# Presentation: 539442

## Progress towards steady state on the National Spherical Torus Experiment (NSTX)

**Progress towards steady state on the National Spherical Torus Experiment (NSTX)**

- Presented by David A. Gates*
- for the NSTX Group
- *Princeton Plasma Physics Laboratory, Princeton, NJ
- at
**MIT PSFC**
- March 9, 2007
- Cambridge, MA

***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***
***JAERI***
***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***
***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***

## Outline

- Introduction
- Progress towards steady state operation
- MHD stability
- Transport
- Non-inductive startup
- Wave physics
- Boundary physics

## The Spherical Torus (ST) is a Low Aspect Ratio Tokamak

- • ST maximizes the field line length in the good curvature
- (stable) region - strong toroidicity
-      - Benefit to stability and confinement
- • Strong natural shaping
-      - Enhances stability and bootstrap current

## NSTX Addresses Key Issues for Fusion Energy Development, ITER Physics and Plasma Science

- Enable attractive CTF for success of DEMO: addresses unique ST development issues, synthesis of ST and tokamak results
- Support USBPO and ITPA activities: can address physics in ITER physics regimes
- Complement and extend toroidal confinement physics: operating regimes unique to ST, overlap with that of conventional aspect ratio tokamaks

- Major Radius R0		0.85 m
- Aspect Ratio A		1.3
- Elongation 		2.8
- Triangularity 		0.8
- Plasma Current Ip		1.5 MA
- Toroidal Field BT		0.55 T
- Pulse Length		1.5 s
- NB Heating (100 keV)	7 MW
- T,tot	 		up to 40%

## STs Can Lead to Attractive Fusion Systems

- Component Test Facility (CTF) will be needed after ITER to carry out integrated DEMO power testing and development
- ST enables highly compact CTF with full remote maintenance and high duty factor, and it provides potentially attractive reactor configuration
- Issue: Plasma formation and current ramp-up

- Peng et al, PPCF **47**,** **B263** **(2005)

## AT/ST requires high bootstrap fraction, fbs, simultaneous with high t

- High *f**bs* and high ***t** * competing requirements (at fixed shape and ***N*)
- Progress for ST and advanced tokamak given by the “sustainable ***t*”    ***sus** **** f**bs****t** ~ S****N**2*, where *S = q**95**(I**p**/(aB**t**))] *is the “shaping factor”
- If ***Nmax** = C**Troyon*, then only shape improves ***sus*
-  ***sus* increases linearly with increasing *S [S = q**95**(I**p**/(aB**t**))]*
- Component Test Facility requires ***t* ~ 20% and *f**bs* ~ 50%, ***sus* = 10%
- ARIES-ST requires ***t* ~ 40% and *f**bs* ~ 90%
- NSTX has optimized shaping with new PF coils for high triangularity and elongation
- NSTX has achieved record values of elongation and shape factor
- Leads directly to record values of the ***sus* for the ST
- For NSTX 100% non-inductive operation with ***N* ~ 7 only with strong shaping

## Continuous shaping enhancements have enabled progress towards steady state

** = 2.75, **= 0.8, S ~ 37

** = 1.8, ** = 0.6, S ~ 22

** = 2.0, ** = 0.8, S ~ 23

** = 2.3, **= 0.6, S ~ 27

** = 3, **= 0.8,  S ~ 41

- Shaping factor has nearly doubled since 2001 - record ** ~ 3
- Major upgrades include
- Power supply modifications
- Control system upgrades
- PF Coil modifications
- Real time equilibrium reconstruction for shape control

- 2001

- 2005

- 2004

- 2002-3

- 2006

## Reduction in control system latency increases elongation (2004)

- Vertical instability determines upper bound on boundary elongation        *** = b/a*
- Control system latency (time between an event and the mitigating control action) major factor in vertical controllability
- Faster control system leads to better plasmas

- Plasma elongation (**) plotted versus
- normalized internal inductance (*l**i*)

- Each point in the plot represents one EFIT equilibrium in the NSTX database

## Upgrade of PF1A coil has enabled simultaneous achievement of high  and 

- Triangularity [=(Rx-R0)/a] increases edge q for given *I**p**, B**t* by increasing length of the plasma boundary on high field side
- Location of X-point (*R**x**,Z**x*) determined on NSTX by PF1A coil
- Coil was modified to enable simultaneous achievement of high  and 

- Modified PF1A Coil

- Old PF1A Coil

## Implementation of rtEFIT improves shape control reproducibility

- Boundary overlay time window

- rtEFIT used for boundary control
- Define locations where boundary should be, control on difference between flux at the control points and flux at the x-point
- Collaboration with General Atomics - very successful
- Has lead to marked improvements in reliability  of shape control

- ~3cm includes MHD perturbations

- Control points

- Gates, PoP 2006

## Shape control modifications substantial increase in NSTX operating space

- Strong correlation between improved shaping and the achieved plasma performance
- Improvements in 2006 are due to improved reliability at high shaping (rtEFIT)

- Pulse averaged t versus pulse length
-  Data are sorted by year and by S

- Pulsed average ***1/2****p*******t** *vs* S *(All years)

## High Performance Can Be Sustained For Several Current Redistribution Times at High Non-Inductive Current Fraction

** ********p and NBI current drive provide up to 65% of plasma current ******
-        Relative to earlier results, High N  H89P now sustained 2  longer

***116313G12***

** **
- TRANSP non-inductive current fractions
** **

- D. Gates, PoP **13**, 056122 (2006)

## MHD-Induced Redistribution of NBI Current Drive Contributes to NSTX “Hybrid”-Like Scenario as Proposed for ITER

- qmin>1 for entire discharge, increases during late n=1 activity

- High anomalous fast ion transport needed to explain neutron rate discrepancy during *n*=1

- Fast ion transport converts peaked JNBI to flat or hollow profile
- Redistribution of NBICD makes predictions consistent with MSE

- J. Menard,  PRL **97**, 095002 (2006)

**n=1 mode onset**

**n=1 mode onset**

## Integrated Modeling Points to Importance of Shaping, Reduced ne, and Increased Te/E for Higher fNI and High N

- n(0)=0.75e20

- n20(0)=0.85,
******=2.2 **
**H****98****=1.1 **
******N**** = 5.6**
**q(0) = 1.15**
- n20(0)=0.36,
******=2.2**
**H****98****=1.1 **
******N**** = 5.6**
**q(0) = 1 @ 0.8 s**
- n20(0)=0.75,
******=2.55**
**H****98****=1.35**
******N**** = 6.6**
**q(0) = 1.4**

*Integrated Modeling Points to Importance of Shaping, Reduced n**e**, and Increased T**e**/****E** for Higher f**NI** and High ****N*

## Fully Non-Inductive Scenario at Higher N Requires Higher Confinement, Higher q, Strong Plasma Shaping

- Need 60% higher T, 25% lower ne
- Lithium?
- higher q0  qmin  2.4 (higher with-wall limit N _<_ 7.2)

******= 2.3, ********X-L**** = 0.75**
** ********R****SEP**** = -1cm**

******= 2.6, ********X-L**** = 0.85**
** ********R****SEP**** = -2mm**

-  Higher  for higher q, P, fBS
** **High  for improved kink stability*** ***

## Dynamic Error Field Correction (DEFC) Extended Pulse

- J. Menard et al., IAEA 2006

- Plasma rotation sustained after
- correction of intrinsic error fields

**• ****Six EF/RWM Coils powered by “SPA” supplies with up to 6 kA-turn currents at 1 kHz **
**• ****Supported by unique RWM sensors (24 B****R**** and 24 B****Z**** internal coils + over 150 other magnetic sensors) with  51 ch. **

## RWM Actively Stabilized at Low, ITER-Relevant Rotation

- First demonstration in an ST
- Plasma rotation reduced by non-resonant n=3 magnetic braking
- No-wall -limit computed by DCON

- Sabbagh et al., PRL 97 (3006) 04500

-  Optimize RWM control
-  Fully understand stabilization
-   physics

## Observed Rotation Follows Neoclassical Toroidal Viscosity (NTV)Theory

- First quantitative agreement with NTV theory
- Due to plasma flow through non-axisymmetric field
- Trapped particle, 3-D field spectrum important
- Computed using experimental equilibria
- Viable physics for simulations of rotation dynamics in future devices (ITER, CTF)

- Magnetic braking due to
- applied n=3 field

- Zhu et al., PRL 96 (2006) 225002

## Both Internal and External Modes Can be -Limiting in NSTX

- Resistive Wall Modes can
- limit T at low-q
- (Sabbagh et al., NF, 44 [2004] 560)

*Maintaining high rotation is a key to stabilizing both modes*

- Non-linear M3D results consistent with experiment

******T ******** 23%**

******T ******** 31%**

******T ****(%)**

**Mode B********(Gauss)**

**q****0**
**(w/o MSE)**

**f********(0) (kHz)**

- Discharge (in black) collapses as rotation flattens and decreases

## NSTX Addresses Transport & Turbulence Issues Critical to Both Basic Toroidal Confinement and Future Devices

- NSTX offers a novel view into plasma T&T properties
- NSTX operates in a unique part of dimensionless parameter space: R/a, T, (, 
- Dominant electron heating with NBI: relevant to -heating in ITER

- Excellent laboratory in which to study electron transport: electron transport anomalous, ions close to neoclassical
- Large range of T spanning e-s to e-m turbulence regimes
- Strong rotational shear that can influence transport
- Localized electron-scale turbulence measurable (e ~ 0.1 mm)

## Dedicated H-mode Confinement Scaling Experiments Have Isolated the BT and Ip Dependences

- Scans carried
- out at constant
- density, injected
- power (4 MW)

- 0.50 s

- 0.50 s

## Dedicated H-mode Confinement Scaling ExperimentsHave Revealed Some Surprises

*Dedicated H-mode Confinement Scaling Experiments**Have Revealed Some Surprises*

-      ***Strong dependence of ************E****** on B******T***

- E,98y,2 ~ BT0.15

- H98y,2 ~       0.9    →  1.1    →  1.4

- 4 MW

- E,98y,2 ~ Ip0.93

- H98y,2 ~       1.4   →  1.3   →  1.1

-      ***Weaker dependence on I******p***

- E~Ip1.3-1.5 at fixed q                  E,98y,2~Ip1.1 at fixed q

- NSTX E exhibits strong scaling at fixed q

- 4 MW

## Variation of Electron Transport Primarily Responsible for BT Scaling

*Variation of Electron Transport Primarily **Responsible for B**T** Scaling*

- Broadening of Te & reduction in e outside r/a=0.5 with increasing BT

- Ions near neoclassical

- Neoclassical

## Ion Transport Primarily Governs Ip Scaling- Ions Near Neoclassical Level -

*Ion Transport Primarily Governs I**p** Scaling**- Ions Near Neoclassical Level -*

- i,GTC-NEO
-     (r/a=0.5-0.8)

- _GTC-Neo_ neoclassical:
- includes finite
- banana width
- effects (non-local)

## Turbulence Measurements + Gyrokinetic Calculations Have Helped Identify Possible Sources of Transport

*Turbulence Measurements + Gyrokinetic Calculations Have Helped Identify Possible Sources of Transport*

- Ion transport during
- H-phase is
- neoclassical

- Electron transport
- reduced, but remains
- anomalous

- Ion and electron transport change
- going from L- to H-modes

- - Localized measurement (axis to edge)
- - Excellent radial resolution (6 cm)

## Strongly Reversed Magnetic Shear L-mode Plasmas Achieve Higher Te and Reduced Transport

- Linear GS2 calculations indicate
- reduced region of tearing instability
- for RS plasma

- GS2 calcs also indicate ETG stabilized by RS

- F. Levinton, APS 2006

## CHI Is Being Used for Solenoid-Free Startup

- Transient CHI: Axisymmetric reconnection leads to the formation of closed flux surfaces

## 160 kA of Closed Flux Current Produced in NSTX by Transient CHI

- 2006 discharges operated at high RF and injector flux
- Magnetic sensors and flux loops used in reconstruction
- Plan to optimize high current CHI discharges, couple to OH, HHFW

- R. Raman et al., PRL 97 (2006) 175002

- Ip decays after ICHI→0 High Ip flux closure

- EFIT reconstruction

## EBW/HHFW Coupling Studies Being Carried Out

*EBW/HHFW Coupling Studies Being Carried Out*

- L-mode

- B-X-O mode coupling understood in
- L-mode at fce; coupling in H-mode low

- HHFW htg efficiency improved at high BT and k||

## NSTX Accesses ITER-Relevant Fast-Ion Phase-Space Regime

***1% neutron rate decrease:***

***5% neutron rate decrease:***

** ****ITER** will operate in new, small * regime for fast ion transport
- k ≈ 1 means "short" wavelength Alfvén modes
- Fast ion transport expected from _**interaction of many modes**_
- NSTX can access multi-mode regime via high fast / total and vfast / vAlfven

*NSTX Accesses ITER-Relevant Fast-Ion Phase-Space Regime*

- E. Fredrickson, Phys. Plasmas **13**, 056109 (2006)

## Alfvén Cascades (RSAE) Observed at Low e on NSTX (also on MAST)

- Frequency chirp indicates evolution of qmin
- 	Use for q-reconstruction, MSE verification

## Angelfish Identified as Form of Hole-Clumps, Consistent with Theory

*Angelfish Identified as Form of Hole-Clumps, Consistent with Theory*

- Mode satisfies Doppler-shifted resonance condition for TRANSP calculated fast ion distribution
- Growth rate estimates from theory is 0.04; from observation is 0.053
- Engineering of fast-ion phase space can suppress deleterious instabilities

- Berk et al., (PoP, ‘99)

## Peak Heat Flux Can Be Reduced By Plasma Shaping

- Flux expansion decreases peak heat flux despite reduced major radius
- Compare **single-null** & double-null configurations with triangularity  ≈ 0.4 at X-point and **high triangularity ********=0.8 double-null** plasmas
- Measure heat flux with IR thermography of carbon divertor tiles
- Peak heat flux decreases as **1**** : ****0.5**** : 0.2**
- ELM character changes: Type I  Mixed  Type V

## Peak Heat Flux Can Be Reduced With No Loss of Confinement

## Lithium Evaporator (LITER) Produced Particle Pumping and Improved Energy Confinement in H-mode Plasmas

- Wtot 20% higher in post-Li
- H98y,2=1.1→1.3 post-Li

## Edge Imaging Has Been Key to Studying Edge Turbulence Phenomena (Blobs, ELMs) 

- Excellent “blob” measurements
- have allowed connection to theory

- ELM dynamics and rotation
- have been measured

## Summary

- NSTX normalized performance approaching ST-CTF level
- Only ST in world with advanced mode stabilization tools and diagnostics
- Unique tools for understanding core and edge transport and turbulence
- Uniquely able to mimic ITER fast-ion instability drive with full diagnostics
- Improved understanding of HHFW and EBW coupling/heating efficiency
- Demonstrated 160kA closed-flux plasma formation in NSTX using CHI
- Developing understanding and unique tools for heat flux and particle control