# Presentation: 024294

##                                            Supported by   

**Macroscopic Stability Research for 2009-13**** **

**Steven A. Sabbagh**
*Columbia University*
*For the NSTX Research Team*

**NSTX 5 Year Plan Review for 2009-13**
**Princeton Plasma Physics Laboratory**
**July 28-30, 2008**

***College W&M***
***Colorado Sch Mines***
***Columbia U***
***Comp-X***
***General Atomics***
***INEL***
***Johns Hopkins U***
***LANL***
***LLNL***
***Lodestar***
***MIT***
***Nova Photonics***
***New York U***
***Old Dominion U***
***ORNL***
***PPPL***
***PSI***
***Princeton U***
***SNL***
***Think Tank, Inc.***
***UC Davis***
***UC Irvine***
***UCLA***
***UCSD***
***U Colorado***
***U Maryland***
***U Rochester***
***U Washington***
***U Wisconsin***

***Culham Sci Ctr***
***U St. Andrews***
***York U***
***Chubu U***
***Fukui U***
***Hiroshima U***
***Hyogo U***
***Kyoto U***
***Kyushu U***
***Kyushu Tokai U***
***NIFS***
***Niigata U***
***U Tokyo***
***JAEA***
***Hebrew U***
***Ioffe Inst***
***RRC Kurchatov Inst***
***TRINITI***
***KBSI***
***KAIST***
***POSTECH***
***ASIPP***
***ENEA, Frascati***
***CEA, Cadarache***
***IPP, J******ü******lich***
***IPP, Garching***
***ASCR, Czech Rep***
***U Quebec***

**                                           ****Supported by   **

- v1.9

## Understanding what profiles and control systems are needed for burning plasmas must occur before such devices 

**Understanding what profiles and control systems are needed for burning plasmas must occur before such devices **

- FESAC US ST mission:
*Develop compact, high ****, burning plasma capability for fusion energy*
- Stability Goal (in one sentence)
- Demonstrate reliable maintenance of high ***N* with sufficient physics understanding to extrapolate to next-step devices
- Knowledge base needed to bridge to these devices
- _Demonstration = Control_ (of modes and plasma profiles):
- Need to determine what control is needed *before* CTF (for greatest simplicity)
- _Understanding = Vary parameters (+operate closer to burning plasma levels)_:
- _Collisionality_: influences *V*** damping
- _Shaping_:
- _Plasma rotation level, profile_:
_*q*__ level, profile_:

_*CTF*_*:** ****N** = 3.8 – 5.9  (W**L** = 1-2 MW/m**2**)**	*_*ST-DEMO*_*:** ****N** ~ 7.5 *
* *Both at, or above ideal no-wall -limit; deleterious effects at ~ 1⁄2 ***N**no-wall*
-  high ***N* accelerates neutron fluence goal - takes 20 years at *W**L* = 1 MW/m2)

***All influence ************-limiting modes:***
***Kink/ballooning, RWM, NTM***

- }

## Advanced tokamak operation demonstrated in a mega-Ampere class spherical torus 

- High  operational space
- Ultra-high ***t* = 39%, near unity in core
- Broad current and pressure profiles
-  ***N* > 7, ***N**/l**i* > 11
- Wall-stabilized, ***N**/****N**no-wall* > 50% at highest ***N*
- Future research moves to develop predictive capability and control for *steady-state operation near with-wall **** limit*
- Extrapolate to next step device with high confidence

- S.A. Sabbagh, et al., *Nucl. Fusion * **46**, 635 (2006).** **

## Development of device hardware empowers fundamental stability understanding for robust extrapolation to next-step STs

- Operate at parameters closer to burning plasma (e.g. low i)
- High plasma shaping ( ~ 3), low *l**i* operation
- Vertical stability, kink/ballooning stability, coupling to passive stabilizers
- Resistive wall mode (RWM) stabilization
- Understand physics of passive mode stabilization vs. *V*** at reduced ***i*
- Non-axisymmetric field-induced viscosity
- Non-resonant and resonant, due to 3-D fields and modes at reduced ***i*
- Control modes and profiles, understand key physics
- Dynamic error field correction (DEFC)
- Demonstrate sustained *V*** with reduced resonant field amplification, under *V*** profile control
- Resistive wall mode control
- Increase reliability of active control, investigate multi-mode RWM physics under *V***, *q* control
- Tearing mode / NTM
- Stabilization physics at low A, mode locking physics under *V***, *q* control
- Plasma rotation control
- Sources (2nd NBI, magnetic spin-up) and sink (non-resonant magnetic braking)
- Mode-induced disruption physics and prediction/avoidance

-  Center stack	upgrade
- - Liquid Li divertor

- - 2nd NBI (incr.)
-  Singly-powered 	RWM control coils
-  coil upgrade (incr.)

## Plasma equilibrium goal to access and maintain stable high N at high shaping

**Plasma equilibrium goal to access and maintain stable high ***********N***** at high shaping**

- Progress
- Central coil PF1A modified (2005) to allow high shaping
- Sustained ** < 2.7, ** < 0.8; transient ** = 3 with record shaping factor, *S**I*  *q**95**(I**p**/aB**t**)* = 41
- Note: Present CTF design has ** = 3.07, lower *S**I*
- Highest ** and *S**I* plasmas reached ***N* ~ 6 in 2008
- Plan summary 2009-2011
- Assess/utilize ** feedback control using real-time EFIT and NBI power to avoid fast kink/ballooning disruptions
- Conduct experiments/analysis to extend high *S**I* plasmas into wall-stabilized, high ***N* > 6 operating space
- Plan summary 2012-2013
- Real-time MSE for evaluation of *q* in real-time EFIT
- Utilize/analyze ** feedback using stability models; *q *profile control with 2nd NBI (incremental)
- Study ST-CTF target shapes (increased A) at low ***i* with favorable profiles, determine sensitivity to variations in *l**i*, **

- D.A. Gates, et al., *Nucl. Fusion* **47**, 1376 (2007).** **

***121241***
***t=275ms***

## Dynamic Error Field Correction used to increase pulse length

- _Feedback System Trained for n=1 DEFC_

- _Important to Correct n > 1 Error Fields_

- Apply preprogrammed n = 1 fields
- Adjust feedback gain, phase, so that feedback cancels those currents
- then remove n=1 EF source to correct intrinsic error fields

- Pre-programmed n = 3 fields, two phases
- Asymmetric response in rotation, pulse length
- n = 3 intrinsic error field present (PF5, TF most likely causes)
- n = 2 error fields found to be less important

## Goal to make resonant and non-resonant DEFC effective over large range of plasma conditions

**Goal to make resonant and non-resonant DEFC effective over large range of plasma conditions**

- FY2009-10
- Optimize DEFC control with expanded sensor arrays
- Plasma response using IPEC
- Test real-time n=3 correction (IPF5)
- Develop tensor-pressure equilibria with IPEC
- FY2011-2012
- Modify IPEC to allow magnetic islands. Compare to experiments involving interaction of 3-D applied fields and NTMs.

- FY2012-2013
- Study EF thresholds at BT < 1T, assess intrinsic error fields at higher BT, Ip
- Utilize 2nd SPA power supply for greater DEFC spectrum flexibility
- Utilizing upgraded RWM coil geometry and IPEC modeling, optimize DEFC poloidal mode spectrum for best *V***. Implement appropriate control system upgrades as required.

- _Ideal Perturbed Equilibrium Code (IPEC_)
- J.-K. Park, et al., Phys. Plasmas **14**, 052110 (2007).

## Active RWM control and error field correction maintain high N plasma

**Active RWM control and error field correction maintain high ********N**** plasma**

- NSTX _record pulse lengths_
- Up to 1.8s (shown previously)
- n = 1 active control
- Upper/lower Bp sensors
- Favorable Br feedback settings found in 2008
- Fast response ~ 1 ms
- n = 3 DC field correction
- best maintains ****
- but - RWMs observed w/o feedback at high ****
- n = 1 feedback now being used as tool in many XPs
- > 200 shots in 2008 with active feedback in 10 XPs
- Present goal to increase reliability, performance
- Feedback success ~ 74%
- RWM more likely when NTM stabilized (e.g. by lithium)
- Poloidal deformation of mode
- Considering system upgrades

## VALEN code reproduces experimental RWM feedback performance

**VALEN code reproduces experimental RWM feedback performance**

- Model simulates experiments
- Actual RWM sensors locations
- Compensation of control field
- Experimental equilibrium reconstruction (including MSE data)
- Proportional gain
- Low *V*** results shown
- Advanced control may greatly improve performance
- Advanced state-space controller may stabilize N/Nwall < 95% (benefit to ITER, next-step ST)
- Plan to test offline 2009-10; implement for initial RWM control testing 2010-11

- Growth rate (1/s)

- N

- DCON
- no-wall
- limit

- Experimental N (control off)
- (collapse)

- With-wall limit

- active control

- passive
- growth

- active
- feedback
- (present)

- Experimental N (control on)

- Advanced
- state-space controller

- reached

## Design work for upgraded non-axisymmetric control capabilities has begun 

**Design work for upgraded non-axisymmetric control capabilities has begun**** **

- Capabilities
- Non-axisymmetric control coil (NCC) – at least _four_ applications
- RWM stabilization       (*n* > 1, higher ***N*)
- DEFC with greater field correction capability
- ELM control (*n* = 6)
*n* > 1 propagation, increased Vcontrol)
- Similar to proposed ITER coil design
- In incremental budget
- Addition of 2nd SPA power supply unit for simultaneous n > 1 fields
- Non-magnetic RWM sensors; advanced RWM active feedback control algorithms
- Alteration of stabilizing plate connections

- Primary
- PP option

- Secondary PP option

- Existing
- coils

- _Proposed Internal _
- _Non-axisymmetric Control Coil (NCC)_
- _(initial designs - 12 coils toroidally)_

- _RWM with __*n*__ > 1 RWM_
- _observed_

- (Sabbagh, et al., Nucl. Fusion **46**, 635 (2006). )

## Significant progress planned for RWM control research(i) Active feedback control

- Plan summary 2009-2011
- Continue use of RWM feedback control as standard tool – build performance statistics
- Investigate underlying active control physics
- Continue feedback control parameter optimizations/analysis, using all RWM sensors
- Determine effectiveness over range of Vprofiles and levels, ***i* level
- Test advanced state-space active stabilization algorithms offline; implement and perform initial tests for RWM control
- Determine the role of plasma response during RWM control using IPEC
- Investigate multiple modes in stabilization, implement methods of decreasing possibility of RWM poloidal deformation
- Multi-mode VALEN code completed (running NSTX, DIII-D, HBT-EP tests now)
- Design high-*n* control coil (NCC); determine need for passive plate modifications
- Plan summary 2012-2013
- Assess RWM control at reduced ***i* with center stack upgrade; during V profile control
- Analyze initial non-magnetic (SXR) RWM sensor data
- compare to magnetic sensors; evaluate future use in RWM feedback
- Implement 2nd switching power amplifier (SPA) to allow independent control of all 6 midplane RWM coils
- Implement NCC and begin to use for RWM stabilization; *n* > 1 RWM, multi-mode study during *n* = 1 stabilization; compare to theory (incremental)
- Examine greater range of V and *q* profiles with 2nd NBI, greater non-resonant braking flexibility using 2nd SPA; NCC (incremental)

## Non-resonant magnetic braking allows V modification to probe RWM “critical rotation” and stabilization physics

**Non-resonant magnetic braking allows V******** modification to probe RWM “critical rotation” and stabilization physics**

- Slowest rotation profiles produced in NSTX are at DIII-D balanced-NBI levels
- Ion collisionality profile variation appears to alter experimental ***crit* profile

- Scalar plasma rotation at q = 2 inadequate to describe stability
- Marginal stability N > Nno-wall, q=2 = 0

* ****crit* doesn’t follow simple ***0*/2 rotation bifurcation relation

- A.C. Sontag, et al., NF **47** (2007) 1005.

- (***0*  steady-state plasma rotation)

- /2 (kHz)

- q

- q = 2

- 0.2

- 0.4

- 0.6

- 0.8

- 1.0

- 1.2

- 1.4

- 1.6

- R (m)

- 1

- 2

- 3

- 4

## Kinetic modification of ideal MHD stability theory investigated to explain experiment

- Ideal MHD RWM growth rate formulation
- Kinetic modifications
- Trapped and circulating ions
- Trapped electrons
- Alfven dissipation at rational surfaces
- Stability depends on
- Integrated _**__**__ profile_: resonances in ***W**K* (e.g. ion precession, diamag. drift)
- Particle _collisionality_

**Kinetic modification of ideal MHD stability theory investigated to explain experiment**

- Hu and Betti, Phys. Rev. Lett **93**, 105002  (2004).

- _Trapped ion component of __**__*W*_*K*_ (plasma integral)_

- Pitch angle integral

- Energy integral

- Mode geometry

- _collisionality_

_**__**__ __profile (enters through ExB frequency)_

## Kinetic modifications show RWM unstable region at relatively high V - in agreement with experiment

- Marginal stability crossed at relatively high V
- in experiment and theory
- far greater than thought needed for robust stability in terms of A at q = 2

- _Theoretical variation of V_

- Marginally stable
- experimental
- profile

- 121083

- _RWM stability vs. V__ (contours of ___*w*_)_

- unstable

- _V__ variation_

- -0.6

- -0.4

- -0.2

- 0.0

- -0.8

- High V

- Low V

- Re(WK)

- Im(WK)

- _V__ variation_

## Kinetic stability model shows increase in stability as ion collisionality decreases – under investigation

- Opposite to results of models using only viscous dissipation
- Determine best theory model by comparing to recent experiments with Li evaporation; further reduced ***i* with upgraded center stack

## Significant progress planned for RWM stabilization research(ii) Passive stabilization physics

- Plan summary 2009-2011
- Determine RWM stabilization requirements for broader range of *V***profiles and ***i* allowed by lithium evaporation, liquid lithium divertor
- Test stability theories against marginal *V*** profile database, new parameter scans (2009 milestone)
- Continue analysis using kinetic ***W* – Hu-Betti-Manickam code
- Compare to latest MARS-K implementation (full kinetic effects modeled - Y. Liu)
- Determine/implement stabilization improvements by altering passive plate electrical reconnections
- Plan summary 2012-2013
- Examine passive stabilization in low ***i* plasmas created at higher *T**i* from higher *B**t* and *I**p*
- Examine RWM stabilization during *V**** *profile feedback control
- Determine a reliable physics model to confidently predict RWM stabilization as a function of rotation and ion collisionality
- Test stabilization physics model for greater range of *V*** profiles with 2nd NBI, greater non-resonant braking flexibility using 2nd SPA unit; with new NCC (incremental)

## Recent experiments and analysis have studied NTM stabilization physics of high- NSTX plasmas

**Recent experiments and analysis have studied NTM stabilization physics of high-********** NSTX plasmas**

- _2/1 onset threshold vs. V____ shear_

- _2/1 Marginal island width for stabilization_

- Flow shear variation achieved by different NB injection and *n* = 3 braking.
- Similar trends observed in co-/counter mix experiments in DIII-D
- Trend likely due to dependence of ’ on local flow shear

- Local Flow Shear:

- Local Mode Drive:

- Local drive required for mode onset

- Polarization Threshold Implies

- Collaboration between R.J. La Haye (GA), S. P. Gerhardt (PPPL), R. Buttery (JET), and M. Marschek (AUG)

- Trend confirms physics of polarization current model
- Expand to additional machines (JET, AUG), and achieve more cases

-  localp

## Establish predictive physics understanding of NTMS

- 2009-2011: Compete Characterization of NTM Onset, Small Island Physics, Restabilization
- Characterize the role of *V*** and the ideal kink limit on NTM onset thresholds
- Characterize triggering events, including sawtooth triggered 3/2 modes and “triggerless” NTMs with qmin > 1
- Finish characterization of the marginal island width for 2/1 and 3/2 modes, including comparisons to conventional aspect ratio devices
- Understand details of how Li conditioning and DEFC assist in stabilizing 2/1 modes
- 2009-2011: Establish a program of relevant NTM modeling
- Implement PEST-III calculations of ’ for realistic NSTX equilibria, including the effects of nearby rational surfaces
- Utilize initial value codes like NIMROD for more sophisticated treatment of transport near the island or rotation shear effects on mode coupling and island eigenfunction.
- 2012-2013: Develop scenarios that mitigate/eliminate deleterious NTM activity
- Quantify the benefits of qmin > 2 operation, and the role of higher order (3/1, 5/2) modes in this case
- Utilize increased toroidal field (new center stack) to scale *****i* in single device
- Utilize 2nd beamline for current profile control, possibly allowing ’ stabilization of NTMs even with qmin < 2

***Collaborations are an essential element of research plan (GA, AUG, JET, U. of Tulsa,...)***

## Non-axisymmetric field-induced neoclassical toroidal viscosity (NTV) important for low collisionality ST-CTF, low rotation ITER plasmas

**Non-axisymmetric field-induced neoclassical toroidal viscosity (NTV) important for low collisionality ST-CTF, low rotation ITER plasmas**

- Significant interest in plasma viscosity by non-axisymmetric fields
- Physics understanding needed to minimize rotation damping from ELM mitigation fields, modes (ITER, etc.)
- NTV investigations on DIII-D, JET, C-MOD, MAST, etc. following quantitative agreement on NSTX
- Expand present studies on NSTX
- Examine larger field spectrum
- Improve inclusion of plasma response using IPEC
- Consider expansions of NTV theory
- Saturation due to Er at reduced ion collisionality, multiple trapping states, matching theory through collisionality regimes
- Examine NTV from magnetic islands
- Stronger dependence on ***B/B*
- Compare to kinetic modeling (e.g. using GTC-Neo upgrade (W. Wang))

- _Measured __*d*__(__*I*__**__*p*__)/__*dt*__ profile and theoretical_
- _NTV torque (__*n*__ = 3 field) in NSTX)_

- W. Zhu, et al., *Phys. Rev. Lett.* **96, **225002 (2006).** **

- _Dominant NTV Force for NSTX collisionality..._

- _...__expected to saturate at lower ____i_

- Can verify at order of magnitude lower i with center stack upgrade

## Clear non-resonant braking observed in recent n = 2 applied field experiments – broader braking profile at lower n

- n = 2 configuration has strong n = 4, but essentially no (resonant) n = 1 component
- Recent experiment shows stronger braking with Li evaporation

- _Rotation evolution during n = 2 braking_

- Braking
- region

- R(m)

- 127488

- _Rotation evolution during n = 3 braking_

- Braking region,
- Peak change
- (1.3 < R(m) < 1.35

- (kHz)

- R(m)

- 124010

-   _t(s)_
- 0.445
- 0.455
- 0.465
- 0.475

-   _t(s)_
- 0.525
- 0.535

- N ~ 4.2

## Develop understanding of field-induced plasma viscosity with application to V profile control

- Plan summary 2009-2011
- Continue testing viscosity theory from resonant /non-resonant fields
- Examine key dependences; alter ***i* using lithium evaporation, LLD
- joint experiments with other devices (MAST 2008)
- Improved plasma internal field response using IPEC; influence of magnetic islands
- Expand analysis to further test theory
- Saturation due to Er at reduced ***i*, multiple trapping states, etc.
- Time-evolved kinetic computations using GTC-Neo; examine saturation at low ***i*
- Determine range of radial placement of torque possible with NCC design
- Begin real-time *V*** control using CHERS measurements, present NBI as source and *n*=3 NTV as sink of plasma toroidal momentum
- Examine effect on modes; initial control of flow shear near for NTM studies
- Plan summary 2012-2013
- Utilize reduced ***i* plasmas using new center stack to controllably access 1/ and lower collisionality regimes of NTV theory
- Use 2nd beam line to vary torque profile at fixed power; include in real-time *V*** control
- Apply greater variation of radial profile of braking torque (for *V*** control; test theory)
- Use 2nd SPA power supply unit; add NCC when available (*n*  6) (incremental)
- Consider momentum _*input*_ with NCC via non-resonant NTV (proposal)

## NSTX Disruption Studies Contribute to ITER, Aim to Predict Disruption Characteristics & Onset For Future Large STs

**NSTX Disruption Studies Contribute to ITER, Aim to Predict Disruption Characteristics & Onset For Future Large STs**

- _Halo Current Magnitudes and Scaling_

***Expand these Results For a Complete Characterization of Disruption Dynamics, Including Prediction Methods.***

- (MA2/T)

- Max Halo Current Magnitude (kA)

- Lower Center Stack

- Inner to Outer Vessel

- Vessel Bottom Near CHI Gap

- Outboard Divertor

- Area Normalized Quench Time (msec/m2)

- Pre-Disruption Current Density (MA/m2)

- Fastest NSTX disruption quench times of 0.4 ms/m2, compared to ITER recommended minimum of 1.7 msec/m2.
- Reduced inductance at high-, low-A explains difference

- New instrumentation in 2008 yields significant upward revision of halo current fractions (now up to 20%)
- reveals scaling with IP and BT.
- Mitigating effect: Largest currents for deliberate VDEs
- Toroidal peaking reduced at large halo current fraction.

- _2006 Instrumentation_

- _2008 Instrumentation_

- _Area-normalized (left), Area and __*L*__*ext*__-normalized _
- _(right) I__p__ quench time vs. toroidal __*J*__*p*__ (ITER DB)_

**NSTX**

## Understand the Causes and Consequencs of Disruptions for Next-step STs and ITER

- 2009-2012: Complete halo current characterization
- Install arrays of instrumented tiles in outboard divertor, measure currents into LLD trays (2009-10)
- Utilize CS upgrade to instrument inboard divertor tiles (2011)
- Understand the halo current paths, toroidal peaking physics, and driving mechanisms, in order to make predicitons for future ST plasmas
- 2009-2011: Complete thermal quench characterization
- Determine the fraction of stored energy lost in the thermal quench, compared to that in the pre-disruption phase, over a variety or plasmas and disruptions
- Utilize fast IR thermography to understand time-scale and spatial distribution of the thermal quench heat flux
- Predict the impulsive heat loading constraints on future ST PFCs
- 2010-2013: Learn to predict and prevent disruptions
- Develop real-time diagnostics useful for predicting impending disruptions for relevant ST equilibria and instabilities
- Test predictive algorithms, to determine the simplest, most robust prediction methods
- Use in conjunction with stability models and mode control systems developed

## Experiments in 2008 part of continuing contribution to ITER high priority MHD research

- ELM Mitigation
- Attempted ELM mitigation with non-axisymmetric field from single row of midplane coils
- Several field configs. with Chirikov parameter > 1
- ELM frequency/duration changed, not fully mitigated
- ELMs mitigated by Li evaporation / _triggered_ by applied field
- Neoclassical Toroidal Viscosity (NTV) Study
- Following quantitative agreement on NSTX to best determine impact of non-axisymmetric fields from RMP fields on ITER *V***
- Braking with *n* = even fields confirm non-resonant effect
- Stronger NTV damping with Li evaporation (***i*, or *T**i* variation?)
- RWM stabilization
- Active feedback system latency artificially increased to simulate the effect of greater time delays due to ITER blanket
- Experiment to simulate impartial toroidal coverage of control coils
- Feedback failed for several phase settings (physics reason TBD)

- Suggested by

- NSTX
- PAC

- ITER
- Org.

- USBPO,
- ITER
- Org.

- USBPO,
- ITER
- Org.

- ITER support experiments/analysis to continue in 2009-13

## VALEN RWM control models validated on NSTX predict significant N increase with proposed ITER internal coil

- 3 toroidal arrays, 9 coils each
- ELM, VS, RWM applications
- Endorsed by ITER STAC
- Configuration similar to proposed NCC coil upgrade for NSTX

- 40° sector

- ITER VAC02 design

- passive

- midplane
- coils

- upper+
- lower
- coils

- all
- coils

- _ITER VAC02 stabilization performance_

- VALEN-3D

- N

## U.S. leads the world in high  ST research and is in position to bridge the gap to next-step STs

- Macroscopic stability research direction
- Transition from establishing high beta operation to reliably and predictably sustaining and controlling it – required for next step device
- Research provides critical understanding for tokamaks
- Stability physics understanding applicable to tokamaks including ITER, leveraged by unique low-*A*, and high ** operational regime
- Specific ITER support tasks
- NSTX provides access to best diagnosed high beta plasmas
- _2009-2011_: allows significant advances in scientific understanding of ST physics toward next-steps, supports ITER, and advances fundamental science
- _2012-2013_: allows demonstration/understanding of reliable stabilization/profile control at lower collisionality – performance basis for next-step STs

## Macroscopic Stability Research Timeline (2008-2013)

-  control; fast IR - PFC disrupt. loads; disruption prediction/detection

- Advanced RWM control algorithms

- Integrated code devel/appl: mmVALEN,Hu-Betti W, MARS-K, GTC-Neo NTV, NIMROD, M3D, PEST3, rDCON

- DEFC physics model at high ; optimize DEFC

***Physics***

***Tools***

- Shaping & Control

- N control w/stability models

- Optimize N > 6 w/high S; N control w/rtEFIT

- RWM

- Role of n > 1, stabilize RWM w/NCC

- Adv. RWM control design, NCC dsgn

- Opt. active RWM control, V; mode deform.

- RWM stab. physics (i,V,A, multi-modes)

- two toroidal SXR arrays

- Adv. RWM, DEFC, V ctrl w/ NCC

- 2nd SPA+ NCC

- DEFC

- NTM

- V physics

- Disruptions

- Advanced B correction w/ NCC

- Advanced modeling; mode avoidance and control

- Onset+Small Island Studies; seeding physics

- V1.1

- NTV, INTV w/ V control; 2nd SPA + NCC

- NTV physics, i, n=2-6, INTV, NCC design

- Prediction, Inboard Halo Currents with CS Upgrade

- Halo current, Thermal Quench Measurements

- real-time V control

- 2nd NBI

- non-mag. RWM detection

- real-time MSE

- Adv. RWM control with V, q profile control

- non-mag. RWM control

- non-mag. RWM detect.

- pCHERS, FIDA, high-resolution edge SXR, internal B with MSE

- MSE-LIF, LLD

- advanced rt-V control

- div. halo current

- CS upgrade