# Presentation: 587640

## Macroscopic Stability Research on NSTX and a ReNeWed Future

_**Macroscopic Stability Research on NSTX and a ReNeWed Future**_

- S.A. Sabbagh1, R.E. Bell2,** **J.W. Berkery1, J.M. Bialek1, S.P. Gerhardt2, R. Betti3, D.A. Gates2, B. Hu3, O.N. Katsuro-Hopkins1,** **B. LeBlanc2, J. Levesque1, J.E. Menard2, J. Manickam2, K. Tritz4, and the NSTX Research Team
*1**Department of Applied Physics, Columbia University, New York, NY, USA*
*2**Plasma Physics Laboratory, Princeton University, Princeton, NJ, USA*
*3**University of Rochester, Rochester, NY, USA*
*4**Johns Hopkins University, Baltimore, MD, USA*
** ****Columbia APAM Plasma Physics Colloquium**** **
**February 6, 2009**** **
**Columbia University, New York, NY**

***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***
***ENEA, Frascati***
***CEA, Cadarache***
***IPP, J******ü******lich***
***IPP, Garching***
***ASCR, Czech Rep***
***U Quebec***

***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***

- v1.1

## Key Research Challenge: Develop Stability and Control Understanding to Produce Continuous High Beta Plasmas

- Motivation
- Future spherical torus (ST) magnetic fusion devices plan to run at high ratios of plasma pressure to magnetic field (beta) and with effectively continuous operation
- Mega-Ampere level high beta ST plasmas have been reached
- Attention now turns to stability physics understanding and mode control to maximize steady-state high beta conditions, minimize beta excursions, and largely eliminate disruptions
- Outline
- Present macroscopic stability research on NSTX
- Related research / device upgrades planned for next five years
- Developing longer-term "ITER-era“ research plan through DOE's ReNeW process
- community input strongly encouraged
- conduits for your input and collaborative discussion

## Understanding what profiles and control systems are needed for burning plasmas best occurs before such devices are built 

- _Understanding what profiles and control systems are needed for burning plasmas best occurs before such devices are built__ _

- 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; + physics for ITER
- _Demonstration = __Control_ (of modes and plasma profiles):
- Need to determine what control is needed *before* CTF
- _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 occur below ***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***

- }

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

- Operate at parameters closer to burning plasma (e.g. order of magnitude lower i (PTRANSP) )
- 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

-  New center stack (Bt = 1T, Ip = 2 MA)
- - Liquid Li divertor

- - 2nd NBI (incr.)
-  Singly-powered 	RWM control coils
-  RWM 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 maintain 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***

## Stabilizer plates for kink mode stabilization
External midplane control coils closely coupled to vacuum vessel
Varied sensor combinations used for feedback
24 upper/lower Bp: (Bpu, Bpl)
24 upper/lower Br: (Bru, Brl)

- Stabilizer plates for kink mode stabilization
- External midplane control coils closely coupled to vacuum vessel
- Varied sensor combinations used for feedback
- 24 upper/lower Bp: (Bpu, Bpl)
- 24 upper/lower Br: (Bru, Brl)

- _NSTX equipped for passive and active RWM control _

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

- n = 1 active, n = 3 DC control
- n = 1 response ~ 1 ms < 1/RWM
-  N/Nno-wall = 1.5 reached
- best maintains ****
- NSTX record pulse lengths
- limited by magnet systems
- n > 0 control first used as standard tool in 2008
- Without control, plasma more susceptible to RWM growth, even at high ****
- Disruption at ****/2 ~ 8kHz near q = 2
- More than a factor of 2 higher than marginal **** with n = 3 magnetic braking

- N

- Bpu,ln=1(G)

- IA (A)

- t (s)

- 129067

- 129283

- n = 1 feedback

- n = 3 correction

- /2 (kHz)

- N > Nno-wall

-  maintained

- Nno-wall = 4

- (DCON)

- With control

- Without control

- (Sabbagh, et al., PRL **97** (2006) 045004.)

## Probability of long pulse and <N>pulse increases significantly with active RWM control and error field correction

- Standard H-mode operation shown
- Ip flat-top duration > 0.2s (> 60 RWM growth times)

- Ip flat-top duration (s)

- Frequency distribution

- Control off (908 shots)

- Control on (114 shots)

- Control on

- Control off

- Control allows <N>pulse > 4
-  N averaged over Ip flat-top

## During n=1 feedback control, unstable RWM evolves into rotating global kink

- Bpn=1(G)

- IA(kA)

- Bn=odd(G)

- Bpn=1(deg)

- RFA

- RFA reduced

- Mode rotation
- Co-NBI direction

- RWM

- 128496

- t (s)

- RWM grows and begins to rotate
- With control off, plasma disrupts at this point
- With control on, mode converts to global kink, RWM amplitude dies away
- Resonant field amplification (RFA) reduced
- Conversion from RWM to rotating kink occurs on w timescale
- Kink either damps away, or saturates
- Tearing mode can appear during saturated kink

## Soft X-ray emission shows transition from RWM to global kink

- Tearing mode appears after 10 RWM growth times and stabilizes

- _Soft X-ray emission shows transition from RWM to global kink_

- edge

- core

- t (s)

- 128496

- spin-up

- RWM

- kink

- edge

- core

- RWM onset time + 35 ms

- _Transition from RWM to kink_

- _Tearing mode appears during kink_

- edge

- core

- 0.645

- 0.646

- t (s)

- q = 2

- filtered
- 1 < f(kHz) < 15

## Low , high N plasma not accessed when feedback response sufficiently slowed

- _Low ____, high ___N_ plasma not accessed when feedback response sufficiently slowed_

- Low **** access for ITER study
- use n = 3 braking
- n = 1 feedback response speed significant
- “fast” (unfiltered) n = 1 feedback allows access to low *V***, high N
- “slow” n = 1 “error field correction” (75ms smoothing of control coil current) suffers RWM at **** ~ 5kHz near q = 2

- RWM

- n = 3 correction

- n = 3 braking

- slow feedback

- fast feedback

- Brun=1(G)

- Bpln=1(G)

- N

- IA (kA)

- /2 (kHz)

- 130639
- 130640

## Low , high N plasma not accessed when two feedback control coils are disabled

- _Low ____, high ___N_ plasma not accessed when two feedback control coils are disabled_

- Low **** access for ITER study
- use n = 3 braking
- n = 1 feedback doesn’t stabilize plasma with 2 of 6 control coils disabled
- scenario to simulate failed coil set in ITER
- Feedback phase varied, but no settings worked
- RWM onset at identical time, plasma rotation

- RWM

- Brun=1(G)

- Bpln=1(G)

- N

- IA (kA)

- /2 (kHz)

- 130641
- 130640
- 130642
- 130643

- 2 control coils out
- feedback phase varied

- All coils on

- n = 3 correction

- braking and feedback

## Experimental RWM control performance consistent with theory

- _Experimental RWM control performance consistent with theory_

- VALEN code with realistic sensor geometry, plasmas with reduced V

- Feedback phase scan shows superior settings

## Significant N increase expected by internal coil proposed for ITER

- _Significant ____N__ increase expected by internal coil proposed for ITER_

- 50% increase in N over RWM passive stability

- (40° sector)

- ITER VAC02 design

- passive

- midplane
- coils

- upper+
- lower
- coils

- all
- coils

- _ITER VAC02 stabilization performance_

- VALEN-3D

- N

- 3 toroidal arrays, 9 coils each

- Growth rate (s-1)

## 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). )

## VALEN computed RWM stability for proposed RWM control coils upgrade - behind passive plates (PP)

- N

- growth rate  [1/s]

**passive**

**Ideal wall limit**

_**Copper Plates**_

**Coils behind secondary PP**

**Coils behind primary PP**

**External coils**

- N

- growth rate  [1/s]

**passive**

**Ideal wall limit**

_**Stainless Steel Plates**_

**Coils behind SS secondary PP**

**Coils behind SS primary PP**

- coils behind copper passive plates perform worse than existing external RWM coil set

- change copper passive plates to SS RWM performs better  than existing external coil set

**External coils**

- (note: idealized sensors used)

## Proposed control coils on plasma side of copper passive plates computed to stabilize to 99% of Nwall

- coils on plasma side Cu secondary PP stabilize to N = 7.04
- coils on plasma side Cu primary  PP stabilize to  N = 7.05
- Ideal wall limit N = 7.06

- (note: idealized sensors used)

- VALEN

## Modification of Ideal Stability by Kinetic theory (MISK code) investigated to explain experimental RWM stabilization 

- Simple critical **** threshold stability models or loss of torque balance do not describe experimental marginal stability
- Kinetic modification to ideal MHD growth rate
- Trapped and circulating ions, trapped electrons
- Alfven dissipation at rational surfaces
- Stability depends on
- Integrated _**__**__ profile_: resonances in ***W**K* (e.g. ion precession drift)
- Particle _collisionality_

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

- Energy integral

- _collisionality_

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

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

- Sontag, et al., Nucl. Fusion **47** (2007) 1005.

- _precession drift_

- _bounce_

- _M__odification of __I__deal __S__tability by __K__inetic theory (__MISK__ code) investigated to explain experimental RWM stabilization _

## Kinetic modifications show decrease in RWM stability at relatively high V – consistent with experiment

- _Kinetic modifications show decrease in RWM stability at relatively high V__ – consistent with experiment_

- Marginal stable experimental plasma reconstruction, rotation profile exp
- Variation of  away from marginal profile _increases_ stability
- Unstable region at low 

- _Theoretical variation of ___

- Marginally stable
- experimental
- profile

- 121083

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

- /a

- /exp

- 2.0

- 1.0

- 0.2

- /2 (kHz)

- Im(WK)

- Re(WK)

- experiment

- 2.0

- 1.0

- 0.2

- unstable

## Kinetic model shows overall increase in stability as collisionality decreases

- _Kinetic model shows overall increase in stability as collisionality decreases_

- Vary  by varying T, n at constant 
- Simpler stability dependence on  at increased 

- _Increased collisionality (x6)_

- Im(WK)

- Re(WK)

- unstable

- _Reduced collisionality (x1/6)_

- w

- /exp

- w

- Re(WK)

- unstable

- 0.2

- 1.0

- 2.0

- 0.2

- 1.0

- 2.0

- Increased stability at /exp ~ 1
- Unstable band in  at increased 

## Hot ions have a strongly stabilizing effect on DIII-D

- without
- hot ions

- with hot ions

***(J.W. Berkery, Mode Control Mtg. 2008)***

- MISK show band of instability at moderate rotation without hot ions, but complete stability with hot ions: (w) ~ 1.0
- DIII-D shot 125701 @ 2500ms and rotation from 1875-2600ms
- This may explain why DIII-D is inherently more stable to the RWM than NSTX
- energetic particle modes might “trigger” RWM by fast particle loss (JT-60U IAEA 08)

## NSTX RWM stability with hot ions under evaluation

***121083***

- Pfast

- Pressure (kPa)

- Ptot

- High level of
- Pfast in NSTX

- Pfast

- Direct effect on calculated growth rate using test profiles (based on TRANSP analysis) for hot ion pressure: (w) ~ 0.2
- Stability using TRANSP runs of RWM marginally stable plasmas now underway
- TRANSP hot ion population smaller at edge in NSTX vs. DIII-D – may explain why RWM apparently less stable in NSTX

- unstable

- Variation due to Pfast

## Lithium wall conditioning, n=1 RWM control, n=3 error correction also shown to control (eliminate) tearing modes

- Physics of tearing mode elimination still under investigation
- Full suppression of modes not seen on all shots
- If lithium wall conditioning a key element, liquid lithium divertor might be used for NTM control

***MHD spectrogram with lithium, n=1 feedback and n=3 correction***

***MHD spectrogram w/o n=1 feedback and n=3 correction***

**n=1 mode drops ******

***CHERS v******t****** at R = 139cm***

**Red with control**
**Black w/o control**

**Red with control**
**Black w/o control**

**No MHD, ******** and rotation maintained**

## Required drive for NTM onset better correlated with rotation shear than rotation magnitude

*NTM Drive at Onset Only Poorly Correlated with q=2 (Carbon) Rotation*

*NTM Drive at Onset Better Correlated with Local Flow Shear*

- For fixed V, order of increasing onset drive: EPM triggers, ELM triggers, and “Triggerless”
-  All trigger types have similar dependence on flow shear
- Dependence likely to related to intrinsic tearing stability, not triggering

- S.P. Gerhardt, submitted to Nucl. Fusion

## 2nd NBI and BT = 1T with center stack upgrade to be used for study and control MHD modes (and much more...)

- Fully non-inductive scenarios require 2nd NBI (7-10MW of NBI heating) for H98  1.2
-  CR will increase from 0.35  1s if Te doubles at lower ne, higher BT
-  Need 3-4 CR times for J(r) relaxation  5s pulses  need 2nd NBI

- Above:  N=5, T=10%, IP=0.95MA
- N=6.1, T=16%, qmin > 1.3, IP=1MA at BT=0.75T possible

- qmin > key rationals 1.5, 2 to be used for NTM control

## 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.
- Expand studies on NSTX
- Examine larger field spectrum
- Improve inclusion of plasma response using IPEC
- Consider developments in NTV theory
- Reduction, or saturation due to Er at reduced ion collisionality, multiple trapping states, bounce/precession resonances, superbanana regime, etc.
- Some effects suggest continued increase in viscosity at reduced i
- Examine NTV from magnetic islands

- _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..._

- _...__will it saturate, decrease at lower ____i __?_

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

- e.g. A.M. Garofalo, APS 2008 invited (DIII-D)

- J.K. Park, APS 2008 invited talk

- No V shielding in core;
- used Shaing erratum

- theory

- measured

- TNTV (N m)

- R (m)

- axis

## Stronger non-resonant braking at increased Ti

- _Stronger non-resonant braking at increased T_i

- Observed non-resonant braking using n = 2 field
- Examine Ti dependence of neoclassical toroidal viscosity (NTV)
- Li wall conditioning produces higher Ti in region of high rotation damping
- Expect stronger NTV torque at higher Ti    (-*d******/dt ~ T**i**5/2 *****)
- At braking onset, *T**i* ratio5/2 = (0.45/0.34)5/2 ~ 2
- Consistent with measured *d******/dt *in region of strongest damping

- Li wall

- no Li

- n = 2 braking

- 130720

- 130722

- no lithium

- Li wall

- R = 1.37m

- Icoil (kA)

-  (kHz)

- 0.0

- -0.4

- -0.8

- 4

- 2

- 0.4

- 0.3

- 0.1

- 0.2

- t (s)

- Ti (keV)

- 0.9

- 1.1

- 1.3

- 1.5

- 0.9

- 1.1

- 1.3

- 1.5

- R(m)

- R(m)

- _(T_i_ ratio)__5/2_

- (1/)(d/dt)

- Li wall

- _Damping profiles_

- No Li

- 2x

## n = 2 non-resonant braking evolution distinct from resonant

- _Non-resonant_:
- broad, self-similar reduction of profile
- Reaches steady-state (t = 0.626s)

- 128882

- R(m)

- (kHz)

- 128882

- t = 0.516s

- t = 0.466s (t = 10 ms)

- _Resonant_:
- Clear momentum transfer across rational surface
- evolution toward rigid rotor core
- Local surface locking at low 

- R(m)

- Steady-state profile
- (from non-resonant braking)

- t = 0.626s
- (t = 10 ms)

- t = 0.816s

- outward
- momentum
- transfer

- N ~ 3.5

## High  ST research plan focuses on bridging the knowledge gaps to next-step STs; contributes to ITER

- 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 well diagnosed high beta ST 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

## DOE ReNeW process to define “ITER-era” (20 yr) research program 

- ReNeW: Research Needs Workshop
- To inform the Office of Fusion Energy Sciences (OFES) in preparing a strategic plan for research in each major area of the Fusion Energy Sciences Program
- To allow U.S. fusion community to explain research goals, methods to achieve them
- Including communication to new administration
- MAIN WEB PAGE: _http://burningplasma.org/renew.html_
- Document
- _Vol 1_: Define scientific research needed to fill “gaps” in present understanding
- “gaps” defined in / modified from Greenwald report, FESAC TAP reports
- Divided into 5 “themes” comprising magnetic fusion research
- _Vol 2_: Define (~ 15) “research thrusts” that will carry out this research
- Basis for detailed program plan to be constructed by OFES

## ReNeW organized into 5 fusion research themes

- _Structure/gaps_
- FESAC Toroidal Alternates Panel Report

- _Structure/gaps_
- Priorities, Gaps, and Opportunities Panel Report
- (“Greenwald Report”)

- _Structure/gaps_
-  Energy Policy Act task group report

- Reports available at: http://burningplasma.org/renew.html

- Spherical Torus sub-theme

## ReNeW ST Panel is on Schedule to Complete Tasks

- _ReNeW ST Panel is on Schedule to Complete Tasks_

- Tasks through March 16-19 Workshop
- Solicit community input: (First call for input DONE – continue to engage community)
- Review issues as described in TAP panel report: (DONE: embodied in community distributed draft of ST section V1.7)
-       ********** WE ARE HERE **********
- Identify scientific research needed to address the issues
- Review and expand on research outlined in the TAP panel report
- Draft write-up of research requirements, make available to community
- Fold in community input on research requirements
- Develop draft “research thrusts” for discussion at March workshop

- _FESAC TAP Report Mission statement_: *Establish the ST knowledge base to be ready to construct a low aspect ratio component testing facility that provides high heat flux, neutron flux, and duty factor needed to inform the design of a demonstration fusion power plant.*

## Interpretation of the FESAC TAP document by some people in the community

- _Pros_
- ST-CTF focus
- _Cons_
- Alienates a significant part of the community (research plan has been characterized by some (to quote) as a “dead end”), so loses potential constituency
- Many have complained that several physics issues have been “swept under the rug”, even at the level of an ST-CTF device
- Interpreted as above, it doesn’t maximize cross-cutting with other magnetic fusion research

- Present STs

- DEMO

- (Research informing DEMO)

- ST-CTF

- Key point: Let’s engage the community and make appropriate, small changes to strengthen these weak points.

## (STRAWMAN) U.S. ST Research Vision consistent with Present ST Mission Statement

- _(STRAWMAN) U.S. ST Research Vision consistent with Present ST Mission Statement_

- ReNeW ST Panel v1.2

- Present STs

- DEMO

- _Scientific Research during ITER era informing DEMO_

-  Blanket development (magnetic, inertial fusion)
-  Fusion materials development

- Potential
- application

- Fusion/Fission
- Hybrid driver

-  Nuclear waste processing
- ...

- ST-CTF

- Potential
- application

- Upgraded
- ST Facilities

- ST Performance
- Extension
- Facility

- ?

- _Tier 1_
-  Start-up and Ramp-up
-  Plasma-material interface
-  Electron energy transport
- - Magnets

- _Tier 2_
-  Stability & SS Control; 3D fields
-  Disruptions
-  Heating & current drive
-  Ion-scale transport
- - Fast particle instabilities

- _Tier 3_
-  NTMs
-  Continuous NBI systems

*Research*

*Facilities and Applications*

## Several Conduits for Participation in ReNeW Process

- ReNeW Forum (web bulletin board)
- Contribute to open discussions; start your own discussions
- Registration instructions: _http://burningplasma.org/forum/_
- Request authorization to ReNeW Forum: e.g. email: _sabbagh@pppl.gov_
- ST topic: _https://burningplasma.org/forum/index.php?showforum=114_
- ST Group: Direct input to/discussion of evolving draft ST section of document
- Posted: _https://burningplasma.org/forum/index.php?showtopic=653_
- Submit short white papers describing your ideas on how to resolve key issues, support/define research thrusts
- Download directions at: _http://burningplasma.org/renew.html_
- Participate in March 2009 Workshops
- Links to websites with info/registration at: _http://burningplasma.org/renew.html_
- Directly contact panel members
- ST Group: 10 Panel members (see next page); more than 30 advisors

## ReNeW Spherical Torus Panel Members

- _ReNeW Spherical Torus Panel Members_

- Contact any panel member for authorization to access the ReNeW Forum (website bulletin board)
- Full advisor list posted at: _https://burningplasma.org/forum/index.php?showtopic=636_

## Backup Slides

## 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
- 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: 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: 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