# Presentation: 668956

## FIRE Advanced Tokamak Progress 

- C. Kessel
- Princeton Plasma Physics Laboratory
- NSO PAC 2/27-28/2003, General Atomics

- 0D Operating Space
- PF Coils
- Equilibrium/Stability
- ECCD/Neoclassical Tearing Modes
- RWM Stabilization
- ICRF, FWCD, LHCD
- TSC-LSC Simulations
- 8.    Further Work ---> PVR

## ARIES-AT Provides a Long Term Target for Advanced Tokamaks

- FIRE-AT will need to show _**how close we can get**_ to this configuration thru control in a burning plasma
**Steady state**
**Strong plasma shaping**
**Large bootstrap fraction, minimal CD**
**High ******
**Transport that supports high f****BS**** and high ******
**Plasma edge solution that supports CD, power handling, divertor solution**

## 0D Operating Space Analysis for FIRE AT

- Heating/CD Powers
- ICRF/FW, 30 MW
- LHCD, 30 MW
- Using CD efficiencies
- (FW)=0.20 A/W-m2
- (LH)=0.16 A/W-m2
- P(FW) and P(LH) determined at r/a=0 and r/a=0.75
- I(FW)=0.2 MA
- I(LH)=Ip(1-fbs)
- Scanning Bt, q95, n(0)/<n>, T(0)/<T>, n/nGr, N, fBe, fAr

- Q=5
- Constraints:
- (flattop)/(CR) determined by VV nuclear heat (4875 MW-s) or TF coil (20s at 10T, 50s at 6.5T)
- P(LH) and P(FW) ≤ max installed powers
- P(LH)+P(FW) ≤ Paux
- Q(first wall) < 1.0 MW/m2 with peaking of 2.0
- P(SOL)-Pdiv(rad) < 28 MW
- Qdiv(rad) < 8 MW/m2

## FIRE’s Q=5 AT Operating Space

**Access to higher t****flat****/********j**** decreases at higher ********N, higher Bt, and higher Q, since t****flat**** is set by VV nuclear heating**
**Access to higher radiated power fractions in the divertor enlarges operating space significantly**

## FIRE’s AT Operating Space

**Q = 5-10 accessible**
******N**** = 2.5-4.5 accessible**
**f****bs**** = 50-90+ accessible**
**t****flat****/t****j**** = 1-5 accessible**
**If we can access.....**
**H98(y,2) = 1.2-2.0**
**Pdiv(rad) = 0.5-1.0 P(SOL)**
**Zeff = 1.5-2.3**
**n/n****Gr**** = 0.6-1.0**
**n(0)/<n> = 1.5-2.0**

## Examples of Q=5 AT 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)

## PF Coils Must Sustain AT Plasmas with Low li and High 

- Ip=4.5-5.5 MA, Bt=6.5-8.5T
- 100% non-inductive in flattop
- Ip can not be too low or we’ll loose too many alphas from ripple
- Inductive + non-inductive
- rampup ----> consumes 19-22 V-s, what is final flux state??
- _Flattop times = 16-50s (from P__fusion__ of 300-100 MW) and TF coil_
- Low li(3) = 0.42, N=4.2
- Divertor coils are driven to high currents

## PF Coil Capability for AT Modes

- Advanced tokamak plasmas
- Range of current profiles: 0.35 < li(3) < 0.55
- Range of pressures: 2.50 < N < 5.0
- Range of flattop flux states: chosen to minimize heating and depends on flattop time (determined by Pfusion)
- Ip limited to ≤ 5.5 MA
- _Lower li operating space led to redesign of divertor coils_
- PF1 and PF2 changed to 3 coils and total cross-section enlarged
- Presently examining magnet stresses and heating for AT scenarios

## Neo-Classical Tearing Modes at Lower Bt for FIRE AT Modes

- Bt=6.5 T

- Bt=7.5 T

- Bt=8.5 T

- Ro

- Ro

- Ro

- Ro+a

- Ro+a

- Ro+a

- fce=182

- fce=142

- fce=210

- fce=164

- fce=190

- fce=238

- 

- 

- 

- 

- 170 GHz

- 200 GHz

- _Target Bt=6.5-7 T for NTM control_, to utilize 170 GHz from ITER R&D
- _Must remain on LFS for resonance_
- ECCD efficiency, _can local ____e be high enough to avoid trapping boundary??_

_**Can we rely on OKCD to suppress NTM’s far off-axis on LFS versus ECCD ?? (enhanced Ohkawa affect at plasma edge)**_

- Avoid NTM’s with j profile and q>2.0
- or do we need to suppress them??

## J. Decker, APS 2002,MIT

**OKCD allows LFS EC deposition, with similar A/W as ECCD on HFS **

## Comments on ECCD in FIRE

- ASDEX-U shows that 3/2 island is suppressed for about 1 MW of power with IECCD/Ip = 1.6%, giving 0.013 A/W
- Ip=0.8 MA and N=2.5
- DIII-D shows that 3/2 island is suppressed for about 1.2-1.8 MW with jEC/jBS = 1.2-2.0
- Ip=1.0-1.2 MA, N=2.0-2.5
- OKCD analysis of Alcator-CMOD gives about 0.0056 A/W
- FIRE’s current requirement should be about 15 times higher than ASDEX-U (scaled by Ip and N2)
- Need about 200 kA, which would require about 35 MW?? Early detection reduces power alot according to ITER
- Do we need less current for 5/2 or 3/1, do we need to suppress them??
- Is 170 GHz really the cliff in EC technology??

- MIT, short pulse results

## Updating AT Equilibrium Targets Based on TSC-LSC Equilibrium

**TSC-LSC equilibrium**
- Ip=4.5 MA
- Bt=6.5 T
- q(0)=3.5, qmin=2.8
- N=4.2, =4.9%, p=2.3
- li(1)=0.55, li(3)=0.42
- p(0)/p=2.45
- n(0)/n=1.4
- Stable n=
- Stable n=1,2,3 with no wall

- √V/Vo

## Original AT Target Equilibrium for FIRE 

-  = 3.65, fbs < 0.75

- N = 2.5, fbs < 0.55

- q(min) = 2.1-2.2
- r/a(qmin) = 0.8
- n(0)/<n> = 1.5
- Ip = 5.5 MA
- Bt = 8.5 T
- No wall stabilization
- N = 2.5
- n=1 RWM stabilized
- N = 3.65

**This needs to be revisited with Ip=4.5 MA and Bt=6.5 T**

## Stabilization of n=1 RWM is a High Priority on FIRE

- Feedback stabilization analysis with VALEN shows strong improvement in , _taking advantage of __DIII-D__ experience, _most recent analysis indicates N(n=1) can reach 4.2_ _

_**What is impact of n=2??**_

## Ideal wall n=1 limit

**Ideal wall n=1 limit**

## FIRE Uses ICRF Ion Heating for Its Reference and AT Discharges

- ICRF ion heating
- 80-120 MHz
- 2 strap antennas
- 4 ports (2 additional reserved)
- 20 MW installed (10 MW additional reserved)
- He3 minority and 2T heating
- Frequency range allows heating at a/2 on HFS and LFS (C-Mod ITB)

- Full wave analysis
- SPRUCE in TRANSP
- Using n(He3)/ne = 2%
- n20(0) = 5.3, <n20> = 4.4
- PICRF = 11.5 MW,  = 100 MHz
- THe3(0) = 10.2 keV
- Pabs(He3) = 60%
- Pabs(T) = 10%
- Pabs(D) = 2%
- Pabs(elec) = 26%

- Antenna design --->D. Swain, ORNL

## ICRF/FW Viable for FIRE On-Axis CD

- _PICES (ORNL) and CURRAY(UCSD) analysis _
- f = 110-115 MHz
- n|| = 2.0
- n(0) = 5x10^20 /m3
- T(0) = 14 keV
- 40% power in good part of spectrum (2 strap)
- ---->  0.02-0.03 A/W
- CD efficiency with 4 strap antennas is 50% higher
- Operating at lower frequency to avoid ion resonances, vph/vth??

- Calculations assume same ICRF ion heating system frequency range, approximately 40% of power absorbed on ions, can provide required AT on-axis current of 0.3-0.4 MA with 20 MW (2 strap antennas)

- E. Jaeger, ORNL

## Benchmarks for LHCD Between LSC and ACCOME (Bonoli)

- Trapped electron effects reduce CD efficiency
- Reverse power/current reduces forward CD
- Recent modeling with CQL and ACCOME/LH19 will improve CD efficiency, but right now........
- Bt=8.5T ----> 0.25 A/W-m2
- Bt=6.5T ----> 0.16 A/W-m2
_**FIRE has increased the LH power from 20 to 30 MW**_

## HFS Pellet Launch and Density Peaking ---> Needs Strong Pumping

- FIRE reference discharge with uniform pellet deposition, achieves n(0)/<n> ≈ 1.25

- Simulation by W. Houlberg, ORNL, WHIST

- P. T. Lang, J. Nuc. Mater., 2001, on ASDEX and JET
- L. R. Baylor, Phys. Plasmas, 2000, on DIII-D

## HFS Launch

**HFS Launch**
**V=125 m/s, set by ORNL pellet tube geometry**
**Vertical and LFS launch access higher velocities **

## TF Ripple and Alpha Particle Losses

- _TF ripple very low in FIRE_
- (max) = 0.3% (outboard midplane)
- Alpha particle collisionless + collisional losses = 0.3% for reference ELMy H-mode
- For AT plasmas alpha losses range from 2-8% depending on Ip and Bt
- ----> are Fe inserts required for AT operation??? _Optimize for Bt=6.5T_

## Fe Shims for Ripple Reduction in FIRE

**TF Coil**

**Outer VV**

**Inner VV**

**Fe Shims**

## TSC-LSC Simulation of Burning AT Plasma in FIRE

- Bt=6.5 T, Ip=4.5 MA
- q(0) =4.0, q(min) = 2.75, q(95) = 4.0, li = 0.42
-   = 4.7 %, N = 4.1, p = 2.35
- n/nGr = 0.85, n(0)/<n> = 1.47
- n(0) = 4.4x10^20, n(line) = 3.5, n(vol) = 3.0
- 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**

## TSC-LSC Simulation of Q=5 Burning AT Plasma

- Ip=4.5 MA, Bt=6.5 T, N=4.1,
- t(flat)/j=3, I(LH)=0.80, P(LH)=25 MW
- fBS=0.77, Zeff=2.3,

## TSC-LSC Simulation of Q=5 AT Burning Plasma

## TSC-LSC Simulation of Q=5 AT Burning Plasma

## AT Physics Capability on FIRE

- Strong plasma shaping and control
- Pellet injection, divertor pumping, impurity injection
- FWCD (electron heating) on-axis, ICRF ion heating off-axis
- LHCD (electron heating) off-axis
**ECCD (LFS, electron heating) off-axis, MHD control**
- RWM MHD feedback control
**NBI ?? (need to examine for AT parameters!!)**
- t(flattop)/t(curr diff) = 1-5
**Diagnostics**

_**Control**_

- MHD
- J Profile
- P-profile
- Rotation

## Ongoing Advanced Tokamak Work ---> PVR

**Establish PF Coil operating limits**
**Revisit Equilibrium/Stability Analysis**
**Use recent GLF23 update in AT scenarios**
- LHCD efficiency updates
**EC with FIRE’s parameters**
**Orbit calculations of lost alphas for scenario plasmas**
- RWM coil design in port plugs and RF ports
- Determine possible impact of n=2 RWM on access to high N
- Examine NBI for FIRE AT parameters