable of 20 MW/m2, leading to ≈ 10-12 MW/m2 conduction heat load to account for radiation and transients
**Divertor radiation??**: 15% of power entering Scrape Off Layer is assumed radiated in divertor slots or about the X-point
**Core radiation** requirement to meet divertor heat load: 2% Be only (Zeff = 1.3) unacceptable, 2% Be + 2% C + 0.12% Ar  (Zeff= 2.2) is acceptable
- Within **volt-second capability** of OH+PF: maximum of 300 V-s, with 10 V-s in breakdown, and about 15-20 V-s spare, based on reference H-mode scenario
**First wall surface heat flux??:**  0.5 MW/m2, with peaking factor of 2.0, leading to 0.25 MW/m2 average

## 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-Neilsen 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

## ITER Hybrid Systems Analysis

| n | n | T | T | N | q95 | Ip | HH | fGr | fBS | fNI | Zeff | fBe | fC | fAr | t/J |
| --- | --- | --- | --- | --- | --- | --- | --- | --- | --- | --- | --- | --- | --- | --- | --- |
| 0.40 | 0.57 | 1.3 | 12.8 | 1.8 | 4.38 | 12.0 | 1.27 | 0.65 | 0.26 | 0.50 | 1.82 | 1% | 0% | .2% | 13.4 |
| 0.05 | 0.66 | 1.3 | 13.1 | 2.0 | 4.38 | 12.0 | 1.42 | 0.70 | 0.25 | 0.45 | 1.80 | 1% | 2% | 0% | 12.2 |
| 0.40 | 0.66 | 2.0 | 13.5 | 2.0 | 4.38 | 12.0 | 1.40 | 0.75 | 0.30 | 0.51 | 2.36 | 3% | 1% | .2% | 13.5 |
| 0.05 | 0.69 | 2.0 | 11.2 | 2.0 | 4.79 | 11.0 | 1.40 | 0.80 | 0.27 | 0.48 | 1.93 | 2% | 0% | .2% | 15.5 |
| 0.40 | 0.76 | 0.6 | 10.1 | 2.0 | 4.79 | 11.0 | 1.29 | 0.95 | 0.33 | 0.52 | 1.61 | 2% | 0% | .1% | 16.8 |
| 0.40 | 0.76 | 1.3 | 10.4 | 2.0 | 4.79 | 11.0 | 1.29 | 0.95 | 0.33 | 0.52 | 1.73 | 3% | 1% | 0% | 16.8 |
| 0.05 | 0.86 | 1.3 | 10.6 | 2.3 | 4.79 | 11.0 | 1.46 | 1.00 | 0.30 | 0.47 | 2.03 | 3% | 2% | 0% | 15.3 |

- BT = 5.3 T, PCD = 33 MW, Paux < 39 MW, Pfusion = 350 MW, Nw = 0.49 MW/m2, tflattop = 3000 s,
- chosen to meet Nw x tflattop > 1475 MW-s/m2

******n**** = 0.05 --> n(0)/<n> = 1.04, ********n**** = 0.4 --> n(0)/<n> = 1.25**
******T**** = 0.60 --> T(0)/<T> = 1.50, ********T**** = 2.0 --> T(0)<T> = 2.50**

## Results

**Fusion power pulse length limitation is most significant** factor in determining Hybrid operating space
- Lowering density does not continuously lead to better operating points
- Higher H98(y,2) allows access to higher fluence and lower n/nGr
**High fusion power is not necessary** or desirable
- Only low N ≈ 2 operating points are required
**Volt-seconds capability** appears to be enough to offer **few thousand second flattops**
**Divertor heat load limits is next most significant** factor for Hybrid operating space
- Combination of conducted power, power radiated in divertor, transient conducted power, and core radiated power
**First wall surface heat load** **limits do not appear to be limiting**
**Available 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%

## Flow Diagram of TSC-TRANSP 1.5D Analysis Combining Strengths of the Two Codes

**TSC**

**TRANSP**

- Discharge simulation with assumed source profiles and evolving boundary

- Plasma geometry
- T, n profiles
- q profile

- Interpretive rerun of discharge simulation with source models, fast ions, neutrals (TSC as expt.)

- Accurate source profiles fed back to TSC

- Analysis with interfaces to TRANSP

- Analysis with interfaces to TSC

**Flow Diagram of TSC-TRANSP 1.5D Analysis Combining Strengths of the Two Codes**

## TSC and TRANSP, a Few* Attributes

_**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 (benchmark with other LH codes)
- Assumed P and j deposition for NB, EC, and ICRF: typically use off-line analysis to derive these

- *In addition, both codes have models for bootstrap current, radiation, sawteeth, ripple loss, pellet fueling, impurities, etc.
- ** TRANSP has predictive capability

## TRANSP NBCD Results for Various Conditions in the ITER Hybrid Simulations, t = 500 s

**TRANSP NBCD Results for Various Conditions in the ITER Hybrid Simulations, t = 500 s**

- IP = 12 MA, PNB = 33 MW, PICRF = variable, ≤ 20 MW

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

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

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

**I****NB**** = 2.4 MA**
**I****NB**** = 2.1 MA**

**I****NB**** = 2.2 MA**
**I****NB**** = 2.1 MA**

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

## ICRF He3 Minority Heating Used as Heating Source to Allow NINB to Drive Current

**ICRF He3 Minority Heating Used as Heating Source to Allow NINB to Drive Current**

- fICRF = 52.5 MHz
- nHe3 = 2% nDT
- EHe3 up to 120 keV

## TSC Simulation Description

- Density evolution prescribed, magnitude and profile
- Impurity is 2% Be for reference, and 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 momentum = i)
- Prescribed pedestal 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

## TSC ITER Hybrid Scenario

- IP = 12 MA, BT = 5.3 T, Vsurf = 0.05V, q(0) = 0.99, q95 = 3.95, li(1) = 0.8,,
- t = 2.2%, n/nGr = 0.79, Wth = 300 MJ, **n****20****(0) = 0.77**, n(0)/<n> = 1.05,
- N = 2.0, **H****98(y,2)**** = 1.33**, Te,i(0) = 22.5 keV, Te,i(0)/<T> = 2.0,  = 1.83,  = 0.46,
- ∆rampup = 150 V-s, P = 65 MW, Paux = 35 MW, PNINB = 33 MW, **Z****eff**** = 1.3**,
**I****NI**** = 5.3 MA, I****BS**** = 3 MA, I****NINB**** = 2.2**, **T****ped**** = 7.5-8 keV**

**GLF23, no stab.**

## TSC ITER Hybrid Scenario

## TSC ITER Hybrid Scenario

**GLF23, no stabilization**

- Shape control points

## Variation of Tped With GLF23 no stabilization

**Variation of T****ped**** With GLF23 no stabilization**

## Variation of Tped with GLF23 with ExB shear stabilization 

**Variation of Tped with GLF23 with ExB shear stabilization **

- TRANSP plasma rotation assuming mom = i

- Lost Wth control

## Benchmark of GLF23 Transport in DIII-D 104276 Hybrid Discharge

**Benchmark of 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 Te and Ti
- Turn on GLF23 in place of expt thermal diffusivities
- Test GLF23 w/o ExB and w EXB shear stabilization

## TSC Simulation Benchmark of DIII-D 104276 Discharge

**TSC Simulation Benchmark of DIII-D 104276 Discharge**

## GLF23 w/o EXB (or -stab) Shows Lower T(0) Values Than Those in Expt

- No -stabilization

- GLF23 turned on

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

**GLF23 w/o EXB (or ********-stab) Shows Lower T(0) Values Than Those in Expt**

- However, these do not agree with GLF23 analysis presented by Kinsey at IAEA
- Cases with ExB shear and -stabilization have not been completed

## Conclusions

- Based on GLF23 no stab.  energy transport, pedestal temperatures appear high (>6.5 keV) to obtain good performance
- Including EXB shear stabilization in GLF23, with velocity from TRANSP assuming mom = I, does not improve the situation
- Higher density operating points can improve this
- Directly applying experimental Hybrid discharge characteristics to ITER may be optimistic
- Lower rotation in ITER by 10x
- Ti ≈ Te in ITER
- Density peaking in expts.  from NB fueling and ExB shear or other particle transport effects, may not exist in ITER
- Application of GLF23 to full discharge simulations is continuing
- No stabilization application is robust in TSC
- With stabilization has start up difficulties, are being resolved
- Will apply to L-mode and H-mode phases
- Benchmark simulations with hybrid discharges is continuing

## (Possible) Future Work

- Determine engineering constraints for use in 0D systems analysis in greater detail
- Is the cryoplant limitation for real??
- Complete higher density and higher Zeff Hybrid 1.5D scenarios
- Examine slight density peaking in 1.5D scenarios
- Turn off stored energy control, examine Q vs. Tped vs. vExB
- Examine higher velocities in ExB shear stabilized cases
- Examine strategies for coupled stored energy and NICD feedback control
- Use alternative energy transport models, less stiff models to see the dependence on required pedestal temperatures
- Expand benchmark simulations to other Hybrid discharges in DIII-D
- Further development of TSC-TRANSP modeling
- Etc.