High performance plasmas on the National Spherical Torus Experiment

Abstract

The National Spherical Torus Experiment has produced toroidal plasmas at low aspect ratio (A = R/a = 0.86 m/0.68 m ∼ 1.3, where R is the major radius and a is the minor radius of the torus) with plasma currents of 1.4 MA. The rapid development of the machine has led to very exciting physics results during the first full year of physics operation. Pulse lengths in excess of 0.5 s have been obtained with inductive current drive. Up to 4 MW of high harmonic fast wave (HHFW) heating power has been applied with 6 MW planned. Using only 2 MW of HHFW heating power clear evidence of electron heating is seen with HHFW, as observed by the multi point Thomson scattering diagnostic. A noninductive current drive concept known as coaxial helicity injection (CHI) has driven 260 kA of toroidal current. A neutral beam heating power of 5 MW has been injected. Plasmas with β₁ (= 2μ₀ < p > /B² = a measure of magnetic confinement efficiency) of 22% have been achieved, as calculated using the EFIT equilibrium reconstruction code. β limiting phenomena have been observed, and the maximum β₁ scales with I_p/B_1. High frequency (> MHz) magnetic fluctuations have been observed. H-mode plasmas are observed with confinement times of > 100 s. Beam heated plasmas show energy confinement times in excess of those predicted by empirical scaling expressions. Ion temperatures in excess of 2.0 keV have been measured, and the power balance suggests that the power loss from the ions to the electrons may exceed the calculated classical input power to the ions.

Full text

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PREPARED FOR THE U.S. DEPARTMENT OF ENERGY,
UNDER CONTRACT DE-AC02-76CH03073
PRINCETON PLASMA PHYSICS LABORATORY
PRINCETON UNIVERSITY, PRINCETON, NEW JERSEY
PPPL-3590
UC-70
PPPL-3590
High Performance Plasmas
on the National Spherical Torus Experiment
by
D.A. Gates, M.G. Bell, R.E. Bell, J. Bialek, T. Bigelow, M. Bitter, P. Bonoli,
D. Darrow, P. Efthimion, J. Ferron, E. Fredrickson, L. Grisham, J. Hosea,
D. Johnson, R. Kaita, S. Kaye, S. Kubota, H. Kugel, B. LeBlanc, R. Maingi,
J. Manickam, T.K. Mau, R.J. Maqueda, E. Mazzucato, J. Menard, D. Mueller,
B. Nelson, N. Nishino, M. Ono, F. Paoletti, S. Paul, Y-K.M. Peng, C.K. Phillips,
R. Raman, P. Ryan, S.A. Sabbagh, M. Schaffer, C.H. Skinner, D. Stutman,
D. Swain, E. Synakowski, Y. Takase, J. Wilgen, J.R. Wilson, W. Zhu, S. Zweben,
A. Bers, M. Carter, B. Deng, C. Domier, E. Doyle, M. Finkenthal, K. Hill, T. Jarboe,
S. Jardin, H. Ji, L. Lao, K.C. Lee, N. Luhmann, R. Majeski, H. Park, T. Peebles,
R.I. Pinsker, G. Porter, A. Ram, M. Rensink, T. Rognlien, D. Stotler, B. Stratton,
G. Taylor, W. Wampler, G.A. Wurden, X.Q. Xu, L. Zeng, and the NSTX Team
July 2001

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HIGH PERFORMANCE PLASMAS ON THE NATIONAL SPHERICAL
TORUS EXPERIMENT
D. A. Gatesa, M. G. Bella, R. E. Bella, J. Bialekb, T. Bigelowc, M. Bittera, P. Bonolid, D. Darrowa,
P. Efthimiona, J. Ferrone, E. Fredricksona, L. Grishama, J. Hoseaa, D. Johnsona, R. Kaitaa, S.
Kayea, S. Kubotaf, H. Kugela, B. LeBlanca, R. Maingic, J. Manickama, T. K. Maug, R. J.
Maquedah, E. Mazzucatoa, J. Menarda, D. Muellera, B. Nelsoni, N. Nishinoj, M. Onoa, F.
Paolettib, S. Paula , Y-K. M. Pengc , C. K. Phillipsa, R. Ramani, P. Ryanc, S. A. Sabbaghb, M.
Schaffere, C. H. Skinnera, D. Stutmank, D. Swainc, E. Synakowskia, Y. Takasel, J. Wilgenc, J.R.
Wilsona, W. Zhub, S. Zwebena, A. Bersd, M. Carterc, B. Dengm, C. Domierm, E. Doylef, M.
Finkenthalk, K. Hilla, T. Jarboei, S. Jardina, H. Jia, L. Laoe, K. C. Leeo, N. Luhmannm, R.
Majeskia, H. Parka, T. Peeblesf, R. I. Pinskere, G. Portern, A. Ramd, M. Rensinkn, T. Rognlienn,
D. Stotlera, B. Strattona, G. Taylora, W. Wamplero, G. A. Wurdenh, X. Q. Xun, L. Zengf, and
the NSTX Team
a Princeton Plasma Physics Laboratory, Princeton University, Princeton, N.J. 08543
b Dept. of Applied Physics, Columbia Univ., NYC, N.Y.
c Oak Ridge National Laboratory, Oak Ridge, Tenn.
d MIT, Cambridge, Mass.
e General Atomics, San Diego, Cal.
f UCLA, Los Angeles, Cal.
g UC-San Diego, San Diego, Cal.
h Los Alamos National Laboratory, Los Alamos, N.M.
i Univ. of Washington, Seattle, Wash.
j Hiroshima Univ., Hiroshima, Japan
k Johns Hopkins University, Baltimore, Md.
l Univ. of Tokyo, Tokyo, Japan
m UC Davis, Davis, Cal.
n Lawrence Livermore National Laboratory, Livermore, Cal.
o Sandia National Laboratory, Albuquerque, N.M.
Abstract
The National Spherical Torus Experiment has produced
toroidal plasmas at low aspect ratio (A = R/a =
0.86m/0.68m ~ 1.3, where R is the major radius and a is
the minor radius of the torus) with plasma currents of
1.4MA. The rapid development of the machine has led to
very exciting physics results during the first full year of
physics operation. Pulse lengths in excess of 0.5s have
been obtained with inductive current drive. Up to 4MW
of High Harmonic Fast Wave (HHFW) heating power has
been applied with 6MW planned. Using only 2MW of
HHFW heating power clear evidence of electron heating
is seen with HHFW, as observed by the multi point
Thomson scattering diagnostic. A non-inductive current
drive concept known as Coaxial Helicity Injection (CHI)
has driven 260kA of toroidal current. Neutral beam
heating power of 5MW has been injected. Plasmas with
βt (=2µ0<p>/B2 = a measure of magnetic confinement
efficiency) of 22% have been achieved, as calculated
using the EFIT equilibrium reconstruction code. β
limiting phenomena have been observed, and the
maximum βt scales with Ip/aBt. High frequency (>MHz)
magnetic fluctuations have been observed. H-mode
plasmas are observed with confinement times of >
100ms. Beam heated plasmas show energy confinement
times in excess of those predicted by empirical scaling
expressions. Ion temperatures in excess of 2.0keV have
been measured, and power balance suggests that the
power loss from the ions to the electrons may exceed the
calculated classical input power to the ions.
I.
INTRODUCTION
The National Spherical Torus Experiment (NSTX) [1]
is a midsize toroidal magnetic confinement device that is
designed to produce plasmas with a low aspect ratio. The
spherical torus has been predicted to have many attractive
features including high normalized pressure or βt (=
2µ0<p>/Bt
scale fluctuations [3] which are believed to degrade
confinement. The spherical torus has been the focus of
increased experimental activity following the
experimental demonstration of high β [4] with good
confinement [5] on the pioneering START [6] device.
An assembly drawing of NSTX is shown in Figure 1.
The extreme geometry is apparent from the drawing. The
column (in red) in the center of the device provides the
toroidal field and is capable of carrying a net vertical
current in excess of 2.5MA (corresponding to the
2) [2] and improved stability against small

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maximum toroidal field on axis of 0.6T). The major
design parameters of the NSTX device are shown in
Table 1.
Figure 1. Cross-section assembly drawing of NSTX
Table 1. Device parameters for NSTX
Parameter
Plasma current (Ip)
Toroidal field (Bt)
Major radius (R)
Minor radius (a)
Neutral beam power (PNBI)
Radio heating power (PRF)
Elongation (κ)
Pulse Length
Achieved
1.4 MA
0.6 T
86cm
68cm
5MW
4MW
2.4
0.5s
Planned
1.0 MA
0.6 T
86cm
68cm
5MW
6MW
2.2
~5 s
The primary benefit of low aspect ratio is the increased
field line length on the inboard (small major radius) side
of the device. This is beneficial since the curvature of the
field lines in this region is in the direction which is
stabilizing for several classes of plasma instabilities. This
is primarily a geometric effect.
Along with this benefit comes an important issue that
must be addressed: the small area on the interior of the
plasma. This geometric limitation creates difficulty for
current drive, which is achieved in conventional
tokamaks by a solenoid transformer, and makes the use of
super-conducting toroidal field coils hard to envision.
The limited area for this transformer on an ST creates a
constraint on the achievable transformer-driven pulse
length for an ST. Much of the research program for
NSTX is aimed at alleviating this constraint.
II.
CO-AXIAL HELICITY INJECTION
Coaxial Helicity Injection [7] is a current drive
technique that uses an externally applied DC voltage to
initiate a toroidal plasma current. This technique was first
used to effect on the HIT device [8]. Poloidal current is
drawn between the outer portion of the vacuum vessel
and the center stack, which are electrically isolated from
each other by two insulators, one at the top of the center
stack and one at the bottom. In order to encourage current
to flow along the desired path, the poloidal field is
maximized across the bottom insulator, where the JθxBt
force points up into the vessel, and minimized across the
top insulator. Current cannot flow purely in the poloidal
direction however, due to the applied toroidal field.
Current instead flows along the nearly helical resultant
field lines, causing the current to make several toroidal
transits before contacting the opposing electrode.
In this manner a poloidal current is used to create a
toroidal current. To date a toroidal current of 260kA has
been driven using only ~25kA of injected poloidal
current, as can be seen in Figure 2. This large
multiplication ratio is the result of the high ratio of the
toroidal field and the poloidal field in the lower insulator
region.
This technique has not yet been used in conjunction
with auxiliary heating due difficulties in obtaining
desirable equilibria. The difficulties are associated with
insulator design issues and poloidal field control in the
upper electrode.
Figure 2 Current and voltage traces from a CHI
discharge on NSTX
III. HIGH HARMONIC FAST WAVE HEATING
The NSTX high harmonic fast wave antenna consists of
12 straps which are powered by six independent RF
sources. The antenna is large, occupying approximately
1/3 of the circumference of the device at the outboard
mid-plane. The large antenna helps reduce peak power
density to values of ~2MW/m2 which has led to reliable
operation without the large density increases often
associated with high power RF heating. The low power
density also reduces antenna conditioning time.
A radio frequency wave heating scheme known as high
harmonic fast wave heating [9] has been used to

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effectively heat electrons on NSTX. Peak temperatures of
1.8keV have been obtained for parallel wave numbers
between 7m-1 and 14m-1. Shown in Figure 3 are 2 shots
one with RF heating and one without, showing significant
electron heating. To date, no parasitic ion heating caused
by coupling to the ion Bernstein wave has been observed,
which was an early theoretical concern for Ti(0) > 1keV.
The phase of each antenna strap can be independently
controlled in real time to vary the launched wave
spectrum between a slow (heating) and a fast (current
drive) phasing. Long range uses for this ability include
current and pressure profile control. Such control could
allow more reliable operation near β limits.
Figure 3 Two comparison discharges, one with 2MW of
RF heating one without, showing strong RF heating
IV.
NEUTRAL BEAM HEATED DISCHARGES
A. High plasma current
Another benefit of low aspect ratio is that the toroidal
plasma current can be much higher for a given applied
axial toroidal magnetic field. The maximum plasma
current attained to date is 1.4MA with a 4.0kG toroidal
field. An important part of this result was the small
inferred halo currents, which are induced py plasma
disruptions and which have been observed to cause
significant damage to conventional tokamaks at high
plasma current. The ability to support high plasma current
is important, since toroidal burning plasmas require a
minimum poloidal magnetic field in order to contain α-
particles from D-T fusion reactions.
B. Extended Pulse
Initial experiments aimed at extending the pulse of
NSTX discharges and have successfully extended the
total pulse length of 1MA discharges to > 0.5s. The
techniques used to date include neutral beam preheat
during the current ramp and raising the toroidal field to
control MHD activity. Evidence suggests that the pulse
length limit on NSTX is set by the central safety factor,
q(0), being of order unity.
Figure 4 Comparison of a a) plasma current b) stored
energy and c) neutral beam power for a 3kG ohmic
discharge (101522) and a 4.5kG discharge with neutral
beam preheat (103887)
C. High Normalized Pressure
Neutral beam heating with injected power reaching
5MW has been used on NSTX to raise the plasma
pressure to a substantial fraction of the predicted limit.
Figure 5 βt vs. IN for NSTX plasmas with strong neutral
beam heating
Plasmas with βt ~ 20% have been achieved on NSTX
Shown in Figure 5 is βt plotted as a function of the
normalized plasma current IN = Ip/aBt (MA/m·T). Ideal
(non-resistive) magneto-hydrodynamic (MHD) modes are
predicted to limit βt to a value that scales linearly with IN
[10]. The observed nearly linear scaling of βt with IN
indicates that the observed limit may be set by ideal
MHD instabilities.

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D. High Confinement
The global energy confinement time on NSTX has
been measured and compared to conventional tokamak
scaling relations. Shown in Figure 6 is the energy
confinement time as determined by magnetic analysis
using the EFIT code [11] plotted against two
conventional tokamak scaling relations, ITER89P and
ITER98y2, which were originally developed for the ITER
project [12]. NSTX confinement times typically exceed
the more pessimistic ITER89P by a factor of two. The
confinement times exceed the H-mode ( H igh
confinement mode) confinement scaling by a factor of up
to 1.4, even though the bulk of the discharges in the plot
Figure 6 Scatter plots of measured energy confinement
times plotted versus the a) ITER89P and b) ITER98y2
scaling relations
were limited discharges and were not in H-mode. The
enhanced confinement time is an important result for the
spherical torus.
E. H-Mode
The highest confinement times achieved to date have
occurred when the plasma was in H-mode. Confinement
times in excess of 100ms have been achieved with plasma
currents of 800kA. The H-mode discharge are
characterized by a sudden drop in the measured
deuterium-α emissions simultaneous with an acceleration
of the density increase.
The H-mode is a relatively new phenomenon on NSTX
and has not yet been fully exploited due to issues with
reproducibility, probably associated with divertor tile
conditioning. Operation in H-mode holds promise for
further improvements in plasma performance.
V.
PROFILES AND POWER BALANCE ISSUES
Measured profiles of density and temperature for a high
performance neutral beam heated discharge from NSTX
are shown in Figure 7. The electron temperature and
density were measured with a multi-point multi-pulse
Thomson scattering diagnostic which measures the
Doppler broadening of laser light classically scattered off
electrons. The ion temperature was measured using
charge exchange recombination spectroscopy (CHERS)
which measures the Doppler broadening of a carbon VI
charge exchange emission line. CHERS requires that at
least one neutral beam source be on to act as a source of
charge exchange electrons. The ion temperatures have
Figure 7 Electron temperature, electron density, and ion
temperature profiles for high performance shot 104001.
been independently confirmed with neutral particle
analyzer measurements, and the electron temperature
measurements are consistent with flux consumption
calculations made using a neoclassical resistivity model
[13]. The electron density measurements have been
confirmed against both reflectometry and interferometry.
These profiles have been used to calculate the classical
collisional power balance for NSTX. The calculation was
done using the TRANSP code, which is a standard
analysis tool developed for use on conventional
tokamaks. The results of this calculation are shown in
Figure 8. The curious result is that ion difference power,
which is the remainder of the ion power flow after neutral
beam ion heating, transient effects, and electron-ion
collisional coupling are taken into account is large and
positive. This implies two important points: a) that there
is a possible anomalous ion heating mechanism, and b)
that the ion confinement is very good.

Page 7

Figure 8 Power balance profiles for shot 104001. Each
point on each trace represents the integral of the power
flow for each mechanism within the enclosed volume.
A mechanism has been proposed [14] whereby a fast
particle driven Alfvén instability, called the
compressional Alfvén eigen-mode (CAE), could heat
thermal ions. The power flow is from the fast particles to
the waves and from the waves to the bulk ions. This
mode has been tentatively identified on NSTX [15]. The
vertical line in the figure is the theoretically predicted
location of the peak CAE mode amplitude. The
correlation between the off axis maximum in the
derivative of the ion-electron coupling is intriguing, but
further work is required to make a firm correlation
between the presence of CAE modes and any anomalous
ion heating mechanism that may be identified on NSTX.
VI. SUMMARY
High performance discharges have been created on
NSTX using strong auxiliary heating, both by HHFW and
by neutral beams. βt ~20% has been achieved using
neutral beam heating. The maximum achievable βt scales
with normalized current. Ion temperatures of 2.0keV have
been measured during beam heating. H-mode has been
observed on NSTX and energy confinement times in
excess 100ms have been measured. High harmonic fast
wave heating has been effectively used to heat electrons
to 1.8keV, with no apparent density rise and without
heating ions. Coaxial Helicity Injection has been used to
drive up to 260kA of toroidal current. Energy
confinement times have been measured and exceed
ITER98pby2 scaling by a factor of 1.4. Power balance
calculations made using the measured kinetic profiles
indicate that the power loss from the ions to the electrons
exceeds the calculated classical input power to the ions.
Compressional Alfvén eigenmodes have been observed
on NSTX and are a potential candidate to explain the
inferred anomalous ion heating.
VII. REFERENCES
[1] S. KAYE, M. ONO, Y.-K. M. PENG, et al., Fusion
Technology 36, 16 (1999)
[2] Y.-K. M. PENG and D. J. STRICKLER, Nucl. Fusion
26, 769 (1986)
[3] G. REWOLDT, et al., Phys. Plasmas 3, 1 (1996)
[4] M. GRYAZNEVICH, R. AKERS, P. G. CAROLAN,
N. J. CONWAY, D. A. GATES, et al., Phys. Rev. Lett.
80, 3972 (1998)
[5] D. A. GATES, R. AKERS, L. APPEL, P. G.
CAROLAN, et al., Phys. Plasmas 5, 1775 (1998)
[6] A. SYKES, et al., Nucl. Fusion 32, 769 (1994)
[7] M. ONO, et al., Phys. Rev. Lett. 59, 2165 (1987)
[8] T. R. JARBOE, et al., Fusion Technol. 15, 7 (1989)
[9] M. ONO, Phys. Plasmas 2, 4075 (1995)
[10] F. TROYON, et al., Plasma Phys. And Control.
Fusion 26, 209 (1984)
[11] S. A. SABBAGH, et al., to be published in Nucl.
Fusion, (2001)
[12] “The ITER Physics Basis”, ITER Expert Groups,
Nucl. Fusion 39 2137-2664 (1999)
[13] J. MENARD, et al., to be published in Nucl. Fusion,
(2001)
[14] D. A. GATES, N. N. GORELENKOV, R. B.
WHITE, submitted to Phys. Rev. Lett. (2001)
[15] E. FREDRICKSON, et al., submitted to Phys. Rev.
Lett. (2001)

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