**Progress towards high performance, steady-state Spherical Torus \***

[M. Ono]{.underline}, M.G. Bell, R.E. Bell, T. Bigelow^1^, M. Bitter, W.
Blanchard, J. Boedo^2^, C. Bourdelle^3^, C. Bush^1^, W. Choe^4^, J.
Chrzanowski, D.S. Darrow, S.J. Diem^5^, P.C. Efthimion, J.R. Ferron^6^,
R.J. Fonck^5^, E.D. Fredrickson, G.D. Garstka^5^, D.A. Gates, L.R.
Grisham, W. Heidbrink^7^, K.W. Hill, J.C. Hosea, T.R. Jarboe^8^, D.W.
Johnson, R. Kaita, S.M. Kaye, C. Kessel, J.H. Kim^4^, M.W. Kissick^5^,
S. Kubota^9^, H.W. Kugel, B.P. LeBlanc, K. Lee^10^, S.G. Lee^11^, B.T.
Lewicki^5^, R. Maingi^1^, R. Majeski, J. Manickam, R. Maqueda^12^, T.K.
Mau^2^, E. Mazzucato, S.S. Medley, J. Menard, D. Mueller, B.A.
Nelson^8^, C. Neumeyer, N. Nishino^13^, C.N. Ostrander^5^, D.
Pacella^14^, F. Paoletti^15^, H.K. Park, W. Park, S.F. Paul, Y.-K. M.
Peng^1^, C.K. Phillips, R. Pinsker^6^, P.H. Probert^5^, S. Ramakrishnan,
R. Raman^6^, M. Redi, A.L. Roquemore, A. Rosenberg, P.M. Ryan^1^, S.A.
Sabbagh^15^, M. Schaffer^6^, R.J. Schooff^5^, C.H. Skinner, A.C.
Sontag^5^, V. Soukhanovskii, T. Stevenson, D. Stutman^16^, D.W.
Swain^1^, E. Synakowski, Y. Takase^17^, X. Tang^12^, G. Taylor, K.L.
Tritz^5^, E.A. Unterberg^5^, A. Von Halle, J. Wilgen^1^, M. Williams,
J.R. Wilson, X. Xu^18^, S.J. Zweben, R. Akers^19^, R.E. Barry^1^, P.
Beiersdorfer^18^, J.M. Bialek^15^, B. Blagojevic^16^, P.T. Bonoli^20^,
M.D. Carter^1^, W. Davis, B. Deng^10^, L. Dudek, J. Egedal, R. Ellis, M.
Finkenthal^16^, J. Foley, E. Fredd, A. Glasser^12^, T. Gibney, M.
Gilmore^21^, R.J. Goldston, R.E. Hatcher, R.J. Hawryluk, W. Houlberg^1^,
R. Harvey^22^, S.C. Jardin, H. Ji, M. Kalish, J. Lawrance^23^, L.L.
Lao^6^, F.M. Levinton^24^, N.C. Luhmann^10^, R. Marsala, D. Mastravito,
M.M. Menon^1^, O. Mitarai^25^, M. Nagata^26^, M. Okabayashi, G. Oliaro,
R. Parsells, T. Peebles^9^, B. Peneflor, D. Piglowski, G.D. Porter^18^,
A.K. Ram, M. Rensink, G. Rewoldt, P. Roney, K. Shaing, S. Shiraiwa^16^,
P. Sichta, D. Stotler, B.C. Stratton, R. Vero^16^, W.R. Wampler^27^,
G.A. Wurden^12^, X.Q. Xu^18^, L. Zeng^9^, W. Zhu^9^

*Princeton Plasma Physics Laboratory, Princeton University, New Jersey,
USA*

^1^ *Oak Ridge National Laboratory, Oak Ridge, Tennessee, USA*

^2^*University of California, San Diego, California, USA*

^3^*CEA Cadarache, France*

^4^*Korea Basic Science Institute, Taejon, Republic of Korea*

^5^*University of Wisconsin, Wisconsin, USA*

^6^*General Atomics, San Diego, California, USA*

^7^*University of California, Irvine, California, USA*

^8^*University of Washington, Seattle, Washington, USA*

^9^ *University of California, Los Angeles, California, USA*

^10^ *University of California, Davis, California, USA*

^11^ *Korea Basic Science Institute, Taejon, Republic of Korea*

^12^ *Los Alamos National Laboratory, Los Alamos, New Mexico, USA*

^13^ *Hiroshima University, Hiroshima, Japan*

^14^*ENEA, Frascati, Italy*

^15^*Columbia University, New York, N.Y., USA*

^16^*Johns Hopkins University, Baltimore, Maryland, USA*

^17^*Tokyo University, Tokyo, Japan*

^18^*Lawrence Livermore National Laboratory, Livermore, California, USA*

^19^*Euratom-UKAEA Fusion Association, Abingdon, Oxfordshire, United
Kingdom*

^20^*Massachusetts Institute of Technology, Cambridge, Massachusetts,
USA*

^21^ *University of New Mexico at Albuquerque, New Mexico, USA*

^22^*Compx, Del Mar, California, USA*

^23^*Princeton Scientific Instruments, Princeton, New Jersey, USA*

^24^*Nova Photonics, Princeton, New Jersey, USA*

^25^*Kyushu Tokai University, Kumamoto, Japan*

^26^*Himeji Institute of Technology, Okayama, Japan*

^27^*Sandia National Laboratories, Albuquerque, New Mexico, USA*

**Introduction** -- The spherical torus \[1\] (ST) research conducted
worldwide has made a remarkable progress in recent years. In the US, the
ST experimental research is carried out in the experimental facilities
including National Spherical Torus Experiment (NSTX) \[2\] and CDX-U
\[3\] at Princeton Plasma Physics Laboratory, PEGASUS \[4\] at
University of Wisconsin, and HIT-II \[5\] at University of Washington.
The Innovative Confinement Concept (ICC) program, the Innovative
Diagnostic Development program, the US theory program and the Virtual
Laboratory for Technology program are four crucial elements of the US ST
research effort. The ST program is presently focusing on two broad
goals. The first goal is to assess the attractiveness of the ST as a
fusion energy concept such as the ST-based CTF (Component Test Facility)
and Demo. The US ST program is indeed well aligned with the recently
developed FESAC fusion energy development path plan \[6\]. The second
goal is to use ST plasma characteristics to foster a deeper
understanding of critical toroidal physics issues.

**Scientifically important ST physics issues** -- The ST configuration
offers the following scientifically important physics conditions to
develop a deeper understanding of high temperature toroidal plasmas as
well as astrophysics plasmas:

• High \<β~T~\>(\~ 40%) & central β~0~ (\~100%). Because of the
favorable MHD stability at low aspect ratio A=R/a \< 2 \[1\], the ST
plasmas have already accessed high average toroidal beta of ≤ 35-40% and
central beta of order unity. This property permits sufficient fusion
power production at relative low confining toroidal field and, thus,
reduces the power plant cost and recirculating power. It should be noted
that the unity beta condition is also relevant for the physics of space
plasmas.

• Strong plasma shaping & self fields (A ≥1.27, δ≤ 0.8, κ ≤ 2.5,
B~p~/B~t~ \~1). Because of the strong toroidicity and shaping produced
in ST plasmas, the investigation in these extreme conditions could lead
to improved and deeper understanding of the toroidal plasmas.

• Large plasma flow (V~rotation~/V~A~ \~0.3). Since at unity beta, the
Alfvén velocity V~A~ approaches the ion thermal velocity, it is
relatively easy to access high Alfvén Mach number plasmas in NSTX. This
property could relax the condition for the wall stabilization by the
plasma rotation for ST reactor.

• Large flow shearing rate (γ~ExB~ ≤ 10^6^/s). With strong plasma
rotation and toroidicity, ST plasmas could generate significant sheared
flows, which could suppress the long wave length turbulence to improve
confinement.

• Supra-Alfvénic fast ions (V~fast~/V~A~ \~4--5). Again, this condition
can be readily created because of the low Alfvén velocity of ST. The
wav-particle interaction in this regime could be of relevance to the
alpha particle physics in burning plasmas such as ITER.

• High dielectric constant (ε \~ 50). The high plasma dielectric
property drastically modify the wave propagation characteristics of some
of the plasma waves. While this property exclude utilization of certain
types of plasma waves such as ECH and lower hybrid waves, it gives rise
to new opportunities for waves such as high harmonic fast waves (HHFW)
and electron Bernstein waves (EBW).

• Large mirror ratios in edge B field. Near the plasma boundary, the
toroidal field could vary as much as a factor of 5, producing a large
mirror ratio. This property could lead to edge power flow modification.

**Physics requirements for the steady-state high-performance plasmas**
-- It is important to note that these unique physics properties of ST
described above could also help ST achieve its long range goal of
steady-state operations at high performance needed for ST-based
reactors. The physics requirements for ST fusion systems can be
summarized as follows:

• MHD stability at high β~~and β~~: To produce required fusion power
at low toroidal field, high β~~ is needed. Since self-driven current
fraction is proportional to ε β~P~ ≡ ε 2~0~〈p〉 / B~P~^2^ and β~T~ ∝
β~N~^2^ / β~P~, very high value of β~N~ is needed for high bootstrap
current fraction. Typically, β~~\~ 20%, β~~\~ 6 is needed for CTF
and much more challenging β~~≥ 40%, β~~\~ 8 is needed for Power
Plants (e.g., ARIES-ST). This power plant regime will require advanced
ST operations with plasma beta near the ideal stability limits and
therefore will likely to require some kind of active feed back
stabilization of MHD modes.

• Transport and Confinement: Since the fusion power production is very
strong funtion of the plasma confinement (P~fusion~ is proportional to
the H-factor to as much as 7^th^ power), it is important to understand
the confinement trends and improve the predictive capability of
confinement. The systems code studies of ST-based CTF and Power Plant
design suggest that the required global confinement should be in the
range of H~98pby,2~ \~ 1.4 - 1.7.

• Power and Particle Handling: Because of the small major radius of ST
reactors, the expected P/R is much larger than that of conventional
fusion reactors by a factor of \~ 2 to 3. While, this is a stringent
requirement, the unique ST geometry may provide a solution to this
problem such as large flux expansion together with innovative plasma
facing component solution such as liquid lithium.

• Solenoid-Free Start-Up: An elimination of in-board ohmic heating
solenoid is required for the ST to function as an attractive fusion
power plant. An in-board ohmic solenoid, along with the shielding needed
for its insulation, increases the size and, hence, the cost of the
plant. Thus ST-based fusion systems including the CTF and power plant
designs assume complete elimination of the ohmic solenoid.

• Integrating Scenarios: While it is often logical to explore each
physics topic independently to facility understanding, it is necessary
to demonstrate all the essential aspects of the physics requirements
simultaneously in an integrated manner to be credible as a fusion
system.

**US ST Facilities --** There are four US ST facilities: The National
Spherical Torus Experiment (NSTX) at the Princeton Plasma Physics
Laboratory (PPPL) is a 1 MA class facility designed to evaluate the
physics principles of the ST, which is characterized by strong magnetic
field curvature and high β~T~, the ratio of the plasma pressure to the
applied toroidal magnetic field pressure. The NSTX facility efficiently
utilized the TFTR site credit to optimize the facility capability in
terms of power supplies and auxiliary heating systems. There are three
smaller facilities dedicated to study targeted innovative ST research
areas. The PEGASUS facility at University of Wisconsin is a few 100 kA
class facility aiming to investigate very low aspect ratio region aiming
to bridge the physics gap between spheromaks and STs. The HIT-II
facility is also a few 100 kA class facility dedicated to develop an
innovative non-inductive plasma start-up concept using coaxial helicity
injection (CHI). The CDX-U facility at PPPL is now focusing its effort
to test concepts utilizing lithium coating and liquid lithium plasma
facing components to develop an attractive power and particle handling
methods for STs. The following Table I shows the parameters of the four
facilities:

  --------------- ------------- ------------- -------------- -------------
  Devices         NSTX          PEGASUS       HIT-II         CDX-U

  Ip (MA)         ≤1.5          ≤ 0.3         ≤0.26          ≤ 0.1

  R (M)           0.86          0.25- 0.45    0.3            0.33

  A(R/a)          ≥1.26         1.1 -- 1.4    1.5            1.5

  B~T~ R (kG-m)   5             0.3           0.8            0.7

  τ−pulse (sec)   ≤ 1.1         ≤ 0.1         ≤ 0.1          ≤ 0.025

  Elongation      ≤ 2.2         1 -- 3.5      1.5            1.55

  OH flux (V-S)   0.7           0.1           0.1            0.1

  NBI (MW)        7             NA            NA             NA

  HHFW (MW)       6             ≤ 1           NA             0.1

  CHI (MA)        0.4           NA            0.2            NA
  --------------- ------------- ------------- -------------- -------------

Table 1

**Progress on MHD stability at high β~~and β~~ in NSTX -** The beta
limit investigations were conducted mainly in NSTX due to its strong
auxiliary heating capability \[7\]. The PEGASUS device also investigated
the plasma beta with ohmic heating taking advantage of the low-aspect
ratio geometry as descrived below. In Fig. 1, the achieved normalized
β~T~ vs normalized β~pol~ data points are shown. In 2001, a rapid
progress was made to reach β~T~ \~ 25% which is near the so-called
no-wall beta limits. In 2002, by realigning outer poloidal field coils,
the n=1 error field component was reduced by a factor of 10. This error
field reduction improved the plasma beta values dramatically as shown in
the figure. The β~T~ value increased to 35% and the β~pol~ value doubled
from 0.6 to 1.4. The beta improvement was also aided by the ready access
to the H-mode which broadened the pressure profiles as broader pressure
profiles improve MHD stability \[8\]. Analysis of many plasmas with high
β~N~ indicates that the no-wall stability limit has been indeed
exceeded, and that wall stabilization is likely a critical player in
achieving these high beta state \[9\]. In Fig. 2, the 1.2 MA discharge
evolution of the β~T~ \~ 35% shot is shown. As can be seen from the
figure, the discharge reached 35% β~T~ shortly after entering H-mode.
The presence of n=1 /m=1 mode seems to regulate the beta value and
keeping the high beta maintained for a period of over energy confinement
time until the end of current flattop. In Fig. 3, we show the high beta
poloidal shot at 800 kA. As can be seen, the loop voltage drops from 2 V
to about 0.1V around 0.3 sec coincident with the rise of β~p~. We
estimate that the non-inductive current drive fraction is about 60% due
to the bootstrap currents and NBI current drive in this phase. The
plasma internal inductance l~i~ stays nearly constant for about 400 msec
much longer than the plasma current skin time. The H-mode access which
kept the pressure profile broad and the EFIT calculated q-minimum
staying around 2 contributed to the MHD stability. The rapid plasma
rotation is likely helping the stability further with the wall
stabilization effect allowing the plasma to stay above the no-wall
stability limit for many tens of wall resistive time. The achieved
parameters β~N~ H89P \~ 15 with β~T~ \~ 15% of the high poloidal
discharges is already comparable to that is required for CTF. The plasma
density however rises continuously leading to an MHD event which
indicates that the particle control is an issue which must be addressed
in order to realize truly steady state regime \[10\].

**PEGASUS High Beta Experiment -** High toroidal beta plasmas are
obtained by operation at very low toroidal field, and cover a regime of
β~T~ vs I~N~ space similar to neutral-beam heated high-β~T~ plasmas in
START and other ST experiments (Fig. 5)\[4\]. As indicated, β~T~ values
up to 25% and β~N~ \~ 5 have been obtained with no evidence of a beta
limit to date. Densities range up to the Greenwald limit (\~I~p~/πa^2^).
Stored energies are consistent with values expected from the ITER98pby1
confinement scaling. Plasma startup is characterized by high current
ramp rates (15-45 MA/s), low internal inductance (*l*~i~ \~ 0.3), and
high elongation. Two-dimensional images of X-ray emission from Pegasus
plasmas have enabled, for the first time, non-perturbative measurement
of the plasma current profile in a spherical torus \[4\]. This
measurement is accomplished by determining intensity contours from the
image and using the contours as inputs in the solution of the MHD
equilibrium state. The shape of the contours is a strong constraint on
the current profile in the equilibrium reconstruction. The safety factor
profile is also well-constrained by this technique. A measured
q-profile, showing near-zero central shear. Plasmas with β~T~ \~ 1 as A
approaches unity in the tokamak-spheromak overlap region appear
accessible with the addition of planned new capabilities which are aimed
at lowering the plasma resistivity and manipulating the evolution of the
q-profile to suppress limiting MHD activity. These include high-power RF
heating, a transient increase in the toroidal field for a stabilized
formation stage, loop voltage control and significantly increased ohmic
volt-seconds, an upgraded equilibrium field system for shape and
position control, and separatrix operation. Operation with a two-strap
high-power Higher Harmonic Fast Wave heating system has begun. Initial
loading tests show an impedance of 1 ohm, and up to 200 KW has been
injected to date. TSC modeling of fast TF rampdown scenarios indicate
accessible paths to regimes of higher current and increased stored
energy.

**Supra-Alfvénic fast ion induced high Frequency MHDs** -- The ST high
beta regimes owing to high beta provide a good test bed to investigate
the wave particle interactions for the Supra-Alfvénic fast ions (V~NBI~
/V~A~ \~4--5). This type of regime is similar to those encountered for
alpha-heated discharges such as ITER. In NSTX NBI heated discharges, a
wide variety of such instabilities has been seen in NSTX at frequencies
ranging from a few kHz to many Mhz \[11-13\]. In the frequency range
below about 200 kHz, a form of the fishbone or energetic particle mode
has been seen, as well as modes that appear to be similar to the TAE
modes of conventional tokamaks. Unlike in conventional tokamaks, the
frequency ranges of these two classes of instabilities have substantial
overlap, complicating the experimental identification and theoretical
analysis. Significant fast ion losses have been correlated, under some
conditions, with the appearance of both of these types of modes.

**Progress on Transport and Confinement -** The global confinement times
in neutral beam heated NSTX plasmas compare favorably to the ITER-89P
empirical scaling expression as well as the ITER-98(pby,2) scaling rule
\[14, 15\]. In recent years, H mode operations have become routine on
NSTX, aided by improved wall conditioning and reduced error fields.
Access to the H mode is easiest in the lower single null configuration,
but H modes have been obtained in double null as well. The power
threshold of several hundred kW in some cases and is exhibiting a
secular fall as wall conditions improve. The in-board gas injection
shows favorable H-mode access compared to the outboard gas injection in
the recent experiments \[8\]. In Fig. 5. NSTX H-mode experimental
confinement data points obtained in quasi-steady conditions are shown
compared to the ITER 98py2 H-mode scaling. The soild circles are the
global confinement time and the rhombus points are the confinement with
energetic component and NBI lost components removed. The H-mode data
shows up to 50 % global confinement improvement compared to the H-mode
scaling. The level of confinement enhancement is comparable to that is
needed for the future ST devices. The global confinement scaling study
has begun on NSTX. Accurate determination of R/a dependence is an active
ITPA research topic. The L-mode data shows a similar scaling as found in
the conventional aspect ratio tokamaks τ~E~^NSTX-L^ \~ I~p~^0.76^
B~T~^0.27^ P~L~^-0.76^. The H-mode data on the other hand shows much
less power degradation P^-0.5^ which is encouraging. But it should be
also noted that the H-mode parameteric dependencies are turning out to
be more complex and non-linear, showing the needs for further refinement
for this low-aspect-ratio high beta ST regimes.

**Transport diffusivities -** The power balance analysis of the NSTX NBI
heated discharges is shown in Fig. 6 with ordering of χ~φ~ \< χ~i~ ≤
χ~neo~ \< χ~e~. The ion thermal conductivity χ~i~ appears to track the
predictions from neoclassical theory quite well, and electron thermal
conductivity χ~e~ that is significantly larger than χ~i~. The momentum
diffusivity χ~φ~ is much smaller than χ~i~ in this analysis,
qualitatively consistent with expectations from neoclassical theory. In
general, the observed diffusivity profiles are unusual in that the
thermal diffusivities are falling with minor radius. This type of
diffusivity profiles tends to give broader pressure profiles which is
favorable for plasma high beta stability. Owing to the small χ~φ~, the
NSTX plasma rotates relatively rapidly at 200 -- 300 km/sec reaching
rather high Mach number of V~rotation~/V~A~ \~0.3. The improved ion
confinement appears to be indeed correlated with the plasma rotation.
The plasma rotation could also provide stabilizing influence on the MHD
modes as discussed earlier. The observed χ~i~ \~ χ~neo~ and χ~φ~ \< χ~i~
suggest long wavelength turbulence may be suppressed. In order to
develop fundamental understanding of the plasma transport, a variety of
theoretical tools are utilized on the problem. The growth rates computed
by GS2 indeed show that the ExB sharing rate is sufficiently greater
than the predicted growth rates of ITG range turbulence as shown in Fig.
7, consistent with the low ion thermal diffusivity \[16, 17\]. In the
ETG range of short wavelength modes, the linear instability growth-rate
is significantly larger than the shearing rate, consistent with
relatively large observed χ~e~. This neo-classical ion transport regime
can provide a unique test bed to investigate the electron transport
physics in NSTX.

**Progress on Power and Particle Handling --** The NSTX boundary physics
research thus far focused on power and particle balance \[10\]. High
heat flux on the target plate has been measured in lower-single null
(LSN) divertor plasmas. The peak heat flux in a lower single null
ELM-free H-mode plasmas with 4.5 MW of heating power has reached 10
MWm^-2^, with a full-width half-max of 2 cm at the outer target plate
approaching the spatial resolution of the IR camera used to make the
measurement. Peak heat flux in H-mode plasmas increases with NBI heating
power. The peak heat flux at the inboard target is typically 0.5 -- 1.5
MW/m^2^, with a profile full-width half max. of \~ 10 cm. The power
flowing to the inboard side is typically 0.2 -- 0.33 of the outboard
power. Similar in-out asymmetry ratios have been observed in the MAST
devices. Double-null high triangularity discharges appear to have much
lower peak heat flux than lower-single null. A preliminary measurement
is shown in Fig. 8 which shows much larger degrees of heat dispersion,
\~ 3 MW/m^2^, consistent with the flux expansion expectation for the
high triangularity discharges as shown in Fig. 8. This is a promising
result to minimize the peak heat load on the divertor plates. It should
be noted that the ARIES-ST configuration also assumes similar high
triangularity configuration to reduce the divertor peak heat load to an
acceptable level.

**Liquid lithium limiter experiment on CDX-U -** The primary research
topic under investigation on CDX-U is the use of liquid metals, and in
particular liquid lithium, as plasma facing components (PFCs) for the ST
and tokamak \[18\]. The primary motivation for these experiments is a
test of liquid metal PFCs as a potential engineering solution to the
problems of high heat flux and erosion of the first wall, which is
expected in a reactor. However, liquid lithium PFCs have attractive
physics advantages as well. A shallow toroidal tray, which encircles the
center stack and forms the lower limiting surface for the plasma, has
been installed in the vessel. The tray has a major radius of 34 cm, is
10 cm wide and 0.5 cm deep, and is fitted with heaters to allow
operation at temperatures up to 500C; the typical temperature during
tokamak operations is 300C. The tray has been filled with approximately
1 liter of liquid lithium; a photograph of part of the tray, installed
in CDX-U and filled with (highly reflective) liquid lithium is shown in
Figure 9. Plasma operations with a bare stainless steel tray and with a
liquid lithium-filled tray have been compared. The use of liquid lithium
as a limiter material results in a significant reduction in the oxygen
impurity in the discharge. Recycling is reduced during lithium
operation, resulting in a requirement for an eightfold increase in the
gas puffing rate in order to maintain a plasma density comparable to
discharges with the bare, fully recycling, stainless steel tray. The
plasma loop voltage during lithium operation is reduced from 2V to 0.5V
at comparable plasma current. The lithium remains quiescent and is
confined to the tray during plasma operations. The performance
enhancement produced by the use of lithium as a PFC is far more evident
than improvements produced by titanium gettering or boronization in
CDX-U. Because of the encouraging results from CDX-U, NSTX will be
testing the lithium techniques to solve the heat and particle issues.

**Progress on Solenoid-Free Start-Up -** The elimination of in-board
ohmic heating solenoid is required for the ST to function as an
attractive fusion power plant. An in-board ohmic solenoid, along with
the shielding needed for its insulation, increases the size and, hence,
the cost of the plant. Thus ST-based fusion systems including the CTF
and power plant designs assume complete elimination of the ohmic
solenoid. Indeed designs for advanced tokamak reactors such as the ARIES
series of devices also eliminate the central solenoid. NSTX is
investigating two alternative approached for the central solenoid-free
start-up. The first one is the coaxial helicity injection developed by
the HIT-II group at University of Washington \[5\]. The other is the
outer PF coil start-up concept.

***Coaxial Helicity Injection*** - This CHI concept is an outgrowth of
the spheromak research. A number of smaller helicity injection
experiments were performed with some success prior to introducing it on
NSTX. On NSTX, CHI has already produced about 400 kA of toroidal current
with a record current amplification of 14 accompanied by distinct n=1
relaxation activities \[19\]. The CHI near term research goal is to
establish an understanding of the current penetration process and to
confirm the existence of closed flux surfaces and to demonstrate
coupling of the CHI produced current to other non inductive current
drive methods. Theoretical modeling of CHI have been pursued with 3-D
MHD codes such as CHIP to understand the reconnection processes during
CHI. \[20\]

***Transient Coaxial Helicity Injection in HIT-II --*** To complement
the so-called slow CHI start-up research, an innovative transient CHI
start-up method has been developed recently on HIT-II device \[5\]. By
applying a short pulse CHI bias voltage, one can create a high quality
(low impurity) CHI discharge which can detach from the electrode and
form a closed flux surface discharge. This well formed clean CHI
discharge provides an ideal target for OH induction. As shown in Fig.
10, this method indeed saves OH volt-second and produce much higher
current than the OH only case. Using this method, the HIT-II produced
record plasma current of 265 kA. This method will be tested on NSTX. If
it works well, one can apply RF heating and current drive instead of OH
to make the plasma start-up and ramp-up completely non-inductive.

***Outer-Poloidal-Field-Coil Start-up*** - Because of the criticality of
OH-solenoid-free start-up issue for ST research, the NSTX is pursuing
alternate methods of OH-solenoid-free start-up. While it is very
difficult to build an central OH solenoid for ST due to the space
limitation, the ST geometry offers an interesting possibility to utilize
outer PF coils for solenoid-free start-up \[21\]. By appropriately
controlling the outer PF coil currents, one can create a high quality
field null as shown in Fig. 11. Using a maximum coil current of 20 kA,
this method can generate up to 10 V of loop voltage while retaining a
good quality field null for the required duration of 3 msec. The Lloyd
condition with strong preionization (E~T~B~T~/B~p~ \> 0.12 kV/m) can be
satisfied for a large plasma volume. The available flux for this case is
about 120 mV-S, which may be sufficient for generating about 300 kA of
plasma current in NSTX. If the Lloyd condition can be relaxed further
using strong heating and current drive as observed on JT-60U \[22\], the
V-S availability can be further increased for this method. This poloidal
field coil start-up will be tested on NSTX since, if proven , it can be
readily applied toward future devices.

**Progress on Integrating Scenarios -** While it is often logical to
explore each physics topic independently to facility understanding, it
is necessary to demonstrate all the essential aspects of the physics
requirements simultaneously in an integrated manner to be credible as a
fusion system. Integrated High Performance Scenarios. In order to
achieve the ST reactor relevant physics goals of 40% β~T~ , I~NI~ =
100%, τ~pulse~ \>\> τ~skin~, the following additional tools are planned
on NSTX.

• Enhanced shaping improves MHD stability. As shown in Fig. 12, the
achieved beta values trend favorably toward higher triangularity and
high elongation regimes. A planned modification of inner PF coils in
NSTX will permit the investigation of the κ =2.4 and δ = 0.8 regime.
This favorable trend is consistent with the MHD stability calculations
predicting stable 40% β~T~ , I~NI~ = 100% for the κ =2.4 and δ = 0.8
regime.

• Since 40% β~T~ , I~NI~ = 100% regime requires wall stabilization, we
will install active feed back coils. NSTX will explore the regime
approaching the ideal wall limit using the active coil system together
with the plasma rotation.

• In order to achieve τ~pulse~ \>\> τ~skin~ in the advanced regime, it
is important to control heat and particles. Maintenance of moderate n~e~
is important for efficient current drive. NSTX plans to install divertor
lithium wall coating and cryo-pump system to control the particle
exhausts. Improved fueling using supersonic gas injector and advance
fueling techniques are also planned.

• The existing 6 MW of HHFW heating \[23\] will contributes to bootstrap
current by raising T~e~. and direct current drive. The electron heating
by HHFW has been demonstrated as shown in Fig. 13 (a) where an electron
ITB was formed and resulted in high Te regime. In Fig. 13 (b), the
current drive was demonstrated by confirming differences in V~loop~ with
co and counter-directed waves maintained for over the skin-time \[24\].
The observed V~loop~ differences can be attributable to \~ 100 kA of
driven current which is consistent with a theoretical estimate. However,
a theoretical modeling calculation indicates that HHFW off-axis CD
efficiency may be reduced in the high thermal ion beta regimes due to
the energetic and bulk ion absorption and trapping effects. The
energetic ion interaction with HHFW has been observed on NSTX \[25\].

• The advanced regime also require off-axis current drive of \~ 100 kA
to maintain the central q \~ 2. A 4 MW 15 Ghz EBW system is planned to
provide needed off-axis CD as well as localized CD to stabilize NTMs as
needed. EBW appear to not suffer from parasitic ion absorptions as in
the case of HHFW. It should be also noted that the recent calculations
indicate that the so-called Ohkawa current (which is induced by trapping
the barely passing particles through heating) can greatly enhance the
EBW driven current efficiency in the off-axis regions due to the large
trapped particle populations in the ST plasmas. The calculated
efficiency ζ~ec~ = 0,4 at large r/a \>0.5 compares favorably to ECCD and
HHFWCD. The EBW lunching efficiency can be optimized using the EBW
emission measurements since the process is reversible. As shown in Fig.
14, the EBW emission measurements in NSTX agrees well with the
theoretical calculations \[26\]. It is believed that by controlling the
edge density gradient, the coupling efficiency can be raised toward 100%
as demonstrated in CDX-U \[3\].

### Conclusions and Discussions -- The ST research is making rapid progress. In the MHD area, NSTX achieved 35% β~T~ with NBI heating and PEGASUS reached 25 % β~T~ with just ohmic heating. NSTX also reached high β~p~ regime (I~NI~ fraction \~ 60%) with β~N~ H~89p~ = 15 at β~~= 15% sustained over τ~skin~. The β and confinement parameters reached are comparable to that is needed for CTF. Neo-classical-like ion confinement was observed in NBI heated discharges with H~98pby,2~ \~ 1.4. Very low χ~φ~ led to a rapid plasma rotation with V~rotation~ \~ 0.3 V~A~. The sheared flow stabilization of long wavelength ITG modes is consistent with the observation. High triangularity double null configuration shows a large reduction in peak heat flux. The CDX-U liquid lithium limiter experiment yielded encouraging results. Two innovative plasma start-up concepts using CHI method developed by HIT-II and outer poloidal field coils are presented. A number of new tools are planned to access the ST power plant relevant advanced regimes of 40% β~T~ and 100% J~NI~ through strong plasma shaping (δ \~0.8, κ \~ 2.4) and profile control. 

This research was supported by DoE contract DE-AC02-76CH03073Work and
DoE Grant DE-FG02-96ER54375.

**References:**

\[1\] Y-K M. Peng, and D. J. Strickler, Nuclear Fusion **26** (1986)
576.

\[2\] E.J. Synakowski et al., Proceedings of the 19^th^ IAEA meeting,
Lyon, France 2002, to be published in Nuclear Fusion.

\[3\] B. Jones, et al., Phys. Rev. Lett. **90**, 165001(2003)

\[4\] G.D. Garstka, et al., Phys. Plasmas **10**, 1705 (2003).

\[5\] R. Raman, et al., Phys Rev. Lett. **90**, 075005-1 (2003).

\[6\] FESAC Panel of Development Path. A Plan for the Development of
Fusion Energy, Preliminary Report to FESAC," issued March 2003.

\[7\] J. Menard et al., at this conference.

\[8\] R. Maingi et al., at this conference.

\[9\] S. Sabbagh et al., Proceedings of the 19^th^ IAEA meeting, Lyon,
France 2002, to be published in Nuclear Fusion.

\[10\] V.A. Soukhanovskii, et al., at this conference.

\[11\] E. Fredrickson, et al., at this conference.

\[12\] N. Gorelenkov et al., at this conference.

\[13\] E. Belova et al., at this conference.

\[14\] B. LeBlanc et al., at this conference.

\[15\] D. Stutman et al., at this conference.

\[16\] C. Bourdelle et al., accepted for publication in Phys. Plasmas
(2003).

\[17\] M. Redi et al., at this conference.

\[18\] D. Majeski, R. Kaita, et al., Private Communication.

\[19\] Jarboe, T.R., Raman, R, Nelson, B.A., *et al.*, "Progress with
helicity injection current drive," *19th IAEA Fusion Energy Conference*,
Lyon, IAEA-IC/P 10 (2002).

\[20\] Tang, X., 8th International ST Workshop, Nov. 18 -- 21, Princeton
Plasma Physics Laboratory, Princeton, NJ (2002).

\[21\] M. Ono and W. Choe, "Out-Board "Ohmic Induction" Coil for
Low-Aspect-Ratio Toroidal Plasma Start-up", Princeton University Patent
Disclosure 03-2003-1.

\[22\] Takase, Y. *et al.*, the Journal of Plasma and Fusion Research,
**78**, 719-721 (2002).

\[23\] M. Ono, Physics of Plasmas **2**, 4075(1995).

\[24\] P.M. Ryan et al., IAEA-CN-94/EX/P2-13, Lyon, France (2002)

\[25\] S. Medley et al., at this conference.

\[26\] G. Taylor, et al., Phys. Plasmas **10**, 1395 (2003)

![](media/image1.png){width="5.788194444444445in"
height="4.711805555555555in"}

Fig. 1. Nomalized plasma toroidal beta vs. normalized plasma polodial
beta. The normalized beta countours are as labeled. The taget beta
regime is shown as a large circle.

![](media/image2.pct){width="5.409722222222222in"
height="4.208333333333333in"}

Fig. 2. The discharge evolution of the β~T~ = 35 % discharge is shown.

![](media/image3.pct){width="4.147916666666666in"
height="5.440972222222222in"}

Fig. 3. The discharge evolution of high β~p~ shot.

![](media/image4.pct){width="5.76875in" height="3.8340277777777776in"}

Fig. 4. Toroidal beta vs normalized current for ohmically heated PEGASUS
discharges.

![](media/image5.png){width="5.048611111111111in"
height="3.6041666666666665in"}

Fig. 5. NSTX H-mode experimental confinement data points are shown
compared to the ITER 98py2 H-mode scaling. The circlar points are the
global confinement time and the rhombus points are the confinement with
energetic component and NBI lost components removed.

![](media/image6.png){width="6.575590551181103in"
height="5.799212598425197in"}

Fig. 6. Thermal and momentum diffusivities calculated from TRANSP power
balance calculations. Shown for comparison is the calculated
neoclassical thermal diffusivity from the NCLASS neoclassical model. The
plasma radius is about 65 cm.

![](media/image7.png){width="5.852777777777778in"
height="3.627083333333333in"}

Fig. 7. Growth rates computed by GS2 show that the EXB sharing rate
exceeds long wavelength modes thereby stabilizing them. The short
wavelength modes on the other hand may dominate (electron) transport.

![](media/image8.jpeg){width="1.4034722222222222in"
height="2.3180555555555555in"}

![](media/image9.pct){width="4.155555555555556in"
height="3.520138888888889in"}

![](media/image10.jpeg){width="1.4354166666666666in"
height="2.3534722222222224in"}

Fig. 8. Divertor heat load flux comparison for the single null and the
high triangularity double null discharges.

![](media/image11.jpeg){width="3.375in" height="2.53125in"}

> Figure 9. View of the lithium-filled toroidal tray in CDX-U through a
> port. The centerstack is the vertical column in the left side of the
> field of view. The tray is indicated by the arrow. Note the highly
> reflective surface, indicative of the liquid lithium in the tray.
>
> ![](media/image12.pct){width="4.582677165354331in"
> height="2.954724409448819in"}
>
> Fig. 10. Comparison of CHI + OH and OH only discharges as labeled. For
> all discharges a constant inductive voltage of 4 V is applied for 2
> ms, followed by 3.2 V for the next 6.8 msec.
>
> ![](media/image14.jpeg){width="2.5493055555555557in"
> height="4.415277777777778in"}
>
> ![](media/image15.jpeg){width="2.8465277777777778in"
> height="2.797222222222222in"}
>
> Fig. 11.The NSTX outer PF-only high quality null formation. (a) Flux
> contours. (b) E~T~ B~T~ / B~p~ contours in kV/m at the time of
> initiation.
>
> ![](media/image16.png){width="4.373611111111111in"
> height="3.1909722222222223in"}
>
> Fig. 12. Achieved plasma beta values vs plasma shaping parameters. (a)
> β~N~ vs plasma elongation. (b) β~T~ vs plasma triangularity.
>
> ![](media/image18.jpeg){width="3.9680555555555554in"
> height="2.5395833333333333in"}
>
> Fig. 13. High Harmonic Fast Wave Heating and Current Drive in NSTX.
> (a) Strong central heating by HHFW with creation of electron transport
> barrier. (b) Differences in V~loop~ with co and counter-directed waves
> indicate \~ 100 kA of current drive consistent with theoretical
> modeling estimate.
>
> ![](media/image20.png){width="4.163194444444445in"
> height="3.3784722222222223in"}
>
> Fig. 14. Observed electron Bernstein wave emission coefficient in
> NSTX. The curves are theoretical values.