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Title
An overview of recent physics results from NSTX
Permalink
https://escholarship.org/uc/item/28s4q3dp
Journal
Nuclear Fusion, 55(10)
ISSN
0029-5515
Authors
Kaye, SM
Abrams, T
Ahn, JW
et al.
Publication Date
2015-03-27
DOI
10.1088/0029-5515/55/10/104002
Copyright Information
This work is made available under the terms of a Creative Commons Attribution License,
availalbe at https://creativecommons.org/licenses/by/4.0/
Peer reviewed
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University of California
| International Atomic Energy Agency Nuclear Fusion
Nucl. Fusion 55 (2015) 104002 (18pp) doi:10.1088/0029-5515/55/10/104002
An overview of recent physics results from
NSTX
S.M. Kaye
1
, T. Abrams
1
, J.-W. Ahn
2
, J.P. Allain
3
, R. Andre
1
, D. Andruczyk
3
,
R. Barchfeld
4
, D. Battaglia
1
, A. Bhattacharjee
1
, F. Bedoya
3
, R.E. Bell
1
,E.Belova
1
,
J. Berkery
5
, L. Berry
2
, N. Bertelli
1
, P. Beiersdorfer
6
, J. Bialek
5
, R. Bilato
7
,
J. Boedo
8
, P. Bonoli
9
, A. Boozer
5
, A. Bortolon
10
, M.D. Boyer
1
,D.Boyle
1
,
D. Brennan
11
, J. Breslau
1
, J. Brooks
12
, R. Buttery
13
, A. Capece
1
, J. Canik
2
,
C.S. Chang
1
,N.Crocker
14
, D. Darrow
1
,W.Davis
1
, L. Delgado-Aparicio
1
, A. Diallo
1
,
D. D’Ippolito
15
, C. Domier
4
, F. Ebrahimi
11
, S. Ethier
1
,T.Evans
13
, N. Ferraro
13
,
J. Ferron
13
, M. Finkenthal
16
, R. Fonck
17
, E. Fredrickson
1
, G.Y. Fu
1
, D. Gates
1
,
S. Gerhardt
1
, A. Glasser
18
, N. Gorelenkov
1
, M. Gorelenkova
1
, I. Goumiri
11
, T. Gray
2
,
D. Green
2
, W. Guttenfelder
1
, R. Harvey
19
, A. Hassanein
12
, W. Heidbrink
20
,
Y. Hirooka
21
, E.B. Hooper
6
, J. Hosea
1
, D. Humphreys
13
, E.F. Jaeger
22
, T. Jarboe
18
,
S. Jardin
1
, M.A. Jaworski
1
, R. Kaita
1
, C. Kessel
1
,K.Kim
1
,B.Koel
11
, E. Kolemen
1
,
G. Kramer
1
,S.Ku
1
, S. Kubota
14
, R.J. LaHaye
13
,L.Lao
13
, B.P. LeBlanc
1
,
F. Levinton
23
,D.Liu
20
, J. Lore
11
, M. Lucia
3
, N. Luhmann Jr
14
, R. Maingi
1
,
R. Majeski
1
, D. Mansfield
1
, R. Maqueda
23
, G. McKee
17
, S. Medley
1
, E. Meier
6
,
J. Menard
1
, D. Mueller
1
, T. Munsat
24
, C. Muscatello
4
, J. Myra
15
, B. Nelson
18
,
J. Nichols
1
,M.Ono
1
, T. Osborne
13
, J.-K. Park
1
, W. Peebles
14
, R. Perkins
1
,
C. Phillips
1
, M. Podesta
1
,F.Poli
1
, R. Raman
18
,Y.Ren
1
, J. Roszell
11
,C.Rowley
11
,
D. Russell
15
, D. Ruzic
3
,P.Ryan
2
, S.A. Sabbagh
5
, E. Schuster
25
, F. Scotti
6
,
Y. Sechrest
24
, K. Shaing
17
, T. Sizyuk
12
, V. Sizyuk
3
, C. Skinner
1
, D. Smith
17
,
P. Snyder
13
, W. Solomon
1
, C. Sovenic
17
, V. Soukhanovskii
6
, E. Startsev
1
,
D. Stotler
1
, B. Stratton
1
, D. Stutman
16
, C. Taylor
12
, G. Taylor
1
, K. Tritz
16
,
M. Walker
13
,W.Wang
1
,Z.Wang
1
, R. White
1
, J.R. Wilson
1
, B. Wirth
10
, J. Wright
9
,
X. Yuan
1
,H.Yuh
23
, L. Zakharov
1
and S.J. Zweben
1
1
Princeton Plasma Physics Laboratory, Princeton University, Princeton, NJ 08543, USA
2
Oak Ridge National Laboratory, Oak Ridge, TN 37831, USA
3
University of Illinois at Urbana-Champaign, Urbana, IL, USA
4
University of California at Davis, Davis, CA, USA
5
Columbia University, New York, NY, USA
6
Lawrence Livermore National Laboratory, Livermore, CA, USA
7
IPP, Garching, Germany
8
University of California at San Diego, San Diego, CA, USA
9
Massachusetts Institute of Technology, Cambridge, MA, USA
10
University of Tennessee, Knoxville, TN, USA
11
Princeton University, Princeton, NJ, USA
12
Purdue University, W. Lafayette, IN, USA
13
General Atomics, San Diego, CA, USA
14
University of California at Los Angeles, Los Angeles, CA, USA
15
Lodestar Research Corporation, Boulder, CO, USA
16
Johns Hopkins University, Baltimore, MD, USA
17
University of Wisconsin, Madison, WI, USA
18
University of Washington, Seattle, WA, USA
19
CompX, Del Mar, CA, USA
20
University of California at Irvine, Irvine, CA, USA
21
National Institute for Fusion Science, Oroshi, Toki, Gifu, Japan
22
XCEL, Oak Ridge, TN, USA
23
Nova Photonics, Princeton, NJ, USA
24
University of Colorado, Boulder, CO, USA
25
Lehigh University, Bethlehem, PA, USA
E-mail: skaye@pppl.gov
0029-5515/15/104002+18$33.00 1 Not subject to copyright in the USA/Contribution of the Department of Energy Printed in the UK
Nucl. Fusion 55 (2015) 104002 S.M. Kaye et al
Received 13 November 2014, revised 18 December 2014
Accepted for publication 7 January 2015
Published 27 March 2015
Abstract
The National Spherical Torus Experiment (NSTX) is currently being upgraded to operate at twice the toroidal field and plasma
current (up to 1 T and 2 MA), with a second, more tangentially aimed neutral beam (NB) for current and rotation control,
allowing for pulse lengths up to 5 s. Recent NSTX physics analyses have addressed topics that will allow NSTX-Upgrade to
achieve the research goals critical to a Fusion Nuclear Science Facility. These include producing stable, 100% non-inductive
operation in high-performance plasmas, assessing plasma–material interface (PMI) solutions to handle the high heat loads
expected in the next-step devices and exploring the unique spherical torus (ST) parameter regimes to advance predictive
capability. Non-inductive operation and current profile control in NSTX-U will be facilitated by co-axial helicity injection
(CHI) as well as radio frequency (RF) and NB heating. CHI studies using NIMROD indicate that the reconnection process
is consistent with the 2D Sweet–Parker theory. Full-wave AORSA simulations show that RF power losses in the scrape-off
layer (SOL) increase significantly for both NSTX and NSTX-U when the launched waves propagate in the SOL. Toroidal
Alfv
´
en eigenmode avalanches and higher frequency Alfv
´
en eigenmodes can affect NB-driven current through energy loss and
redistribution of fast ions. The inclusion of rotation and kinetic resonances, which depend on collisionality, is necessary for
predicting experimental stability thresholds of fast growing ideal wall and resistive wall modes. Neutral beams and neoclassical
toroidal viscosity generated from applied 3D fields can be used as actuators to produce rotation profiles optimized for global
stability. DEGAS-2 has been used to study the dependence of gas penetration on SOL temperatures and densities for the MGI
system being implemented on the Upgrade for disruption mitigation. PMI studies have focused on the effect of ELMs and 3D
fields on plasma detachment and heat flux handling. Simulations indicate that snowflake and impurity seeded radiative divertors
are candidates for heat flux mitigation in NSTX-U. Studies of lithium evaporation on graphite surfaces indicate that lithium
increases oxygen surface concentrations on graphite, and deuterium–oxygen affinity, which increases deuterium pumping and
reduces recycling. In situ and test-stand experiments of lithiated graphite and molybdenum indicate temperature-enhanced
sputtering, although that test-stand studies also show the potential for heat flux reduction through lithium vapour shielding.
Non-linear gyro kinetic simulations have indicated that ion transport can be enhanced by a shear-flow instability, and that
non-local effects are necessary to explain the observed rapid changes in plasma turbulence. Predictive simulations have shown
agreement between a microtearing-based reduced transport model and the measured electron temperatures in a microtearing
unstable regime. Two Alfv
´
en eigenmode-driven fast ion transport models have been developed and successfully benchmarked
against NSTX data. Upgrade construction is moving on schedule with initial physics research operation of NSTX-U planned
for mid-2015.
Keywords: NSTX, spherical torus, overview
(Some figures may appear in colour only in the online journal)
1. Introduction
Recent analysis of data from the National Spherical Torus
Experiment (NSTX) has focused on topics critical to the
development of the research plan and achievement of physics
goals for the NSTX-Upgrade. NSTX (R/a = 0.85/0.65 m,
κ = 1.8–2.4, δ = 0.3–0.8, I
p
up to 1.5 MA, B
T
up to
0.55 T) concluded operation in October 2010 in order to
begin Upgrade construction activities. NSTX-U [1] has three
primary research goals:
1. To advance the spherical torus (ST) concept for a fusion
nuclear science facility (FNSF) [2]. Achieving this goal
requires demonstrating 100% non-inductive sustainment
at a performance level that extrapolates to 1MWm
−2
neutron wall loading in FNSF, as well as developing non-
inductive start-up and ramp-up techniques for an FNSF
with a small, or no, solenoid. The latter is a particularly
unique requirement for an ST-based FNSF [3].
2. To develop solutions for the plasma–material interface
(PMI). High-flux expansion snowflake or X-divertor,
coupled with radiative detachment, will be employed
for mitigating high heat fluxes (q
peak,div
40 MW m
−2
,
P
heat
/S 0.5MWm
−2
, up to five times greater than
in NSTX), and high-Z and liquid lithium plasma-facing
components (PFC) will be assessed as PMI solutions.
3. To explore unique ST parameter regimes to advance
predictive capability for ITER and beyond. In order
to address this goal, NSTX-U will access reduced
collisionality (up to a factor of ten lower than that in
NSTX) and high-β (β
n
6) with an ability to vary the q-
and rotation profiles for enhanced stability, confinement
and non-inductive current drive. Models for thermal and
fast ion transport will be developed and/or tested.
The increased capabilities of NSTX-U (R/a =
0.95/0.55 m, κ 2.8) will facilitate the research necessary to
achieve these goals. The major capabilities include a factor of
∼two increase in both plasma current I
p
and toroidal magnetic
field B
T
to 2 MA and 1 T respectively, and an additional off-
axis neutral beam that will double the available nominal input
beam power to 12 MW. The first two enhancements will allow
for high-β operation at almost an order-of-magnitude lower
collisionality [4], and the second neutral beam, coupled with
up to 6 MW of high harmonic fast wave (HHFW) heating and
current drive, will provide means to vary the q- and rotation
2
Nucl. Fusion 55 (2015) 104002 S.M. Kaye et al
profile, and it will provide significant non-inductive current
drive directly or indirectly. This will enable NSTX-U to
sustain stable, high-performance discharges in near steady-
state conditions for pulse lengths up to 5 s, which is a duration
of many current diffusion times. Advanced control algorithms,
using the above capabilities as well as applied n = 1
to 3 non-axisymmetric magnetic perturbations (MPs) from
a midplane, and eventually, possibly an off-midplane, coil
system will be utilized for stable operation. A cryo-pump is
being considered for density control. A disruption prediction,
avoidance and mitigation (PAM) system will be deployed using
a large number of measurements and models for prediction and
avoidance [5–7] and a massive gas injection (MGI) system
at different poloidal locations to optimize gas penetration
for reducing potential wall damage due to disruptive forces.
Lithium evaporation and boronization will be used as the
primary wall conditioning techniques, and off-line research
centred on liquid lithium on metal substrates will support a
planned phasing in of high-Z PFCs.
The capabilities of NSTX-U complement those of the
companion ST, MAST-U [8], which plans to commence
operation shortly after the start of NSTX-U. The elements
of recent research on NSTX-U, both experimental analysis
and theory that address the topics mentioned above, will be
reported in this paper.
2. Advance the ST concept for a Fusion Nuclear
Science Facility
This section will focus on two main issues, the non-inductive
current drive research that will allow for achievement of
the 100% non-inductive goal in the three phases of the
discharge (start-up, ramp-up and sustainment), and the ability
to maintain magnetohydrodynamic (MHD) stable plasmas for
long duration.
2.1. Non-inductive start-up and ramp-up
NSTX-U is striving for fully non-inductive operation to
establish the physics basis of an ST-FNSF with small or no
solenoid, and the starting point is plasma initiation and current
ramp-up. An envisioned strategy for non-inductive operation
in NSTX-U is shown in figure 1. NSTX-U will initiate the
plasma using co-axial helicity injection (CHI), as has been
done in NSTX [9]. Plasma guns [10] will be tested on NSTX-U
for start-up once development on smaller devices is complete.
Once the plasma is initiated, electron cyclotron heating (ECH)
is being considered for heating the plasma from an initial
temperature of 10 s of eV to up to several hundred eV, along
with HHFW to heat the plasma further to the 1–3 keV level (red
phase). At this point, HHFW and/or neutral beam injection
(NBI) current drive would be used to ramp the plasma current
up to full current (blue phase), at which time neutral beam
current drive and bootstrap current would be utilized to sustain
the full current non-inductively (black phase).
CHI has been successfully used for plasma formation in
NSTX with currents up to 200 kA and especially with coupling
to inductive ramp-up [9]. Axisymmetric Tokamak Simulation
Code (TSC) [11] simulations, reported previously [12],
have been validated successfully against NSTX discharges,
Figure 1. Strategy for fully non-inductive operation in NSTX-U.
and these simulations are the basis for start-up scenario
development on NSTX-U, where CHI is projected to be able
to provide up to 400 kA of start-up current. An example of the
TSC-produced flux evolution in an NSTX plasma is shown in
figure 2. For this case, a 5 ms voltage pulse is applied across
the injection electrodes, providing sufficient current to allow
the discharge to fill the vessel. The voltage is rapidly reduced
to zero at t = 5 ms (first panel in figure 2). The closed flux
surfaces form in this axisymmetric simulation as a result of
generation of a strong toroidal loop voltage that drives the
toroidal current.
The underlying physics of CHI start-up have been studied
with resistive MHD simulations using the NIMROD code [13]
in 2D in order to improve the flux surface closure and
current drive using this technique [14, 15]. In one set of
simulations, time-varying boundary conditions and poloidal
field coil currents emulating experimental conditions were
used, and closed flux surfaces were produced [14]. A more
simplified model, with constant poloidal field currents and
time varying injector currents, was used to study the minimum
conditions for flux closure [15]. In these simulations, the
injector voltage is adjusted so that the J × B force overcomes
the field line tension and open field lines fill the vessel. Flux
closure occurs under the right conditions approximately 0.5 ms
after the injector voltage is turned off, as in the experiment.
These simulations indicate that the magnetic diffusivity
strongly controls flux closure, with no flux closure at high
diffusivity corresponding to temperatures lower (T
e
1 eV)
than those measured in the experiment. As the diffusivity
decreases with T
e
increasing towards the experimental values
(10–25 eV), flux closure occurs with increasing volume of
closed flux. Field line tracing was used to confirm the
formation of the X-point and flux closure, as is shown in
figure 3 for two temperatures, T
e
= 14 and 24 eV. Additional
scans indicate other dependences controlling flux closure. If
the injection voltage is slowly reduced to zero or the injector
flux footprints are too far apart, closure is inhibited. The
first is due to smaller generated toroidal loop voltage, and the
second due to the longer time scales required for the oppositely
directed flux to come together and form an X-point.
3
Nucl. Fusion 55 (2015) 104002 S.M. Kaye et al
Figure 2. Time evolution of flux surface closure with CHI in an NSTX plasma.
Figure 3. Poincar
´
e plots soon after flux closure for two
experimentally relevant electron temperatures (a)14eVand
(b)24eV[15]. (Copyright 2013, American Institute of Physics.)
The simulations have shown that during X-point
formation, the current density is localized to an elongated
current sheet (figure 4) whose width scales as η
1/2
, where η is
the magnetic diffusivity. This, along with the computed strong
inflow and outflow characteristics, suggests that the X-point
formation may be a Sweet–Parker-type reconnection.
Once the plasma is initiated, the core electron temperature
has to be increased in order for HHFW or NBI to couple to the
NSTX-U plasma for continued ramp-up of the plasma current.
Figure 4. Toroidal current in the reconnection region. [14, 15].
(Copyright 2013, American Institute of Physics.)
In order to do this, and as part of a possible future upgrade to
the NSTX-U facility, a 28 GHz O-mode ECH system that is
capable of injecting 1 MW for pulse lengths of up to several
seconds can be utilized. TSC simulations predict that the
central electron temperature can increase from 10 to 100 eV
in 20 ms when 0.6 MW of ECH power is coupled to a CHI
plasma. This increase in temperature will reduce significantly
the plasma current decay rate of CHI plasmas and allow for
coupling to medium and high harmonic fast wave heating
and NBI.
The level of heating and current drive that HHFW can
provide is critically dependent on how much power is lost
along open field lines in the SOL. Recent experimental results
have shown that up to 60% of the coupled power can be
lost when edge densities are high enough that fast waves can
propagate close to the launcher [16]. Thus, it is important to
understand the physics of RF deposition and especially this
edge loss mechanism in order to optimize discharge scenarios
that lead to RF loss minimization. Understanding the physics
can aid in the projection to ICRF efficiency and deposition
in ITER.
4
![Figure 6. Fraction of power lost to the SOL as a function of density in front of the antenna for nφ = −21 (solid red) and nφ = −12 (dashed blue) as calculated by AORSA. The vertical lines indicate the cutoff density for the two respective nφ . (Reproduced from [18] courtesy of IAEA.)](/figures/figure-6-fraction-of-power-lost-to-the-sol-as-a-function-of-2bgcjxt1.webp)
![Figure 19. Linear growth rates (top) and real frequencies (bottom) of microturbulent modes in an unlithiated discharge (black) and a lithiated one (red). The black and red vertical lines indicate the location of the top of the density pedestal for the unlithiated and lithiated cases, respectively. (Reproduced from [59] courtesy of IAEA.)](/figures/figure-19-linear-growth-rates-top-and-real-frequencies-31xjxbbd.webp)
![Figure 9. (Top) Spectrogram of envelope of high-frequency CAE fluctuation amplitude, overlaid with kink mode frequencies. (Bottom) Relative phase between the n = 1 kink and high-frequency CAE envelope [23]. (Copyright 2013, American Institute of Physics.)](/figures/figure-9-top-spectrogram-of-envelope-of-high-frequency-cae-j0ibazuh.webp)