# Presentation: 546110

## FIRE Physics Basis(detailed version)

**C. Kessel for the FIRE Team**
*Princeton Plasma Physics Laboratory*
_**FIRE Physics Validation Review**_
_**March 30-31, 2004**_
_**Germantown, MD **_

## FIRE Description

_**H-mode**_
**I****P**** = 7.7 MA**
**B****T**** = 10 T**
******N**** = 1.80**
****** ****= 2.4%**
******P**** = 0.85**
************** ****= 0.075%**
**q(0) < 1.0**
**q****95**** ≈ 3.1**
**li(1,3) = 0.85,0.66**
**T****e,i****(0) = 15 keV**
******T****e,i******** = 6.7 keV**
**n****20****(0) = 5.3**
**n(0)/********n******** = 1.15**
**p(0)/********p******** = 2.4**
**n/n****Gr**** = 0.72**
**Z****eff**** = 1.4**
**f****bs**** = 0.2**
**Q = 12**
******burn**** = 20 s **

**R = 2.14 m, a = 0.595 m, ********x**** = 2.0, ********x**** = 0.7, P****fus**** = 150 MW**

_**AT-Mode**_
**I****P**** = 4.5 MA**
**B****T**** = 6.5 T**
******N**** = 4.2**
****** ****= 4.7%**
******P**** = 2.35**
************** ****= 0.21%**
**q(0) ≈ 4.0**
**q****95****, q****min**** ≈ 4.0,2.7**
**li(1,3) = 0.52,0.45**
**T****e,i****(0) = 15 keV**
******T****e,i******** = 6.8 keV**
**n****20****(0) = 4.4**
**n(0)/********n******** = 1.4**
**p(0)/********p******** = 2.5**
**n/n****Gr**** = 0.85**
**Z****eff**** = 2.2**
**f****bs**** = 0.78**
**Q = 5**
** ********burn**** = 32 s**

- port

- plasma

- divertor

- baffle

- passive
- plate

- VV

## FIRE Magnet Layout

**TF Coil**

**CS1**

**CS2**

**CS3**

**PF1,2,3**

**PF4**

**PF5**

**Error field correction**
** ****coils**

**Fast vertical and radial position control coil**

**RWM feedback coil**

**Fe shims**

## Toroidal Field Coils

**16 TF coils**
**BeCu inboard legs**
**OFHC Cu outboard legs**
**Coil stress and heating limit TF pulse length (factor of ≥ 1.18 over ******** allowable)**
**H-mode**
**B****T**** = 10 T and P****fus**** = 150 MW**
**---->  20 s flattop**
**AT-mode**
**B****T**** = 6.5 T and P****fus**** = 150 MW**
**---->  48 s flattop**
**Maximum TF ripple at R+a is 0.3%**
**0.3% ********-particle loss H-mode**
**8% ********-particle loss AT-mode**
**Expect to use Fe shims for AT-mode**

## Poloidal Field Coils

**Center Stack 1, 2U&L, 3U&L**
**PF1,2,3,4,5 U&L**
**All CS and PF coils are CuCrZr**
_**H-mode**_
**Fiducial equilibria in discharge**
**IM, SOD, SOH, SOB, EOB, EOH, EOD**
**Flexibility of PF coils**
**0.55 ≤ li(3) ≤ 0.85 (SOB,EOB)**
**0.85 ≤ li(3) ≤ 1.15 (SOH,EOH)**
******ref****-5 ≤ ********(Wb) ≤ ********ref****+5**
**1.5 ≤ ********N**** ≤ 3.0**
**Full operating space available within stress (1.3 margin) and heating allowables, except at EOB, li=0.85 where ******** ≤ ********ref****-2**
**Rampup consumes ≈ 40 V-s**
**Flattop consumes ≈ 3 V-s**

**CS1**

**CS2**

**CS3**

**PF1,2,3**

**PF4**

**PF5**

## Poloidal Field Coils

_**AT-mode**_
**Fiducial equilibria in discharge**
**IM, SOD, SOH, SOF, SOB, EOB, EOH, EOD**
**Flexibility of PF coils**
**0.35 ≤ li(3) ≤ 0.65 (SOB & EOB)**
**2.5 ≤ ********N**** ≤ 5.0**
**7.5 ≤ ********flattop****(Wb) ≤ 17.5**
**Full operating space is available for Ip ≤ 5.0 MA**
**PF coils can provide pulse length limitation -----> 41 s for access to op. space at Ip = 4.5 MA, and scales with Ip, li, ********p****, and ******** **
**Inductive + non-inductive rampup consumes 19-22 V-s, final state can be optimized**
**Plasma current 100% non-inductive in flattop**

- shape control feedback points

## Poloidal Field Coils

_**TSC simulations**_
**Free-boundary calculations with heating, CD, bootstrap current, energy and current transport, impurities, PF coils, structure and feedback systems, etc.**
**--->  check of equilibrium coil currents**
**---> Volt-second consumption**
**---> Feedback control of vertical position, radial position, plasma current and shape**

## Vertical Stability

**Design passive structures to slow vertical instability for feedback control and provide a stability factor f****s**** > 1.2**
**Passive stabilizers are 2.5 cm thick Cu, toroidally continuous on upper outboard and inboard sides**
**For most unstable plasmas (full elongation and low pressure), over the range 0.7 < li(3) < 1.1, the stability factor is 1.3 < f****s**** < 1.13 and growth time is 43 < ********g****(ms) < 19**

**Passive **
**stabilizers**

**Cladding**
**(ports provide poloidal cuts)**

**Cladding**
**(large number of poloidal cuts)**

## Internal Control Coils

**8 OFHC Cu coils (2nd redundant coils) above and below the midplane**
**Fast vertical position feedback control**
******Z****RMS**** = 1 cm, 65-90 kA-turn, 50-75 V/turn ------> 7-14 MVA (peak)**
**Fast radial position feedback control (antenna coupling)**
**Analysis not completed, assuming I and V similar to vertical control**
**Fast radial feedback is coupled with slower outer PF shape control**
**These coils also used in startup to tailor field null**

## Resistive Wall Mode (RWM) Coils

- ICRF Port Plug

- RWM Coil

_**DIII-D experience**_
**Modes are detectable at the level of 1G**
**The C-coils can produce about 50 times this field**
**The necessary frequency depends on the wall time for the n=1 mode (which is 5 ms in DIII-D) and they have ********wall ****≈ 3**
_**FIRE projection**_
**FIRE has approximately 3-4 times the DIII-D plasma current, so we might be able to ****measure down to 3-4 G**
**If we try to guarantee at least 20 times this value from the feedback coils, we must produce ****60-80 G at the plasma**
**These fields require approximately ****I = f(d,Z,********)B****r****/********o**** = 5-6.5 kA**
**Assume we also require ********wall ****≈ 3**
**Required voltage would go as ****V ≈ 3********o****(2d+2Z)NI/********wall ****≈ 0.25 V/turn**

## Error Field Correction Coils

**Static or slow dynamic Cu coils**
**Located outside TF and PF coils**
**Compensating TF and PF coil/lead/etc. misalignments and other under field conditions**
**These coils are NOT used for RWM feedback**
**Extrapolated threshold to induce locked modes ≈ 1 ******** 10****-4**** T (****very**** ****uncertain****!)**
**Correction coils should be capable of reducing (m=1,n=1), (2,1), and (3,1) error fields, simultaneously**
**And provide factor of ≈ 5 reduction in net error field B****r2,1****net**

- 3 distributed coils provides poloidal mode control allowing multiple (m,n=1) suppression

**Recent C-Mod data shows that applied B****r2,1**** of 6********10****-4**** T removed mode-locking ----->  Important since C-Mod does NOT have external rotation source**

_**No analysis performed**_

**ITER Error Coils**

## ICRF Heating and CD

**Frequency range 70-115 MHz**
**2 strap antennas**
**4 ports, total power 20 MW**
_**H-mode**_
**B****T**** = 10 T, minority He3 and 2T at 100 MHz**
**Frequency range allows heating at a/2 on HFS and LFS**
_**AT-mode**_
**B****T**** = 6.5 T, ion heating at minority H and 2D at 100 MHz**
**Frequency range allows ion heating at a/2 on HFS and LFS**
**Electron heating/CD at 70-75 MHz**
**CD efficiency ********20**** = 0.14-0.21 A/W-m****2**** **

- SPRUCE analysis
- nHe3/ne=2%
- PICRF=11.5 MW
- =100 MHz
- THe3(0)=10.2 keV
- PHe3=60%
- PT=10%
- PD=2%
- Pe=26%

## ICRF Heating and CD

**B****T**** = 10 T**

**B****T**** = 6.5 T**

**Vacuum Toroidal Field Resonances**

**Be**

**Be**

## ICRF Heating and CD

**Want to reduce power required to drive on-axis current**
**2 strap antenna and port geometry provides only 40% of ICRF power in good CD part of the spectrum**
**4 strap antenna can provide 60% of power in good CD part of spectrum**
**Expanding antenna cross-section and going to 4 straps reaches 80% in good CD part of spectrum**

## ICRF Heating and CD

**AORSA full wave analysis continues including fast alpha and Be impurity effects**

**75 MHz**

**70 MHz**

**P****e****=0.44**
**P****T****=0.15**
**P****Be****=0.30**

**P****e****=0.65**
**P****T****=0.32**
**P****Be****=0.0**

******20****=0.14**

******20****=0.17**

## Lower Hybrid Current Drive

**Frequency 5 GHz**
**Spectrum n****|| ****≈ 1.8-2.5, ********n****||**** = 0.3**
**Power of 30 MW, in 2 ports**
**Upgrade to baseline design**
_**H-mode**_
**Used for NTM control for B****T**** = 10 T**
**Used for non-inductive CD for hybrid discharges**
_**AT-mode**_
**Used for bulk CD for B****T**** = 6.5 T**
**CD efficiency ********20**** ≈ 0.16 A/W-m****2**** at 6.5 T (30-50% higher from 2D FP calcs.)**
**Used for NTM control**

**f  > 2********f****LH**
**RF power flux is 53 MW/m****2**
**Need 0.57 m****2**** per waveguide for 30 MW**
**Each waveguide is 5.7 cm(tall) ******** 0.65 cm**
**Have ≈ 1500 waveguides**

## Lower Hybrid Current Drive

**Trapped electron effects reduce CD efficiency**
**Reverse power/current reduces forward CD**
**Less than 1.0 MW is absorbed by alphas**
**Recent modeling with CQL and ACCOME/LH19 improves CD efficiency, 30-50% increase, but right now........**
**B****T**** = 8.5T ----> 0.25 A/W-m2**
**B****T**** = 6.5T ----> 0.16 A/W-m2**
**Benchmarks with ACCOME, CURRAY and LSC**
**3.7 GHz, 750 kW, 1000s sources available**
**ITER estimate for 5 GHz, 1.0 MW, CW sources was 1.15 euro/watt**

**ACCOME**

**TSC-LSC**

## Electron Cyclotron

- =ce=170 GHz

- pe=ce

**Rays are bent as they approach ******** = ********pe**
**Rays are launched with toroidal directionality for CD**

**Frequency of 170 GHz to utilize ITER R&D**
**LFS, O-mode, fundamental**
**FIRE has high density and high field**
**Cutoff of EC when ******** = ********pe**
_**AT-mode**_
**Lower B****T**** = 6.5 T**
**LFS deposition implies trapping reduction of CD, however, Ohkawa effect provides more CD than standard EC**
**Current required, scaled as Ip********N****2**** from DIII-D and ASDEX-U expts for (3,2) mode**
**----> drive 200 kA to suppress from saturated state**
**----> requires 100 MW!**

## Electron Cyclotron

- Bt=6.5 T

- Bt=7.5 T

- Bt=8.5 T

- Ro

- Ro+a

- fce=182

- fce=142

- fce=210

- fce=164

- fce=190

- fce=238

- 

- 

- 

- 

- 170 GHz

- 200 GHz

- qmin

- (3,1)

- J. Decker, MIT

**145≤********≤155 GHz**
**-30****o****≤********L****≤-10****o**
_**midplane launch**_
**10 kA of current for 5 MW of injected power**

**r/a(q****min****) ≈ 0.8**
**r/a(3,1) ≈ 0.87-0.93**
**Does (3,1) require less current than (3,2)?**
**Local *************, *************, Re****m**** effects so close to plasma edge?**
_**170 GHz may be adequate, but 200 GHz is better fit for FIRE parameters**_

## Neutral Beam Injection (Difficult)

**R****tan**** = 0 m**

**R****tan**** = 0.75 m**

**R****tan**** = 1.7 m**

_**R**__**tan**__** = 0 m**_
**16 TF Coils**
**Need 1 MeV to get 50% of power inside a/2**
_**R**__**tan**__** = 0.75 m and higher**_
**Must go to 12 TF coils****, pinwheel ports, and Fe inserts**
**Need > 1 MeV to get 50% of power inside a/2**
_**Plasma rotation for R**__**tan**__** = 1.7 m**_
**Assumed 120 keV & 8 MW ----> deposited r/a > 0.65**
**Dominated by j********B rotation giving v/v****Alfven**** ≤ 0.5%**

## Power Handling

_**First wall**_
**Surface heat flux**
**Plasma radiation, Q****max**** = P********+ P****aux**
**Volumetric heating**
**Nuclear heating, q****max**** = q****peak****(Z=0)**
_**VV, Cladding, Tiles, Magnets....**_
**Volumetric heating**
**Nuclear heating, q****max**** = q****peak****(Z=0)**
_**Divertor**_
**Surface heat flux**
**Particle heat flux, Q****max**** = P****SOL****/A****div****(part)**
**Radiation heat flux, Q****max**** = P****SOL****/A****div****(rad)**
**Volumetric heating**
**Nuclear heating, q****max**** = q****peak****(divertor)**

**plasma**

- VV

- Clad

- Tile

## Power Handling

_**Pulse length limitations**_
**VV nuclear heating (stress limit), 4875 MW-s -----> P****fus**** (q****VV****nuclear****)**
**FW Be coating temperature, 600****o****C -----> Q****FW**** & P****fus**** (q****Be****nuclear****)**
**TF coil heating, 373****o****K -----> B****T**** & P****fus**** (q****Cu****nuclear****)**
**PF Coil heating-AT-mode, 373****o****K -----> Ip, li, ********p, and ********  ****(not limiting)**** **
_**Component limitations**_
**Particle power to outboard divertor < 28 MW**
**Radiated power on (inner&outer) divertor/baffle < 6-8 MW/m****2**

## Power Handling/Operating Space

_**FIRE H-mode Operating Space**_
******N**** limited by NTM or ideal MHD with NTM suppression**
**----->  maximum P****fus**
**Higher radiated power in the divertor allows more operating space, mainly at higher ********N**
**----->  maximum P****fus**
**Majority of operating space limited by TF coil flattop**
**-----> ********flattop**** ≤ 20 s**
**High Q (≈15-30) operation obtained with **
**Low impurity content (1-2% Be)**
**Highest H****98**** (1.03-1.1)**
**Highest n/n****Gr**** (0.7-1.0)**
**Highest n(0)/********n******** (1.25)**

**H****98(y,2) ****≤ 1.1**

## Power Handling/Operating Space

_**FIRE AT-mode Operating Space**_
******N**** is limited by ideal MHD w/wo RWM feedback  ----->  maximum P****fus**
**Higher radiated power in the divertor allows more operating space, mainly at higher ********N**
**----->  maximum P****fus**
**Majority of operating space limited by VV nuclear heating**
**-----> ********flattop**** ≤ 20-50 s**
**Design solutions to improve VV nuclear heating limit, could reach PF coil limit, function of Ip**
**Number of current diffusion times accessible is reduced as ********N****, B****T****, Q increase**

**H****98(y,2) ****≤ 2.0**

## Particle Fueling/Pumping

|  | Gas Fueling System | Pellet Fueling System | Remarks |
| --- | --- | --- | --- |
| Design fueling rate | 200 torr-l/s for 20 s | 200 torr-l/s for 20 s | Pumping cap. 200 torr-l/s |
| Operational fuel rate | 100-175 torr-l/s | 100-125 torr-l/s | Isotopic fueling |
| Fuel isotope | D(95-99%), T,H(5-1%) | T(40-99%), D(60-1%) | D-rich edge, T-rich core |
| Impurity fuel rate | 25 torr-l/s | Prefer gas |  |
| Impurity species | He3, Ne, Ar, others |  |  |
| Rapid shutdown | Massive gas puff | Killer pellet or liquid jet | Disruption/VDE |
| Pellet sizes |  | 3-4 mm |  |

**Require ≈ 1-2********10****21****  tritons/s for FIRE H-mode**
**---> 0.1-0.2 g T injected per shot (20 s)**
**---> 5% of injected tritium consumed**
**HFS launch, limited to 125 m/s (test actually performed at ORNL to find pellet speed limit)**
**LFS and VL can reach much higher velocities**
**VL is at major radius, therefore not expected to provide improvement over LFS **

## Particle Fueling/Pumping

**16 cryocondensation/diffusion pumps, 8 above and 8 below midplane, every other port**
**Backed by turbo/drag pumps**
**H****2****O pumped on 1 m long 30****o****K entrance duct**
**H and impurities pumped by cryocondensation, liquid He**
**He pumped by turbo/drag pump located outside bio-shield, viscous drag compression (200 l/s conductance)**
**Cooling requirement for 16 cryopumps at 200 torr-l/s and nuclear heating (0.03 W/cm****3****) is 48 W, and liquid He flow rate is 64  l/hour for all 16 pumps**
**Regeneration is done into the turbo/drag pumps**

**Pumpdown and vessel bakeout utilize midplane pump, to provide minimum of 2000 l/s to reach 10****-7**** or less base pressure**

## Particle Fueling/Pumping

- V = 125 m/s
- Parks, 2003

**WHIST simulation of FIRE H-mode discharge (Houlberg)**
**Assume uniform pellet deposition**
**Obtains some density peaking with sufficient pumping**

## MHD Stability

_**Sawtooth **__**H-mode**_
**Unstable to internal kink, r/a(q=1) ≈ 0.35 m**
**-----> coupling to other global modes?**
**Porcelli sawtooth model (********W****MHD****+ ********W****KO****+ ********W****fast****),  incorporated into TSC indicates effect on fusion performance is weak**
**Pedestal/bootstrap broadens j profile**
**Rapid reheat of sawtooth volume**
******-particles providing stabilization**
**Complete stabilization would require RFCD since FIRE does not have high energy minority species**
**The q=1 surface can be removed from the plasma by **
**1.2 MA off-axis CD**
**Reduction of Ip to 6.0 MA**

## MHD Stability

_**Neoclassical Tearing Modes **__**H-mode**_
**Stable or unstable?**
**Sawteeth and ELM’s are expected to be present and can drive NTM’s**
**Typical operating point is at low ********N**** and ********P**
**Can lower ********N**** further if near threshold**
**Lower Hybrid CD at the rational surfaces**
**Compass-D demonstrated LH stabilization**
**Analysis by Pletzer and Perkins showed stabilization was feasible (PEST3)**
**Lowers Q(=P****fus****/P****aux****)**
**EC methods require high frequencies at FIRE field and densities  ---->  280 GHz**

**TSC-LSC**

**(3,2) surface**
**12.5 MW**
**0.65 MA**
**n/n****Gr**** = 0.4**
**Q = 6.8**

## MHD Stability

**FIR-NTMs**

**ususal NTMs**

**S. G****ünter et al., PRL 2001**

**ASDEX-U**

_**Current profile modification**_

## FIRE MHD Stability

- Despite (3,2) NTM excellent confinement: H98y=1.4, N = 3.3

- (LHCD ctr-CD in start-up phase)

**JET**

_**Current profile modification**_

## MHD Stability

_**Ideal MHD Stability **__**H-mode**_
_**n=1 external kink and n=∞ ballooning modes **__**H-mode**_
_**Stable without a wall/feedback**_
**Under various profile conditions ********N**** ≤ 3**
**ballooning unstable in pedestal region depending on pedestal width and magnitude**
_**Intermediate n peeling/ballooning modes **__**H-mode**_
_**Unstable**_**, primary candidate for ****ELM’s**
**Type I ELM’s are divertor lifetime limiting, must access Type II, III**
**P****loss****/P****LH**** ≈ 1.0-1.6 in flattop, not > 2 like many present experiments**
**FIRE has high triangularity (********x**** = 0.7) in Double Null and high density **
**Active methods to reduce ********W****ELM**** include pellets, impurities, ergodization,...**

**Self-consistent ohmic/bootstrap equilibria**

## MHD Stability

_**Neoclassical Tearing Modes **__**AT-mode**_
**Unstable or Stable?**
**q(********) > 2 everywhere, so rational surfaces are (3,1), (5,2), (7,3), (7,2)...**
- r/a(qmin) ≈ 0.8
- r/a(3,1) ≈ 0.87-0.93
- Local *, *, Rem effects so close to plasma edge? L-mode or H-mode conditions
**Examining EC stabilization at 170 GHz**
**LFS absorption, Ohkawa CD dominates**
**Scaling from (3,2) expts indicates high power ----> ****early detection required**
**LH using two spectra, one for bulk CD and other for NTM suppression**

## MHD Stability

_**Ideal MHD Stability **__**AT-mode**_
- n= 1, 2, and 3...external kink and n = ∞ ballooning modes
_**n = 1 stable without a wall/feedback for **__****__**N**__** < 2.5-2.8**_
_**n = 2 and 3 have higher limits without a wall/feedback**_
_**Ballooning stable up to **__****__**N **__**< 6.0,**_ unstable in pedestal region of H-mode edge plasmas.
- RWM stabilization with feedback coils, **VALEN analysis** indicates 80-90% of ideal **with wall** limit for n=1
_**n = 1 stable with wall/feedback to **__****__**N**__**’s around 5.0-6.0**_
_**n = 2 and 3 appear to have lower **__****__**N**__** limits in presence of wall**_, possibly blocking access to n = 1 limits
- Intermediate n peeling/ballooning modes
_**Unstable**_ under H-mode edge conditions

- Growth Rate, /s

- N

- N=4.2

**Bialek, Columbia Univ**.

## MHD Stability

**JSOLVER Equilibria**

**TSC-LSC Simulation Equilibria**

## MHD Stability

_**Other MHD Issues H-mode and AT-mode**_
**Alfven eigenmodes**** and energetic particle modes**
**Snowmass assessment indicated stable for H-mode, although access to shorter pulse high P****fus**** plasmas should destabilize **
**AT-mode not analyzed**
**Error fields**** from coil misalignments, etc.  ---->  install Cu window coils outside TF coil, stationary to slow response**
**FIRE does not have an external source of rotation**
**Transport, sheared rotation**
**Resistive instabilities, sheared to bulk rotation**
**RWM, bulk rotation**
**Plasma self-rotation (C-Mod), is it sufficient for some stabilization**

## Disruption Modeling

|  |  | Comments |
| --- | --- | --- |
| frequency | 10-30% | 30% plasma development, 10% for repetitive operation |
| number | 300 (900) | 300 at maximum Wmag and Wth |
| thermal energy | 35 MJ |  |
| themal quench duration | 0.2 ms (0.1-0.5) | single or multi-step quench |
| fraction of Wth to divertor | 80-100% (or ver low for mitigation)  | conduction to targets, 2-1 asymmetry |
| fraction of Wth to first wall/baffle | < 30% (or nearly 100% for mitigation) | radiation |
| inside to outside divertor split | 5 to 1 | need data on DN |
| poloidal localization | 3x typical SOL width(1x to 10x) | this may not be reliable |
| magnetic energy | 25 MJ |  |
| current quench | 6 ms (2-600 ms) |  |
| Ip decay rate | 1 MA/ms typical, 3 MA/ms maximum |  |
| fraction of Wmag to FW by radiation | 80-100% | peaking factor of 2 |
| fraction of Wmag to FW  by conduction | 0-20% |  |
| VDE frequency | 10% of disruptions | uncertain due to up-down symmetry |
| halo current fraction, Ihalo/Ip | 0.4 (0-0.5) |  |
| halo current toroidal peaking | 2 (1.2-4) |  |
| (Ihalo/Ip)  TPF | 0.5 typical maximum | experimental data boundary is 0.75 |
| runaway electron current | 50% of Ip | Highly uncertain, high density in FIRE makes this much weaker |
| localization of runaway deposition | < 1 m2 | PFC and wall alignment |

## Disruption Modeling

_**Experimental database used to project for FIRE**_
**Thermal quench time**
**I****halo****/Ip ******** TPF**
**dIp/dt rates for current quench**

## Disruption Modeling

_**TSC simulation of disruption**_
**Critical structures modeled; VV’s (SS), passive plates (Cu), cladding (Cu), divertor (Cu), baffle (Cu), midplane port regions (SS)**
**Zero-net current constraint on divertor, baffle, midplane port regions**
**Provide poloidal current paths for halo/structure currents**

## Disruption Modeling

**Thermal quench, ********t = 0.2 ms**

**Ip drop, -2.9 MA/ms**

**Diamagnetic flux**

**Rapidly drop pressure over 0.2 ms**
**Use hyper-resistivity to broaden current**
**Plasma temperature drops to 15-30 eV, current is shared with halo region depending on T****halo**** (2-7 eV) and halo width**
**Ip drops at rate determined largely by T****halo**
**Plasma shrinks rapidly, then plasma is converted to a circuit**

## Disruption Modeling

_**TSC simulation produces for Engr. analysis**_
**Toroidal structure currents, fields and forces**
**Poloidal structure current, fields, and forces**
**Plasma toroidal currents on a grid**
**Halo/poloidal plasma currents at structure interfaces**
**Global plasma and PF coil data**

## Disruption Mitigation

_**Utilize fueling technology to mitigate electromagnetic effects of disruptions**_
**Massive gas puff into DIII-D  ---->  peak halo currents reduced by 50% by He and D puffing, and toroidal asymmetry reduced**
**Ne, Ar, and CH****4**** pellets into DIII-D  ----> peak halo currents reduced by 50% with Ne and Ar pellets, and toroidal asymmetry reduced from 3 to 1.1**
**Cryogenic liquid jet being developed**
**Low Z impurity pellets (LiD) if runaway electrons not an issue**
**Snowmass assessment indicated large radiated power to FW could cause Be melting **

**DIII-D Massive**
**Gas Puff System**

## FIRE Transport and Confinement

_**Energy Confinement Database**_
******E****98(y,2)**** = 0.144 M****0.19**** Ip****0.93  ****B****T****0.15  ****R****1.97  ********0.58  ****n****20****0.41  ********0.78  ****P****-0.69  ****(m, MA, T, MW)**
******p*********/********E**** = 5**
**Z****eff**** = 1.2-2.2 (f****Be**** = 1-3%, f****Ar**** = 0-0.3%)**
_**Pedestal Database (Sugihara, 2003)**_
**P****ped****(Pa) = 1.824********10****4****M****1/3****Ip****2****R****-2.1****a****-0.57********3.81****(1+********2****)****-7/3****(1+********)****3.41****n****ped****-1/3****(P****tot****/P****LH****)****0.144**
**----> T****ped**** = 5.24 ± 1.3 keV**
**---->  ********ped****??**
_**L-H Transition**_
**P****LH****(MW) = 2.84M****eff****-1****B****T****0.82****n****L20****0.58****Ra****0.81 **** (2000) ----> 26 MW in flattop**
**P****LH****(MW) = 2.58M****eff****-1****B****T****0.60****n****L20****0.70****R****0.83**** a****1.04 **** (2002) ----> 18.5-25 MW in flattop**
**DN has less or equal P****LH**** compared to favored SN (Carlstrom, DIII-D; NSTX; MAST)**
_**H-L Transition & ELM’s**_
**P****loss**** > P****LH**** although hysterisis exists in data**
**Type I ELM’s typically require P****loss**** > 1.( )********P****LH****, expts typically > 2********P****LH**
**Type II ELM’s require strong shaping, higher density, DN ---> reduced P****div****, H****98****=1**
**Type III ELM’s, near P****loss**** ≈ P****LH****, or high density, reduced H****98**
**Active methods ----> pellets, gas puffing, impurity seeding, ergodization**

## Pedestal Physics and ELM’s

_**Pedestal physics**_
**Intermediate n peeling/ballooning modes**
**----> ballooning ****destabilized**** by high p’ and low j**
**----> peeling modes**** destabilized**** by high j and low p’**
**Stronger shaping raises p****ped**
**Stability analysis distinguishes n****ped**** and T****ped**** through *************ped**** (n****ped****/T****ped****2****) ---> j****BS**
**Higher n****ped**** leads to mode envelope narrows and lowers j****BS****  ---> smaller ********W****ELM**

**weak shaping**

**strong shaping**

**ELITE projections for FIRE**

## Pedestal Physics and ELM’s

_**Type I ELM trends**_
**Reduced ********W****ELM****/W****ped**** with increasing *************ped****  ----> ****inconsistent with higher T****ped**** for high Q**
**Reduced ********W****ELM****/W****ped**** with increasing ********||****i**** ----> ****inconsistent with higher T****ped**** for high Q**
******W****ELM****/W****ped**** correlated with ********T****ped****/T****ped**** as n****ped**** varied, very little change in ********N****ped****/N****ped**

_**Type II ELM’s**_
**ASDEX-U with ****DN ****and high n ----> H****98**** = 1-1.2 and reduction in divertor heat flux by 3******
**JET with high ******** and high n ----> mixed Type I+II, no reduction in confinement and 3******** reduction in ELM power loss**

**P****in**

**W****th**

**P****rad**

**P****ELM**

**JET**

## Pedestal Physics and ELM’s

_**Active methods for ELM mitigation**_
**JET argon seeding in Type I, f****rad**** > 0.65, H****98**** ≈ 1, n/n****Gr****> 0.7, Q**** div**** reduced by 2******
**Type III, f****rad**** > 0.7, H****98**** ≈ 0.7-0.9, n/n****Gr**** > 0.7**
**Pellets that trigger ELMs, avoiding large infrequent Type I ELMs**
**Ergodization of plasma edge region, use coils to produce high (m,n) field that perturb only ELM region**

- JET

## POPCON Operating Space vs. Parameters

**T(0)/********T********, n(0)/********n********, ********p*********/********E****, H****98****, f****Be****, f****Ar**

**H****98(y,2)**** must be ≥ 1.1 for robust operating space**

## 1.5D Integrated Simulations

**Tokamak Simulation Code (TSC)**
**Free-boundary**
**Energy and current transport**
**Density profiles assumed**
**GLF23 & MMM core energy transport**
**Assumed pedestal height/location**
**ICRF heating, data from SPRUCE**
**Bootstrap current, Sauter single ion**
**Porcelli sawtooth model**
**Coronal equilibrium radiation**
**Impurities with electron density profile**
**PF coils and conducting structures**
**Feedback systems on position, shape, current**
**Use stored energy control**
**Snowmass E2 simulations for FIRE**
**Corsica, GTWHIST, Baldur, XPTOR**

## 1.5D Integrated Simulations

**FIRE H-mode, GLF23**

## 1.5D Integrated Simulations

| FIRE | Q | Paux(MW) | Tped(keV) |
| --- | --- | --- | --- |
| TSC
GLF | 10.3 | 13.5 | 4.5 |
|  | 10.0 | 7.5 | 3.8 |
|  | 10.0 | 10.0 | 4.1 |
|  | 10.0 | 12.5 | 4.4 |
|  | 10.0 | 15.0 | 4.7 |
|  | 10.0 | 20.0 | 5.4 |
|  |  |  |  |
| Baldur
MMM | 4.5 | 30.0 | 2.5* |
|  | 7.0 | 10.0 | 2.5* |
|  |  |  |  |
| XPTOR/12
GLF | 5.0 | 20.0 | 3.0 |
|  | 10.0 | 20.0 | 4.0 |
|  | 15.0 | 20.0 | 5.0 |
|  |  |  |  |
| Corsica
GLF | 4.0 | 12.5       | 2.5 |
|  | 6.0 | 12.5 | 4.0 |
|  | 10.0 | 12.5 | 5.0 |

## 0D Advanced Tokamak Operating Space

**Scan** ----> q95, n(0)/n, T(0)/T, n/nGr, N, fBe, fAr
**Constrain** ----> LH = 0.16, FW = 0.2, PLH ≤ 30 MW, P ≤ 30 MW, IFW = 0.2 MA,
- ILH = (1-fbs)Ip, Q
**Screen**  ----> flattop(VV, TF, FW heating), Prad(div), Ppart(div), Paux< Pmax

## Observations from 0D Analysis for Burning Plasma AT

**In order to provide reasonable fusion gain Q≥5, **_**can’t operate at low density**_** to maximize CD efficiency**
_**Density profile peaking is beneficial**_** (pellets or ITB), since broad densities increase required H****98**** and P****CD**
**Access to **_**high density**_** relative to Greenwald density, in combination with **_**high bootstrap**_** current fraction gives the lowest required H****98**
_**H**__**98 **__**≥1.4 are required to access **__****__**flattop**__**/**__****__**curr diff **__**> 3**_**, however, the ELMy H-mode scaling law is known to have a ******** degradation that is not observed on individual experiments**
_**Radiative core/divertor solutions**_** are a critical area for the viability of burning AT experiments due to high P********+P****CD****, suggesting impurity control techniques**
**Access to higher radiated power fractions in divertor enlarge operating space significantly**
**Access to higher ********flattop****/********j**** decreases at higher ********N****, higher B****T****, and higher Q, since ********flattop**** set by VV nuclear heating**

## Examples of FIRE Q=5 AT Operating 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)**

## 1.5D Integrated Simulations AT-mode

**Ip=4.5 MA  **
**Bt=6.5 T  **
******N****=4.1  **
**t(flat)/********j=3.2  **
**I(LH)=0.80**
**P(LH)=25 MW **

**f****BS****=0.77  **
**Z****eff****=2.3 **
**q(0) =4.0 **
**q(min) = 2.75 **
**q(95) = 4.0 **
**li = 0.42, **
****** ****= 4.7%, ********P**** = 2.35**

## 1.5D Integrated Scenarios AT-mode

**t = 12-41 s**

## 1.5D Integrated Scenarios AT-mode

**n/nGr = 0.85 **
**n(0)/<n> = 1.4**
**n(0) = 4.4x10^20**
**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**

## Perturbation of AT-mode Current Profile

**5 MW perturbation to P****LH**
**Flattop time is sufficient to examine CD control**

**t = 12 s**
**t = 25 s**

**t = 25 s**
**t = 41 s**

## Conclusions

- The FIRE device design provides **sufficient/flexible/relevant** operating space to examine burning plasma physics
- Sufficient to provide burning conditions (**Q ≥ 10** inductive and **Q ≥ 5** AT, does not preclude **ignition**)
- Flexible to accommodate **uncertainty and explore** various physics regimes
- Relevant to **power plant plasma physics and engineering** design
- The **subsystems on FIRE**, within their operating limits, are suitable to **examine burning plasma physics**  ---->  subject to R&D in some cases
- Auxiliary heating/CD
- Particle fueling and pumping
- Divertor/baffle and FW PFC’s
- Magnets
- Diagnostics

## Conclusions

**Burning plasma conditions** can be accessed and studied in both **standard H-mode** and **Advanced Tokamak** modes.  The range of AT performance has been expanded significantly since Snowmass
- FIRE can reach **1-5 ********j****, and examine current profile control**
- Design improvement to FW tiles could **extend flattop times further**
- FIRE can reach **80-90% of ideal with wall limit**, with RWM feedback
- FIRE can reach high **I****BS****/I****P**** (77% in 1.5D simulation)**
- Identified that **radiative mantle/divertor** solutions significantly expand operating space
- FIRE will pursue Fe shims for AT operation
- The **physics basis** for FIRE’s operation is based on **current experimental and theoretical results**, and projections based on these continue to provide **confidence** that **FIRE will achieve the required burning plasma performance**

## Issues/Further Work

- Magnets
- Ripple reduction, design Fe shims for AT mode
- Continue equilibrium analysis
- Complete plasma breakdown and early startup
- Complete internal control coil analysis
- RWM coil design/integration into port plugs, time dependent analysis
- Error field control coil design
- Heating and CD
- Continue ICRF antenna design, disruption loads, neutron/surface heating
- Engineering of 4 strap expanded antenna option
- More detailed design of LH launcher, disruption loads, neutron/surface heating
- Complete 2D FP/expanded LH calculations for FIRE specific cases
- Continue examination of EC/OKCD for NTM suppression in AT mode
- Pursue dynamic simulations/PEST3 analysis of LH NTM stabilization for both H-mode and AT-mode

## Issues/Further Work

- Power Handling
- Pulse length limitations from VV nuclear heating, design improvements
- FW tile design, material choices, impacts on magnetics
- Continue divertor analysis, UEDGE and neutrals analysis for integrated heat load, pumping,and core He concentration solutions
- Continue examination of ITPA ELM results and projections, encourage DN strong triangularity experiments
- DN up-down imbalance, implications for divertor design (lots of work on DII-D)
- Disruption mitigation strategies, experiments
- Particle Handling
- Continue pellet and gas fueling analysis in high density regime of FIRE
- Neutrals analysis for pumping
- Be behavior as FW material and intrinsic impurity
- Impurity injection, core behavior, and controllability
- Particle control techniques: puff and pump, density feedback control, auxiliary heating to pump out core, etc.
- Wall behavior, no inner divertor pumping, what are impacts?

## Issues/Further Work

- MHD Stability
- LH stabilization of NTM’s, analysis and experiments (JET, JT-60U and C-Mod)
- Examine plasmas that appear not to be affected by NTM’s (current profile)
- Early (before they are saturated) stabilization of NTM’s with EC/OKCD
- Continue to develop RWM feedback scheme in absense of rotation
- Identify impact of n=2,3 modes on wall/feedback stabilized plasmas
- Examine impact of no external rotation source on transport, resistive and ideal modes
- Alfven eigenmodes/energetic particle modes, onset and accessibility in FIRE
- Plasma Transport and Confinement
- Continue core turbulence development for H-mode, ITPA
- Establish AT mode transport features, ITB onset, ITPA
- Pedestal physics and projections, and ELM regimes, ITPA
- Impact of DN and strong shaping on operating regimes, Type II ELMs
- Improvements to global energy confinement scaling, single device trends
- Expand integrated modeling of burning plasmas