NSTX-U: An Essen.al Science
Facility for US Fusion Innova.on
Na.onal Academy Panel Visit to PPPL
S. Gerhardt, for the NSTX-U Team
April 11, 2018
Princeton Plasma Physics Laboratory

Here is why we are excited for NSTX-U• NSTX-U contributes to optimization of the tokamak
concept though optimization of aspect ratioà potentially transformational behavior when approachingfusion relevant plasma parameters• NSTX-U is a Key Component of the PPPL
Programmatic Strategy Towards Fusion Powerà Supports our vision for innovation in confinement
optimization, materials• NSTX-U has a unique role within the world ST
program• à Unique focus on core physics & impact of Li on
plasma behavior• NSTX-U will provide answers to critical fusion
science questions early in its operation.
• à Exciting science from the commissioning
campaign

2

Outline• Why study spherical tori/tokamaks (STs)à And why use NSTX-U for that mission.

• Brief Discussion of NSTX-U OperaFons andRecovery• Key ScienFﬁc Deliverable of the NSTX-U Program• NSTX-U in the Larger ST/Fusion Context
3

Aspect ra.o is important tokamak geometricalparameter that has not yet been fully explored
Aspect raFo A = R / a
R = Major radiusa = Minor radius
Conven&onal
(A=4)
a

a

NOTE: Cross-sections illustratedifferences in natural elongation.
All modern tokamaks maximize elongation
to increase stability, confinement

Spherical(A=1.25)

Spherical tokamak / torus (ST) A < 2
Conven.onal tokamak typically A > 2.5
4

Spherical shape increases plasma stability• Key plasma physics and fusion parameter isnormalized pressure = β ∝ p / B2
Plasma pressure MagneFc pressure
Magne.c ﬁeld line

Stable region

Unstable region

ST ﬁeld line trajectoryfavors the stabilizing
good curvature regionàHigher βàChanges turbulence,
conﬁnement proper.es

A=4, κ=2

A=1.3, κ=3

5

STs access very high toroidal βtD.J. Schlossberg, PRL 2017

Pegasus (US): βt ~ 30-100%

βt

(%)

A=1.25
B=0.15T

A=1.35
B=0.3TSTART (UK), NSTX (US): βt ~ 30-40%
MAST (UK): βt ~ 15-20%

A=1.45
B=0.4T

Increasing A and B

A=1.1-1.2
B=0.025T

DIII-D (US): βt ~ 12%
Conventional A=2.5, B=0.8T

Normalized Current
IN = IP / aBT (MA / m⋅T)
STs can maintain ahigher βN = βt/(IP/aBT)

Key Science Question:
What Aspect Ratio and Field
Maximizes the Pressure(βB2) in Steady State?
6

Data indicates ST conﬁnement improves towards fusion
relevant collisionality
ST scaling: τE, th ∝ νe* -0.8 β-0.0
Does favorable ST conﬁnement
scaling with ν* conFnue at lowerν*?
• Only 20% diﬀerence in scaling
values in NSTX regime• NSTX-U will access higher Te
and lower ν* to disFnguishbetween scalings
Could this be a game changer for
conﬁguraFon opFmizaFon?

ITER basis: τE, th ∝ νe* -0.1 β-0.9

BTτE
[T-s]

Normalized electron collisionality
νe* ∝ ne / Te2
7

NSTX-U vital for addressing key ST / fusion ques.ons
Will access new physics with 2 new tools:

Highest normalized pressure athigh T & Low Collisionality

Sustain plasma without transformer

1. New Central Magnet

2. Tangential 2nd Neutral Beam

2× field, current, power
4× heat flux, 5× pulse length
Up to 10× higher nTτE

8

NSTX-U is one of the two largest facili.es
in a world-wide ST research program
NSTX-U, USA

PEGASUS, USA

PI3,
CanadaLTX-β / CDX-U, USA

Proto
ST40, UK Sphera, Italy

GLOBUS-M2, Russia

SUNIST, China

MAST-U, UK

KTM,
Kazakhstan

QUEST/CPD, Japan
VEST, Korea

HIST, Japan

LATE, Japan

TST-2, Japan

UTST, Japan

TS3/4, Japan

17 interna.onal facili.es for ST research and broader fusion science
9NSTX-U and MAST-U are the most capabledevices in a world-wide ST research program
MAST-U (UK)NSTX-U
Similar features:• Major radius R = 0.8-1m• Plasma current up to 2MA• Pulse durations 1s à up to 5s• Strong neutral beam heating

Complementary Research:Core emphasis• Highest plasma beta in large

Boundary emphasis• “Long-leg” divertor for powerST, facilitated by passive
exhaust research
conductors and RWM control • Only large ST with oﬀ• Only large ST with RF heating
midplane 3D magnetic field
coils
10

STs will provide leading contribu.ons to development
and understanding of advanced divertors
NSTX-U: Short-leg flared divertor +radiation to mitigate heat flux

MAST-U: World-leading pumped longleg + flexible flaring, radiation

Super-X divertors

Conventional
Divertor

Importance of this research elevated by research showing
narrow scrape-oﬀ-layers at high plasma current (poloidal ﬁeld)
11

NSTX-U design enables access to 2-3× higher
plasma pressure, temperature than MAST-U
NSTX-U NSTX-U central magnet
provides 1.5× higher
toroidal ﬁeld current
à ~1.5 - 2× higher BT2

MAST-U(depending on plasma shape)

Conducting
plates can
suppress global
kink instabilities,~1.5× higher βTp ∝ βT BT2
2-3× higher
Allows an extended range of collisionality, higher self-driven currents
12Outline• Why study “spherical tori/tokamaks” (STs)à And why use NSTX-U for that mission.

• Brief Discussion of NSTX-U OperaFons andRecovery• Key ScienFﬁc Deliverable of the NSTX-U Program• NSTX-U in the Larger ST/Fusion Context
13

In FY2016, NSTX-U had scien.ﬁcally produc.ve 1st year
NSTX-U in high demand: ~4x more run-time requested than available• Achieved H-mode on 8th day of 10 weeks of operation• Surpassed magnetic field and pulse-duration of NSTX• Matched best NSTX H-mode performance at ~1MA
• Identified, corrected dominant error fields• Commissioned all magnetics, profile diagnostics• Operated NB systems with up to 12MW NBI available• Discovery: New 2nd NBI suppresses Alfvénic
instabilities– APS invited, PRL 2017
• Implemented techniques for controlled plasma shut
down, disruption detection / new tools for mitigation

mid January
early Marchearly April
early April

Several papers on NSTX-U results published, in-press, or
submitted• PF1A upper divertor coil failed, ending the run
14

NSTX-U Recovery will restore NSTX-U as a
criFcal facility for fusion science research• InvesFgaFons following the coil failure idenFﬁed addiFonalissues with the design of the device• The “Recovery Project” scope was deﬁned by an extensive
Extent of CondiFon analysis during 2017, resulFng in
numerous changes to improve the reliability of the device,including:• New inner-PF coils with improved design and manufacture• New plasma facing component Fles
• Improved vacuum boundary• These technical acFviFes in parallel with a complete overhaulof PPPLs engineering and quality procedures• Aiming for project baselining in the fall of this year, withresumpFon of operaFons in 202015

Outline• Why study “spherical tori/tokamaks” (STs)à And why use NSTX-U for that mission.

• Brief Discussion of NSTX-U OperaFons andRecovery• Key ScienFﬁc Deliverable of the NSTX-U Program• NSTX-U in the Larger ST/Fusion Context
16

NSTX-U will address cri.cal ques.ons related toperformance of the core plasma
Key Questions:• How will confinement and stability
scale at fusion relevant plasma
parameters?

Core• Can the ST plasma be sustained w/oa transformer?

17

NSTX-U will access and measure novel regimes ofturbulent transport at lower aspect ra.o & high beta

• Results from non-linear
gyrokineFc codes (GYRO)

Normalized transport rateχe (ρs2cs/a)

Guttenfelder, PRL• Experimental Result:Conﬁnement improves as (2011), PoP (2012)10collisionality is decreased• Energy conﬁnement scaling1from “micro-tearing” modes
simulated to be similar to
NSTX results

NSTX experiment

NSTX-U
regimesimulation

0.1

10
1
0.1
• Other microturbulent modes
Normalized collision frequency(KBMs) may become
dominant at low
Key diﬀerences between high- and low-A:
collisionality?
Low-A: electromagneFc electron turbulent

NSTX-U will deploy advanced
turbulence diagnosFcs tounderstand transporttransport is dominant channel, ions neoclassicalConven.onal-A: ion transport dominant,electrons and ions transport both dominated byelectrostaFc turbulence

18STs have the Poten.al For High Self-Driven
Currents Due to Inherently Higher Stability• Bootstrap Current: Current
internally generated by the 888
77
plasma pressure itself.
666
1/ 2

f BS ~ A⎛ βN ⎞⎜⎟ qcyl⎝ li ⎠

A = device aspect ratio
li = plasma internal inductance(Low value è broad current profile)

betaN
betaN

Plasma-driven current fraction

β55N

BetaN vs.li - XP948
vs.li - beta
Gridlines
n BetaN= 1 no-walllimit line

βN/li 14 13 12 11

Unstable RWM
Stable/Controlled RWM
Active control
Passive control

444
33
222

MAST-U

4

Theoretical limit
w/o stabilization

11
000
0.0
0.0
0.0

10

4li: High βN value forhigher A tokamak

0.2
0.2
0.2

0.4
0.4
0.4

li

0.6
0.6
0.6

0.8
0.8
0.8

lili

S. Sabbagh et al., Nucl. Fusion 53, 104007 (2013)

ST plasmas can achieve very high values of βN/li,
with associated high non-inductive current

19

NSTX-U will be leading ST able to access, study, andexploit kine.c resis.ve wall mode (RWM) stabiliza.on

Conducting
plates3D coils

Stability Improvement at High βN/li
J. Berkery, et al., PoP 21 (2014) 156112

Resonant Field Amplification (G/G)

Conducting Plates Resultin the Slowly Growing“Resistive Wall” Branchof the External Kink

unstable
RWM

Normalized β / current profile peaking (βN / li )Stability results consistent with drii-kineFc
theory including mode-parFcle resonances
20

NSTX-U Will Enable Physics Assessment of HighCurrent 100% Non-InducFve Scenarios
1

fGW = 0.72nτeT 1020 m-3 s keV• ConvenFonal aspect
raFo H-mode scaling
implies non inducFve
currents up to ~0.9 MA• ST-scaling implied
signiﬁcantly higher
performance• Performance will besimilar to 100% noninducFve operaFons at
A=3.

100% NI NSTX-U TRANSP

0.1

BT = 1.0 T
PNBI = 10.2 MW
βN = 4 – 4.3
H98=1
0.01
0

0.5

Ip (MA)1

1.5

Adapted from Gerhardt et al., NF 52 (2012) 083020
21

NSTX-U Will Enable Physics Assessment of HighCurrent 100% Non-InducFve Scenarios
1

fGW = 0.72nτeT 1020 m-3 s keV• ConvenFonal aspect
raFo H-mode scaling
implies non inducFve
currents up to ~0.9 MA• ST-scaling implies
signiﬁcantly higher
performance• Performance will besimilar to 100% noninducFve operaFons at
A=3.

100% NI NSTX-U TRANSP

BT = 1.0 T
PNBI = 10.2 MW
βN = 4 – 4.3
HST=1

0.1

BT = 1.0 T
PNBI = 10.2 MW
βN = 4 – 4.3
H98=1
0.01
0

0.5

Ip (MA)1

1.5

Adapted from Gerhardt et al., NF 52 (2012) 083020
22

NSTX-U Will Enable Physics Assessment of HighCurrent 100% Non-InducFve Scenarios
1

fGW = 0.72nτeT 1020 m-3 s keV• ConvenFonal aspect
raFo H-mode scaling
implies non inducFve
currents up to ~0.9 MA• ST-scaling implies
signiﬁcantly higher
performance• Performance will besimilar to 100% noninducFve operaFons at
A=3.

100% NI NSTX-U TRANSP

BT = 1.0 T
PNBI = 10.2 MW
βN = 4 – 4.3
HST=1
10.2à
15.6 MW

0.1

10.2à 15.6 MW
BT = 1.0 T
PNBI = 10.2 MW
βN = 4 – 4.3
H98=1
0.01
0

0.5

Ip (MA)1

1.5

Adapted from Gerhardt et al., NF 52 (2012) 083020
23

NSTX-U Will Enable Physics Assessment of HighCurrent 100% Non-InducFve Scenarios
1nτeT 1020 m-3 s keV, core averaged values• ConvenFonal aspect
raFo H-mode scaling
implies non inducFve
currents up to ~0.9 MA• ST-scaling implies
signiﬁcantly higher
performance• Performance will besimilar to steady state
operaFons in otherdevices.

100% NI NSTX-U TRANSP

JT-60U,
100% NI
DIII-D,
100% NI

W7-X
0.1

0.01
0

0.5

Ip (MA)1

1.5

Adapted from Gerhardt et al., NF 52 (2012) 083020
24

NSTX-U is excellent testbed for studying α-par.cle
physics applicable to burning plasmas, ITER
6-heatedSpherical
Tokamaks

42-heatedconventional aspectratio tokamaks(ITER, DEMO)

0
0.0

0.1

0.2

0.3

0.4

0.5

& fast/& total~ (EP-instability drive)/(EP-instability damping)25

NSTX-U is excellent testbed for studying α-par.cle
physics applicable to burning plasmas, ITER
• NSTX-U: Fast-ion dynamic range spanning ST
and conventional A burning plasma regime

64α-heated
Spherical
Tokamaks

2 α-heatedconventional aspectratio tokamaks(ITER, DEMO)– BT 2× NSTX à stabilize modes (Vfast < VAlfven )– Tangential 2nd NBI à flexible fast-ion distribution§ Vary pitch angle, pressure profile à Assess modestability as a function of ion phase space

Beam-heated
NSTX-U

Conventional
Beam-heated tokamaks
0
0.2
0.3
0.0
0.1
0.4

0.5

β fast/β total~ (EP-instability drive)/(EP-instability damping)

26

NSTX-U is excellent testbed for studying α-par.cle
physics applicable to burning plasmas, ITER
• NSTX-U: Fast-ion dynamic range spanning ST
and conventional A burning plasma regime

64α-heated
Spherical
Tokamaks

2 α-heatedconventional aspectratio tokamaks(ITER, DEMO)– BT 2× NSTX à stabilize modes (Vfast < VAlfven )– Tangential 2nd NBI à flexible fast-ion distribution§ Vary pitch angle, pressure profile à Assess modestability as a function of ion phase space

Beam-heated
NSTX-U

Conventional
Beam-heated tokamaks
0
0.2
0.3
0.0
0.1
0.4

0.5

β fast/β total

E.D. Fredrickson, et al., Phys. Rev. Lett. 18
265001 (2017)~ (EP-instability drive)/(EP-instability damping)

NSTX-U will:à Assess and utilize phase-spaceengineering to optimize the scenario
performanceà Extend predictive capabilities forenergetic particle dynamics in burning
plasma regime

27

Op.miza.on of the edge pedestal is cri.cal for
Fusion Performance
Pressure (kPa)

Pedestal is a narrow region of steep
gradients immediately inside the separatrix
L-modeH-modej||Radius (m)Key Questions:• What physics sets the height andwidth of the pedestal?
• How can we control the pedestal?
28

NSTX-U Oﬀers Exci.ng Opportuni.es For Op.miza.onof the Edge Pedestal

2: Higher lithium deposition
à Broader profiles & higher
confinement

A. Diallo et al., Nucl. Fusion 53 (2013) 093026
0.16

NSTX−ELMy
NSTX −fit (0.4 ± 0.1)(βped)(1.05 ± 0.2)
θ

0.14

ped 0.8
NSTX − BCP Theory 0.2(βθ )0.12Pedestal width [ψn]1: Pedestal Widths is Broader thanPredicted by Standard Theory(EPED) for Conventional-Aà Challenge to theory
à Opportunity for optimization –improved pedestal stability

0.1

0.08
0.06
0.04

C-Mod theory
DIII-D theory
ped 0.5
0.07(β )D3D & MAST − Theory 0.07(

C−Mod − Theory 0.08(
0.08(βped
)0.5
θ

0.02

θ

0
0.1

0.2

0.3

0.4

0.5

(βped
)1/2
θ

0.6

0.7

0.8

0.9

29

NSTX-U Oﬀers Exci.ng Opportuni.es For Op.miza.onof the Edge Pedestal
1: Pedestal Widths is Broader thanPredicted by Standard Theory(EPED) for Conventional-Aà Challenge to theory
à Opportunity for optimization–
improved pedestal stability
2: Lithium conditioning of PFCs
à Broader profiles à Eliminationof ELMs
à Higher confinement

R. Maingi et al., Nucl. Fusion 52 (2012) 083001

Without Li
With Li

H98~1

NSTX-U will have world leading
capabiliFes to diagnose, study, andcontrol the pedestal parameters.
D.P. Boyle, et al., J. Nucl. Mater. 438 (2013) S979 (PhD thesis work)30

Mi.ga.on / control of edge heat and par.cle
exhaust is cri.cal issue for magne.c fusion
Power exhaust width outside main
plasma can be very narrow (few mm)
Key Questions:• How does the scrape-off-layer
width (λq) scale with plasma
parameters?

λq• How can the severe heat loads
reaching the divertor be mitigated?
Edge power and particles are
purposely guided away from plasmacore using a magnetic “divertor”
31

NSTX-U will play important role in understanding how
power exhaust width extrapolates to future devices
XGC1 simulations predict
turbulence will widenedge heat flux in ITER
C.S. Chang et al 2017 Nucl. Fusion 57 116023

2MA NSTX-U
Range of NSTX-U
λq projectionsC-Mod

XGC1 studies of NSTX-U
initiated and ongoing
Preliminary predictions
show λq ~ 3 mm
(unpublished)∝ Plasma current

32

NSTX-U aims to explore liquid metal PFCs
• Need for high heat ﬂux exhaustmoFvates long term research
direcFons for NSTX-U• High-Z Fles with pre-ﬁlled Li• Flowing liquid lithium divertor withexternal feed into reservoir

High heat ﬂuxes in NSTX-U (>20 MW/
m2 unmiFgated in standard divertors)
make NSTX-U an excellent facility forthis research• Part of a liquid metal PFC program at
PPPL• LTX-β (staFc liquid Li pools)• EAST collaboraFon (coaFngs, slow ﬂow
liquid lithium limiter)

LTX-β

See talk by Jaworski tomorrow for more details on liquid-metal PFCs

33

Outline• Why study “spherical tori/tokamaks” (STs)à And why use NSTX-U for that mission.

• Brief Discussion of NSTX-U OperaFons andRecovery• Key ScienFﬁc Deliverable of the NSTX-U Program• NSTX-U in the Larger ST/Fusion Context
34

NSTX-U Research is Highly Collabora.ve,
Facility Supports Wide range of Users
362 data users
40 international

54 collaborating institutions
32 domestic, 22 international

29 graduate students
25 post-doctoral researchers

FY2016 numbers

35CollaboraFons Support NSTX-U Mission While
Advancing Fusion CapabiliFes at Other FaciliFes• Lithium use and related technologies have spread: EAST
dropper, granule injector on EAST and DIII-D, collaboraFons
with AUG and W7-X.
• Solenoid free start-up technologies developed on NSTX
being used on QUEST.
• Scenario & equilibrium reconstrucFon development on
MAST-U• Proﬁle and divertor control development on DIII-D
Transient CHI on QUEST

MAST-U and NSTX-U Shape Comparison à
RealFme Control of NB
Voltage on DIII-D

B PowderB powder dropping in AUGà
36

Private Companies are Pursuing ST Research
General Fusion - Canada

Tokamak Energy - UK

Future Plans: Compress plasmas via
converging liquid metal liner.

ST-40: 2 MA, 3 T, R0=0.4 m Cu magnets
Merging compression and reconnecFon
formaFon scheme
Future Plans: HTS development and devices

ST is the most stable compact plasma
conﬁguraFon (FRCs, spheromaks, STs…)

PI3

These faciliFes focus on developing technology
using ST plavorm – synergy with DoE research

ST-40
37

NSTX-U Addresses Key Science Issues from
Compact Fusion System• How will turbulence and transport scale to the low-collisionality
regime?
àKey research enabled by increased magneFc ﬁeld of the Upgrade• Can the plasma current be sustained 100% non-inducFvely at
fusion relevant β and with high conﬁnement?

àKey research enabled by increased magneFc ﬁeld and 2nd neutral beamàNearby conducFng plates facilitate key low-A RWM stability research• How do energeFc parFcle stability and conﬁnement properFes
trend into the burning plasma regime?
à2nd neutral beam and upgraded of TF allows fast-ion phase space to becomprehensively exploredàCan span burning plasma regimes at both low- and convenFonal A• How can the power exhaust be handled in compact fusion
systems?
àExploring innovaFons in ﬂared divertors and lithium systems forhandling large divertor heat ﬂuxes.

38Summary• NSTX-U contributes to optimization of the tokamak
concept though optimization of aspect ratioà Potentially transformational behavior in STs whenapproaching fusion relevant plasma parameters• NSTX-U is a Key Component of the PPPL
Programmatic Strategy Towards Fusion Powerà Supports our vision for innovation in compact high-β
configurations with lower capital costs, materials innovation• NSTX-U has a unique role within the world ST
programà Unique focus on core physics & impact of Li on plasma
behavior• NSTX-U will provide answers to critical fusion
science questions early in its operation.
à Exciting science from the commissioning campaign

39