Review of deuterium-tritium results from the Tokamak Fusion
Test Reactor*
K. M. McGuire,t H. Adler, P. Alling, C. Ancher, H. Anderson, J. L. Anderson,a)J. W. Anderson, V. Arunasalam, G. Ascione, D. Ashcroft, Cris W. Barnes,a) G. Barnes,
S. Batha,b) G. Bateman, M. Beer, M. G. Bell, R. Bell, M. Bitter, W. Blanchard,
N. L. Bretz, C. Brunkhorst, R. Budny, C. E. Bush,c) R. Camp, M. Caorlin, H. Carnevale,
S. Cauffman, Z. Chang,d) C. S. Chang,e) C. Z. Cheng, J. Chrzanowski, J. Collins,
G. Coward, M. Cropper, D. S. Darrow, R. Daugert, J. DeLooper, R. Dend~/) W. Dorland,g)
L. Dudek, H. Duong,h) R. Durst,d) P. C. Efthimion, D. Ernst,i) H. Evenson, ,N. Fisch,
R. Fisher,h) R. J. Fonck,d) E. Fredd, E. Fredrickson, N. Fromm, G. Y. Fu, T. Fujita,i>
H. P. Furth, V. Garzotto, C. Gentile, J. Gilbert, J. Gioia, N. Gorelenkov,k) B. Grek,
L. R. Grisham, G. Hammett, G. R. Hanson,c) R. J. Hawryluk, W. Heidbrink,l)H. W. Herrmann, K. W. Hill, J. Hosea, H. Hsuan, M. Hughes,m) R. Hulse, A. Janos,
D. L. Jass.by, F. C. Jobes, D. W. JOhnSOn L. C. Johnson, M. Kalish, J. Kamperschroer,J. Kesner,') H. Kugel, G. Labik, N. T. Lam, d) P. H. LaMarche, E. Lawson, B. LeBlanc,
J. Levine, F. M. Levinton,b) D. Lgesser, D. Long, M. J. Loughlin,n) J. Machuzak,il R. Majeski,
D. K. Mansfield, E. S. Marmar,') R. Marsala A. Martin, G. Martin, E. Mazzucato,
M. Mauel,O) M. P. McCarthy, J. McChesney,h) B. McCormack, D. C. McCune, G. McKee,d)D. M. Meade, S. S. Medley, D. R. Mikkelsen, S. V. Mirnov,k) D. Mueller, M. Murakami,c)J. A. Murphy, A. Nagy, G. A. Navratil,o) R. Nazikian, R. Newman, M. Norris, T. O'Connor,
M. Oldaker, J. Ongena,p) M. Osakabe,q) D. K. Owens, H. Park, W. Park, P. Parks,h)S. F. Paul, G. Pearson, E. Perry, R. Persing, M. Petrov/) C. K. Phillips, M. Phillips,m)S. Pitcher,S) R. Pysher, A. L. Qualls,c) S. Raftopoulos, S. Ramakrishnan, A. Ramsey,
D. A. Rasmussen,C) M. H. Redi, G. Renda, G. Rewoldt, D. Roberts,d) J. Rogers,
R. Rossmassler, A. L. Roquemore, E. Ruskov,l) S. A. Sabbagh,o) M. Sasao,q) G. Schilling,
J. SchiveIJ, G. L. Schmidt, R. Scillia, S. D. Scott, I. Semenov,k) T. Senko, S. Sesnic,
R. Sissingh, C. H. Skinner, J. Snipes,i) J. Stencel, J. Stevens, T. Stevenson, B. C. Stratton,
J. D. Strachan, W. Stodiek, J. Swanson,t) E. Synakowski, H. Takahashi, W. Tang,
G. Taylor, J. Terry,i) M. E. Thompson, W. Tighe, J. R. Timberlake, K. Tobita,il H. H. Towner,
M. Tuszewski,a) A. von Halle, C. Vannoy, M. Viola, S. von Goeler, D. Voorhees,
R. T. Walters, R. Wester, R. White, R. Wieland, J. B. Wilgen,c) M. Williams, J. R. Wilson,
J. Winston, K. Wright, K. L. Wong, P. WOSkOV,i) G. A. Wurden,a) M. Yamada,
S. Yoshikawa, K. M. Young, M. C. Zarnstorff, V. Zavereev,u) and S. J. Zweben
Plasma Physics Laboratory, Princeton University, Princeton, New Jersey 08543(Received 14 November 1994; accepted 24 February 1995)
After many years of fusion research, the conditions needed for a D-T fusion reactor have been
approached on the Tokamak Fusion Test Reactor (TFTR) [Fusion Technol. 21, 1324 (1992)]. For the
first time the unique phenomena present in a D-T plasma are now being studied in a laboratory
plasma. The first magnetic fusion experiments to study plasmas using nearly equal concentrations of
deuterium and tritium have been carried out on TFTR. At present the maximum fusion power of
10.7 MW. using 39.5 MW of neutral-beam heating, in a supershot discharge and 6.7 MW in ahigh-,Bp discharge following a current rampdown. The fusion power density in a core of the plasma
is =2.8 MW m- 3, exceeding that expected in the International Thermonuclear Experimental
Reactor (ITER) [Plasma Physics and Controlled Nuclear Fusion Research (International Atomic
Energy Agency, Vienna, 1991), Vol. 3, p. 239] at 1500 MW total fusion power. The energy
confinement time, 7£, is observed to increase in D-T, relative to D plasmas, by 20% and the nj(O)Tj(O) 7£ product by 55%. The improvement in thermal confinement is caused primarily by adecrease in ion heat conductivity in both supershot and limiter-H-mode discharges. Extensive
lithium pellet injection increased the confinement time to 0.27 s and enabled higher current
operation in both supershot and high-,Bp discharges. Ion cyclotron range of frequencies (ICRF)
heating of a D-T plasma, using the second harmonic of tritium, has been demonstrated. First
measurements of the confined alpha particles have been performed and found to be in good
agreement with TRANSP [Nucl. Fusion 34, 1247 (1994)] simulations. Initial measurements of thealpha ash profile have been compared with simulations using particle transport coefficients from He
gas puffing experiments. The loss of alpha particles to a detector at the bottom of the vessel is welldescribed by the first-orbit loss mechanism. No loss due to alpha-particle-driven instabilities has yet
been observed. D-T experiments on TFTR will continue to explore the assumptions of the ITER
design and to examine some of the physics issues associated with an advanced tokamak
reactor. © 1995 American Institute of Physics.

2176

Phys. Plasmas 2 (6), June 1995

1070-664X195/2(6)/21761131$6.00© 1995 American Institute of Physics

I. INTRODUCTION

For nearly 40 years, fusion researchers have studied theconfinement, heating, and stability of hydrogen (H) and deuterium CD) plasmas while reactor designs were based on using deuterium-tritium (D-T) fuelY Since December 1993
on the Tokamak Fusion Test Reactor (TFTR), it has become
possible to make a systematic -study of the differences between D and D-T fuels. These studies are needed to validate
the assumptions underlying reactor design such as that of theInternational Thermonuclear Experimental Reactor (ITER).
During the past year (1994), TFTR has created 280 D-T
discharges with tritium concentrations up to 60%, ion temperatures (Ti ) up to 44 keY, electron temperatures (Te) up to
13 keY, fusion power up to 10.7 MW, central fusion power
densities to 2.8 MW m- 3, fusion energy per pulse to 6.5 MJ.
The experimental D-T program on TFTR3 has significantly
extended the limited-objective D-T experiments previously
performed on the Joint European Tokamak (JET) which
achieved 1.7 MW of fusion power the -10% -tritium fueladmixtures. 4
The principal goals of the TFTR deuterium-tritium experiments are the following:(I) Safe operation of the tritium handling and processing
systems, and successful machine and diagnostic operation in a high radiation environment with 14 MeV neutrons;(2) documenting changes in confinement and heating goingfrom deuterium to tritium plasmas;(3) evaluating the confinement of a particles, including theeffect of a-induced instabilities, and measuring a hea~:­ing, and helium ash accumulation;*Paper lRVl, Bull. Am. Phys. Soc. 39, 1516 (1994).
tInvited speaker.
aJpermanent address: Los Alamos National Laboratory, Los Alamos, New
Mexico 87545.
bJpermanent address: Fusion Physics· and Technology, Torrance, California
90503.
c)Permanent address: Oak Ridge National Laboratory, Oak Ridge, Tennessee
37831.
dJpermanent address: University- of Wisconsin, Madison, Wisconsin 53706.
e)Permanent address: Courant Institute, New York University, New York
10003.
flpermanent address: Culham Laboratory, Abingdon, Oxford, England.
gJperrnanent address: University of Texas, Institute for Fusion Studies, Austin, Texas 78712.
hlPermanent address: General Atomics, San Diego, California 92186.
i)Permanent address: Massachusetts Institute of Technology, Cambridge,
Massachusetts 02139.
j)Perrnanent address: JAERI Naka Fusion Research Establishment, Naka,Japan.
k)Permanent address: TRINITI, Moscow, Russia.
I)Permanent address: University of California, Irvine, California 92717.
m)Permanent address: Grnmrnan Corporation, Princeton, New Jersey 08540.
n)Permanent address: JET Joint Undertaking, Abingdon, England.
o)Permanent address: Columbia University, New York, New York, 10027.
p)Permanent address: Ecole Royale Militaire, Brussels, Belgium.
q)Permanent address: National Institute of Fusion Studies, Nagoya, Japan.
r)Permanent address: Ioffe Physical-Technical Institute, Russia.
s)Permanent address: Canadian Fusion Fuels Technology Project, Toronto,
Canada.
tlpermanent address: EBASCO, Division of Raytheon, New York, New York
10048.
u)Permanent address: RRC Kurci1atov Institute, Moscow, Russia.

Phys. Plasmas, Vol. 2, No.6, June 1995

(4) demonstrating the production of -10 MW of fusion
power.
In this paper, a brief description will be given of the D-Dexperiments leading up to the D-T campaigns in: TFTR. The
optimization of the D-T power within the constraints imposed by the available heating power, the energy confinement, and the plasma stability are discussed. Finally, the possibilities for further impro"ements in the D-T fusion
performance of TFTR are discussed and how they will address key design considerations of a tokamak reactor utilizing deuterium-tritium fuel.
The experiments described in this paper were conducted
at a major radius of 2.45 -to 2.62 m, toroidal field at theplasma center from 4.0 to 5.6 T, and phlsma current from 0.6
to 2.7 MA. Deuterium and tritium neutral beams with energies up to 115 keV were injected to heat and fuel the plasma
with a total injected power up to 39.5 MW. Ion cyclotron
range of frequencies (ICRF) power,up to 8 ¥W has also
been used. The plasma boundary i~ defined by a toroidal
limiter composed of carbon-composite tiles in high heat flux
regions, and graphite tiles elsewhere.

II. TRITIUM SYSTEMS AND OPERATIONS

Initial tokamak experiments at low tritium concentration
were conducted in November 1993 and experiments at hightritium concentration began on 9 December 1993.
The tritium system on TFTR can handle concentrationsof tritium from relatively low levels of =0.5% up to 100%and is run routinely with up to 5 g of tritium (50 kCi)
on-site. s The tritium gas is brought on-site in an approved
shipping canister and transferred to a uranium bed where it isstored. The uranium bed is heated to transfer the gas to theneutral beam or torus gas-injection systems. The gas is theninjected into the torus or neutral beams and pumped by theliquid-helium cryopanels in the beam_ boxes. During plasma
operation, some of the gas is retained in the graphite-limiter
tiles in the vacuum vessel. The quantity of tritium in thevacuum vessel is restricted by PPPL requirements to 20 kCi.
The gas on the cryopanels is transferred to a gas holding tank(GHT) for inventory measurement, and subsequently is oxidized and absorbed onto molecular sieve beds. These beds
are shipped off-site for reprocessing or burial.
Since the start of D-T operation, 1.2 x 1020 D-T neutrons, equivalent to 340 MJ of fusion energy, have been produced. The activation of the vacuum vessel -2 weeks after
D-T operation is about 100 mremlh at vacuum vessel
flanges, permitting limited maintenance and access to somemachine areas.
In summary, the tritium processing systems are operatingsafely and are supporting the TFTR experimental runschedule. Operation and' routine maintenance ofTFTR during
D-T have been demonstrated. Shielding measurements havedemonstrated that the number of D-T experiments will not
be limited by either direct dose from neutrons and gammas
or dose from the release of activated air or release of tritiumfrom routine operations and maintenance.
McGuIre et at.

2177

.,

:

4

10

TFTR'(DT)ITER~ 10' ------------~-------JET(DT*,~.......
,....
10.2: A • ,
Q)'
~ 10.5
• A

.. .,

a..

•

c:

•• ••

0

'00
:J
u..

10-8

10'"

....

A World
Tokamaks

•

Ohmic

•
•

RF
NBI·D
NBI·DT

*

•

1980

1990
YEAR

2000

~
I])

:;:
0

5

'iii

+
+

o

0.5

I

·• $.·t••
If
1.0

•

..1 • I·t +• •
• I
1.5

• -

• •

•

8.
<::

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Supershot
High-13p(rampdown)

l-

b
2010

,

I

§' 10

.2

u·L----'-_-L_-L_--L_....!.l.._--L_ _-.l-.'---I

1970

,

I •
I

I
I

,

I I
•

It
••

• ,1 L-mod?
2.0

2.5

3.0

Plasma current (MA)FIG. J. Progress of tokamaks in obtaining fusion power from D-D and
D- T reactions for OH. NBI. and RF-heated plasmas.

FIG. 2. Peak D- T fusion power for TFfR discharges in the supershot,high-,8p. and L-mode regimes.

III. TFTR MACHINE PERFORMANCE

which increased from typically 160 ms to a maximum of 270
ms in D-T plasmas. For the first time, the fusion performance of TFTR at the highest beam power and plasma current is not limited by plasma energy confinement, but rather
by stability near the beta limit.

TFrR experiments over the last 10 years have emphasized the optimization of high performance plasmas as well
as studies of transport in high temperature plasmas. Figure I
shows the progress of tokamaks in obtaining fusion powerfrom D-D and D-T reactions for Ohmically heated (OH),neutral beam injection (NBI), and radio frequency (RF)heated plasmas. With increasing Ti and density, the fusion
power from OH tokamaks steadily increased during the1970s. With the advent of high power NBI in 1973, the fusion power was raised substantially relative to the OH plasmas of that time. Then finally in the 1990s, with D-T on JET
and TFTR, the tokamak is producing substantial fusion
power.
In TFTR the highest performance plasmas are supershotswith peaked density profiles, which have performance, as
measured by the parameter ni(O)Ti(O)'TE, enhanced by afactor of -20 over comparable L-mode plasmas, or a factorof -5 over standard H-mode plasmas with a broad density
profile. The enhanced confinement of supers hots is correlatedwith the peaking of the density profile, ne(O)/(n e). In plasmas with constant beam power, the confinement enhancement over L mode rises to -3 as ne(O)/(ne> increases to 3. 6
An important feature of the supershot regime is that the confinement time does not decrease with heating power, in contrast to L-mode and H-mode plasmas where 'TE - P he~{2.
This feature is also evident in the local transport coefficientsfor supershots and L-modes, and suggests that the basic
mechanism causing transport is substantially modified in supershots relative to L-mode plasmas.
During the past year (1994), as a result of extensive wallconditioning with lithium pellets'? supershots have been produced at I p =2.7 MA corresponding to q",=3.8. This represents a significant extension of the supershot regime fromplasma currents of 2.0 to 2.7 MA. Typically, two Li pellets(-2 mm diameter) are injected into the plasma in the Ohmic
phase of a pulse prior to beam injection, and two Li pellets
are injected into the post-beam injection Ohmic phase inpreparation for the next discharge. Each pellet deposits approximately one monolayer of Li on the vacuum vessel firstwall. This conditioning results in an energy confinement time
2178

Phys. Plasmas, Vol. 2, No.6, June 1995

IV. FUSION POWER

TFrR has an extensive set of fusion neutron detectors(five fission detectors, two surface barrier detectors, four activation foil stations, a collimated scintillating fiber detector,8and a 10-channel neutron collimator with 25 detectors) toprovide time and space resolution as well as energy discrimination of the D-T and D-D neutron fluxes. 9 The systems
were calibrated in situ by positioning an intense D-T neutron generator source at many locations within the vacuum
vessel. In addition, the activation system is absolutely calibrated by neutronics modeling of the neutron scattering. The
yield measured by the fission, surface barrier, and 4He recoil
detectors is linear with measurements by activation foils over
six orders of magnitude. The system of multiple measurements and calibrations has allowed high accuracy, ±7%, determination of the fusion energy production. Neutronemission profiles which are peaked in the center of theplasma are measured by the neutron collimator.
As shown in Fig. 2, the highest fusion power of to.7
±0.8 MW was achieved in a supershot discharge at I p =.2.7
MA. The highest fusion power in a current rampdown(high-,Bp) experiment was 6.7 MW achieved in a 1.5 MA
discharge.
Figure 3 shows the time evolution of the D-T fusion
power from a sequence in December 1993, May 1994, andNovember 1994. leading up to the shot producing the highest
instantaneous power of 10.7 MW at 39.5 MW of input power
for an instantaneous Q of 0.27. Here Q is defined as theinstantaneous total fusion power divided by the total injected
NBI power. Shine-through, first-orbit loss, dW/dt terms, etc.,
are not subtracted from the total injected NBI power in determining Q. Each D-T fusion event is counted as producing
17.6 Me V of energy. Normally the neutral beam heating
pulse length is limited, typically to 0.7-0.8 s, to reduce neutron activation of the tokamak structure. In this sequence, theMcGuire et al.

NOV 1994

1010.0.....;;-------------,
10MW,,"'• .....--1-...
'" ........, ..
.;··"~ ~'.i_ - - - - ·x

,X ....

.... fFUSION
POWER..------.:7'"., ."
..

(MW)

.i.'

~

5

x DHe

I

I

)('

o

3.0

0.01

3.5
4.0
Time (Seconds)

I

0.0

I

I

0.2

I

I

0.4

I

I

0.6

Minor Radius (m)FIG. 3. Time evolution of the D-T fusion power from a sequence in December 1993, May and November 1994, leading up to the shot producing
the highest instantaneous power of 10.7 MW at 39.5 MW of input power for
an instantaneous Q of 0.27.

neutral beam power and the amount of lithium conditioning
were progressively increased. Only shots with tritium NBI
are shown in Fig. 3; shots with deuterium NBI only wereinterspersed between the tritium shots for conditioning of thewalls. The 10.7 MW shot in the sequence had a minor disruption after 0.5 S of NBI when exceptionally good confinement increased the plasma pressure near the beta limit. The
Troyon normalized 13, f3N(= I08f3-ra B -r/lp , where f3r is thetotal toroidal 13 and a is the plasma minor radius) reached1.8_ The parameter of relevance for fusion yield isf3"'=2/Lo(p2)1I2a(m)/[Ip (MA)·Br(T)], where (p2)112 is
the root-mean-square plasma pressure, which reaches 2.8 forthis plasma. Values of f3N=3.0 with 13"'=4.2 have beenachieved in high fusion power high-,Bp discharges in which
the current was ramped down (for current profile control purposes) from 2.5 to 1.5 MA
The measured neutron emission profiles agree well withthose calculated by TRANSP using measured plasma parameters as shown in Fig. 4. 10 The beam voltage is approximately 105 keV for the "Case shown. The beam neutrals are

C
0

.~

'EQ)
Co

-em
.....
~'E

c;:1-'0

OC
(\$

1.0

"E
0

..c
0

0.0
Radius(m)FIG. 4. Measured profiles of neutron emission compared with those calculated by TRANSP for measured plasma parameters.
Phys. Plasmas, Vol. 2, No.6, June 1995

FIG. 5. Comparison of tritium and helium particle diffusivities and convective velocities. The diffusivitiesof tritium, helium, and heat are of similar
magnitudes. These are attractive characteristics for future reactors, _like
ITER.

injected with full, half, and third energies. The fractions of
the neutral currents at full energy are 0.49 for tritium and0.43 for deuterium. The fractions at half energy are 0.38 fortritium and 0.39 for deuterium. The neutron emission is due
to beam-thermal, beam-bearn, and thermonuclear reactions.
The separation between these reactions is discussed in Ref.

11.
V. TRANSPORT ANO CONFINEMENT IN D-T

A. Tritium particle transport

Tritium operation in TFTR 12,13 has provided a unique
opportunity to study hydrogenic 'particle dynamics in reactor
relevant plasmas. The enhancement factor of """ 100 in D-T
neutron cross section, compared to that for D-D reactions,allows easy diagnosing of both trace tritium particle trans~-,·port and influx from the limiter. The study differences inparticle transport between deuterium and tritium, experiments were performed with small concentrations of tritium
prior to the walls becoming loaded with tritium. These experiments entailed the use of either deuterium containing a
trace tritium concentration «2%) or.small puffs of pure tritium gas puffing into a deuterium-bearn-heated discharge.
These experiments showed relatively rapid radial tritiumtransport such that the effective tritium particle confinement
time TpCT) is approximately equal to the energy confinement
time TE and that the tritium particle transport coefficients arecomparable to He particle transport coefficients in similar
deuterium plasmas. 14 Figure 5 shows the tritium transportcoefficients, Dr(r) and V r(r), as determined from multiple
regression analysis. In addition, the transport coefficients of
4He measured by charge-exchange recombination spectroscopy on similar plasma discharges are shown forcomparison. 15 Also included in the plot is the deuterium thermal conductivity determined from eqUilibrium power balance analysis. Thediffusivities are all similar in magnitude
and profile shape: Dr~DHe-XD' The similarity of the diffusivities has been observed in previous perturbative transport experiments on TFTR and is a prominent characteristicof transport due to drift-like microinstabilities. 15- 17 In addiMcGuire et a/.

2179

4

(j) 1.3
S~
>-

O

c:
~

3

(])

e>
c:
w

'0

~

1.2

~

Q)

"0
"0

2

-

1.1

W
\-'

.9

(/)

<II

Ip(MA) Pb(MW)X Pure D·NBI
...

w

E
<J)
<II

a::

• PureT-NBI... PUreT·NBI

0.9
0

3.5

4.0

4.5

20-22
6-18

3.0

Average jon Mass (amu)Time (sec)tion, the similarity in the diffusivities has been shown to be
attractive with regard to helium ash removal for future reactors, such as ITER. IS
In both the tritium gas puffing and in the subsequent
high power deuterium-tritium neutral beam heating experiments, spectroscopic measurements have shown that the influx of tritium from the limiters is relatively small «5%)and that the edge fueling from the limiter is predominantly
deuterium. The relatively rapid transport to the core together
with the relatively low influx of tritium from the walls affects
the ratio of n DI n T in the plasma core.

1.8
1.6

2.5

2.0

5.0

FIG. 6. Comparison of plasma stored energy in comparable D-D and D- T
discharges. The plasma stored energy is larger in the D- T plasmas. Energy
confinement time increases from 160 ms to 200 ms. The productn,(O)T,(O)TE increases from 1.9 to 3.5 (1020 m- 3 s keY).

1.6-2.0 6-30

0 D·NBI + T·NBI 1.8-2.0 20-30\-'

FIG. 7. The energy confinement time in these supershot discharges increased with the average mass of the hydrogenic ions. This is observed insupershot and H-mode regimes.

MW of tritium and 8 MW of deuterium NBI are shown in
Fig. 8. There is a 20%-30% increase in Ti(O) and only a5%-10% increase in the ne(O) going from D-D to D-T
plasmas.
There are a number of expected differences between Tand D-neutral-beam heating, which are modeled using the
SNAP and TRANSP codes. For T-NBI, the beam deposition
profile is broadened, the beam heating of thermal ions isincreased, and the heating of electrons is decreased. The
fusion-generated alpha particles are expected to primarilyheat the electrons. Taken together, these effects tend to cancel, producing small net changes in the total ion or electron
heating powers when changing from D- to T-NBI in the plas-

B. Isotope effects in supershots

The experiments performed in December 1993 and May
1994 provided a clear demonstration that the plasma confinement in D-T supershots is better than in similar D-only plasmas, as shown in Fig. 6. The plasma energy is determinedfrom magnetic data and includes the energy in the unthermalized injected deuterons and tritons. The isotopic content
of the plasma must be measured in order to understand any
transport variations observed. In D-neutral-beam-heatedplasmas, the absence of a significant tritium density is confirmed by the low level of D-T fusion neutron emission. In
T-neutral-beam-heated plasmas, the wall recycling deuterium
influx is a significant source of particles and leads to a significant deuterium density nd throughout the plasma, as measured and discussed in Ref. 18. These plasmas were generated using co- and counter-tangential neutral beam injection(15-30 MW) into low edge-recycling plasmas, with plasma
currents of 1.6-2.5 MA. The stored plasma energy, electron,and ion temperatures increased in deuterium-tritium plasmas
compared with similar deuterium plasmas, corresponding toan increase in T£ from 160 ms to 200 ms and in the product2o2onj(O)Tj(O)TE from 1.9X10 to 3.5X10 m- 3 keY s. Theenergy confinement time in these supershot discharges increased with the average mass of the hydrogenic ions asshown in Fig. 7. This improvement in thermal energy confinement with ion mass is observed for both supershots lo.ls
and limiter H modes l9 in TFfR. The ion temperature andelectron density profiles for an I p = 1.6 MA plasma with 8
2180

Phys. Plasmas, Vol. 2, No.6, June 1995

40

4

30

Tj

Tj

! ! ! !
ii•&(keV) 20t&

18 MW, 1.6 MA

10

•
A

• T-NBI ( D:T", 50:50)
A D-NBI

...•

.

Al

0
6
5

4

ne
(10 19 m·3)

•

•

t

1 •A •

ne

A

•

A

•

A

3

0

•

A

•

A

2

magnetic
axis

t 1

t 1

edge~ ..

Dllr
2:5

•*••

3.0

3.5

I\t1ajor RacfIUS (m)FIG. 8. Ion temperature and electron density profiles for an lp= 1.6 MA
plasma with 8 MA of tritium and 8 MW of deuterium NBI. There is a20%-30% increase in Ti(O) and a 5%-10% increase in the n,(O) goingfrom D-D to D- T plasmas.
McGuire et al.

7r---------------------------------~

D-T(pure T-NBI => ..s0:50
thermal D:T mix);

'I·

PNS' = 18 MW
Ip = 1.6MA
R = 2.52m

0.0

0.5
Normalized minor radius (rIa)

1.0

FIG. 9. Ion thermal conductivity is reduced by !Ifactor of -2 in D:':'t
plasma compared to D-D plasmas. The improvement in XI increases withplasma tritium content.
.'

mas studied. The power balance analy~Js indicates that th~higher Ti gradient measured during a 50150 D-T plasma
relative to a pure D plasma is due to a reduction of the ion
thermal diffusivity X:of by a factor of 2'f~r r/a~O.5 (Fig. 9).
The lack of substantial change in the density gradient, de7
spite the broader beam deposition m:ofile with T-NBI, indicates a drop in the core electron particle diffusivity D by

-30%.
The~limiter H modes produced
on
TFTR
in
high~j3p D-T. ' .
plasmas- have energy confinement enhancements >4 relative to the ITER-89P scaling20 wIMle ,~<;)ffespoiiding D. pla~~
mas had enhancements of -3.2. The confinement was im7
proved across the plasma duriJ;lg the H~mode phase. In
particular, the ion heat conductivity was observed to de~~easeby a factor of 2,3 across the transition to H mode 17 (Fig.
10). The edge localized modes (ELM's) are much larger during the D-T H modes. This suggests that ITER D-Tplasmas
may be more susceptible'to gi!lnt EUvi'Sihan inferred from
D-only experiments. The power threshold for the transition.
to an H mode is.similar to D and D-T discharges. 21
One focus of the present experimental campaign is theturbulence and transport characteristics of D-T _plasmas
which have indicated improved ion confinement properties
during D-T operation. Initial results from the reflectometer
indicate that there appears to be no difference in the local
iiln in D-T plasma compared to similar D-D cases. (Fig~11). An extensive study of isotope scaling effects on confinement and fluctuations is planned for the. near future.

2..9

3.0

~.1

Time (sec)
FIG. 10. Comparison of H-ma'cte transitions in D:-D and D-T plasmas. The
increase in TE is larger in D- T plasmas. The edge localized modes are larger
in D-T plasmas.

tion .of classical effects associated with high energy particle
orbits in the inhomogeneous magnetic field, and instabilities
in the plasma resulting in a loss of alpha particles. The operating point for a reactor is determined in part by the confinement of alpha particles, the transfer of energy from thealphas to the background plasma, and the accumulation of
low energy alpha ash in the plasma which displaces tIle deuterium and tritium ions. TFTR experiments are aimed atstudying this broad range of alpha particle physics and documenting them for conditions relevant to the reactor regime.

w~------------------

VI. a CONFINEMENT AND a HEATING

A. Effects of alpha particles
The behavior of alpha particles from D-T reactions is afundamental consideration for the performance of a future
D-T reactor for two reasons. First, if a significant fractioIi of
the alpha particles is not confined, then the confinement requirements for ignition would increase. Second, if a small
unanticipated fraction (a few percent) of the alphaparticles islost in ITER. and the resulting heat flux is localized, damage
to first-wall components could result. The heat load on thevessel components from alpha particles is due to a combinaPhys. Plasmas, Vol. 2, No.6, June 1995

FIG. 11. Initial results for the refiectometer indicate that there appears to be
no difference in the local iiln in D-T plasmas compared to similar D-D
cases.
McGuire et al.

2181

c£'

Jf' ~~h

b 30
C
C
o

90 Q detector

T"""

~

u

~

norm.

.m 10

§ 8

t5 6
~

'8 4

-'~

First-orbit /loss model

«l

.c

0~

in loss is small «50% of the first orbit loss which corresponds to =<3% of the total alpha birth rate) but clearlyvisible on the probes. The same effect is also seen in D-D
plasmas, for D-D fusion products. The present understanding is that the ICRF, which primarily increases the V.L of theresonant particles, heats the alpha particles and a part of the
population crosses the passing-trapped boundary and entersthe first-orbit loss cone, resulting in increased loss.23
C. Alpha heatingr~

2+----r----r----r----r---~--__+

o

0.5

1.5

1

2

2.5

3

Plasma current (MA)FIG. 12. The plasma current dependence of the neutron-normalized total
D- T alpha loss signals. The agreement between the calculated and mea-

sured alpha loss versus plasma current is within the estimated uncertainties
in the calculation.

B. Single-particle effects
An extensive study of fusion product losses in deuterium
experiments had been conducted prior to beginning D-Texperiments. 22 During the D-T experiments the scintillator
probes located at 90°, 60°, 45°, and 20° below the outer
midplane detect alpha particle losses. The results from the
90° detector during D-T (shown in Fig. 12) match the firstorbit loss model in both magnitude and pitch angle distribution. For detectors closer to the midplane, the first orbit loss
model does not adequately fit the losses from D-D or D-T
plasmas. Collisional and stochastic toroidal field ripple
losses are being investigated to explain the pitch angle distribution observed there.
The probes are also used to study the effect of ICRF onenergetic particles. In a deuterium-tritium plasma, the alphaparticle losses are observed to increase with the applicationof ICRF as shown in Fig. 13. The magnitude of the increase

The electron heating in D-T supershot plasmas has been
analyzed for evidence of heating by fusion-produced alphaparticles. During the NB-heated phase of the discharge, thealpha heating contributes ~ 1 MW out of ~ 10 MW of heating power to the electrons, making its detection difficult. The
first method of detecting alpha heating is to analyze andsimulate the steady-state power balance of the electrons.
Simulations using the measured plasma parameters (exceptTe ), and the Xe experimentally inferred from the deuterium
comparison discharge indicate that alpha heating may be responsible for about half of the observed 2 keV increase in Te
going from D to D-T plasmas. The second method is to
examine the transient response of the electrons to a SUdden
change in the alpha heating. In a pair of nominally identical
D and D-T plasmas, a lithium or boron pellet was injected~0.2 s after the termination of NBI. The initial density increase and Te decrease upon injection of the pellets were
nearly identical in the two cases. By the time the pellet injection, most of the circulating beam ions have thermalized,
but the alpha particles have not due to their longer slowing
down time. In addition, much of the tritium in the plasma is
calculated to have been pumped out of the plasma by theconditioned graphite limiter. The Te reheat rate after pellet
injection for the condition discussed above is measured to be~85% higher in the D-T plasma relative to the D plasma, in
agreement with TRANSP calculations of the expected alphaheating of the electrons. Additional experiments at higheralpha particle densities and pressures are planned.
VII. CONFINED ALPHA MEASUREMENTS

5

••

o·-~~
3.5

~

4.0

TIME (sec)

FIG. 13. (a) Neutron-normalized alpha loss rate to a detector 90° below the
midplane as a function of time and (b) the corresponding RF power evolution. 9.1 MW of D and 11.6 MW of T neutral beam power were injectedduring the time indicated by the shaded region.
2182

Phys. Plasmas, Vol. 2, No.6, June 1995

The first experimental results have been obtained with
two of the new alpha particle diagnostics of confined alphasfrom D-T reactions. The alpha charge-exchange diagnostic
obtained data during ablation of a Li pellet fired into a 1.0
MA D-T plasma, after the neutral beams were turned off. In
the case shown in Fig. 14, the different analyzer energy
channels give an energy spectrum of the alphas in the plasmacore in the range 2 MeV down to 0.5 MeV. The measured
shape of the energy spectrum of the alphas is in good agreement with a TRANSP calculation, although an absolute calibration of the diagnostic is not yet available. Chargeexchange recombination spectroscopy has been used tomeasure the alpha particles with energies up to 600 keV ina
D-T pulse soon after the T-beams have been turned off, butwith D beams remaining on to allow the measurement. The
signal predicted from the alpha distribution function calculated by the TRANSP code is within a factor of 2 agreementwith the measured absolute intensity, demonstrating that this
McGuire et al.

105~--------~--------~PELLET CHARGE EXCHANGE
PARTICLE DETECTION

(MJ)

0.15
(sec)

Data normalized
toTRANSP

0.1q=~~~*=:=::::;::~=:;;~w~==:=i40

Li PELLET 200 ms AFTER NBI

a.u 20r=Ocm

103+-~~~~~--~~~~
0.2

0.7

1.2

1.7

Time (sec)

2.2

Alpha Energy (MeV)
BO. 14. Energy spectrum of confined alpha particles measured by the alphacharge exchange diagnostic at r=O cm is compared with TRANSP calculation.

technique can be used to make absolute measurements of thealpha density. Further work is in progress to' evaluate theeffects to stochastic ripple diffusion and sawtooth oscillations on the alpha energy and radial distributions and to compare them quantitatively with theory.
VIII. a-ASH ACCUMULATION

3.0

4.0

3.5

Time (sec)The production, transport, and removal of helium ash is
an issue that has a large, impact in determining the size andcost of ITER. 24 The present experiments on TFTR are providing the first opportunity to measure helium ash buildup,
assess helium transport coefficients, and examine the effects
of edge helium pumping on central ash densities in D-T
plasmas. In addition, the, importance of the -central helium
source in determining the helium profile' shape and amplitude
is being examined.
Initial measurements of radial ash profiles have been
made using charge-exchange· recombination spectroscopy.
Differences between similar D-D and D-T supershots in thetime history and amplitude of the thermal helium spectrum
enables the alpha ash profile to be deduced. These measurements have been compared to predictions from the TRANSP
code, using transport coefficients from earlier helium puffing
experiments in deuterium plasmas and the TRANSP calculation of alpha particle slowing down and transport upon thermalization. The ash profiles are consistent with the TRANSP
modeling, indicating that the ash readily transports from thecentral source region to the plasma edg~ and recycles. These
measurements provide' evidence that, in the presence of acentral helium ash source, the ash transport and confinement
time are roughly consistent with external helium gas puffing
measurements. This suggests that helium transport in theplasma core will not be a fundamental limiting factor forhelium exhaust in a reactor with supershot-like transport.
Further dedicated experiments will be performed to determine the alpha ash particle transport coefficients in D-T
plasmas.
Phys. Plasmas, Vol. 2, No.6, June 1995

FIO. 15. The effects of MHD on confinement suggests that the MHD can be
responsible for up to a 30% decrease in the energy confinement time in theworst cases. In cases of weak MHD, typical of most of the higher current
plasma (Ip>2 MA, qs/o<4), the effect is usually less than 5%.

IX. MHO STABILITY IN 0-T PLASMAS
A. MHO activity in the initial TFTR 0-T plasmas

Low m and n (min = 211 , 3/2, 111, etc.) coherent MHD
modes have been observed in the initial D-T plasmas on
TFfR. The amplitude, frequency of occurrence, and effecton plasma performance are similar to those observed in comparison D-only plasmas. Modeling of the effect of MHD on
confinement suggests that the MHD can be responsible forup to a '30% decrease in the energy confinement time in theworst cases,25 consistent with the observations. Incases of
weak MHD, typical of most of the higher current plasmas(Ip>2.0 MA, qsh<4), the effect is usually less than 5% (Fig.
15). The decrease in the neutron rate is consistent with thechanges in the equilibrium plasma: it is not necessary to
invoke anomalous losses of fast beam ions to explain thisdecrease. Enhanced losses of fusion a's, correlated with thepresence of MHD, are observed in D-T plasmas. The losses
are similar to those previously reported for D-D plasmas,26
and represents a small fraction of the total alpha population.
Fishbone and sawtooth activity have also been observedin D.:... T plasmas. At present there is no evidence that thefusion a's have affected to sawtooth or fishbone stability.
There is a tendency for the fishbone activity to be stronger in
McGuire et al.

2183

11

Ballooning mode
bursts ~

10

~

"""
l-

Q)

7

-en-E
::>

3.4

0

3.0

a:

2.8

«
a:

Ballooning
mode

3.2
5keV

0
J

«

Ballooning
mode

Magnetic Axis

2.6

~

./\

2.4 _

1

180 f..lsec-II-

180 f..lsec-I

FIG. 16. Contours of the electron temperature prior to a high-f;I disruption showing the n= 1 kink and ballooning precursors.

D-T plasmas; however, that may be more correlated with thesomewhat broader pressure profiles often found in D-T plasmas, as compared to D-only plasmas under similar conditions.
B. fJ limit and disruptions in 0-T plasmas

Currently, the D-T fusion power which TFrR can produce is limited by pressure-driven instabilities which cancause major or minor disruptions. The disruptive f3 limit in
D-only NBI-heated plasmas and D-T NBI-heated plasmas
appears to be similar. The f3 limit follows approximately thedependence on plasma current and magnetic field predictedin the Troyon formula. 27 The high f3 disruption in D-only or
D-T plasmas appears to be the result of a combination of ann= 1 internal kink coupled to an external kink mode and atoroidally and poloidally localized ballooning mode. 28 Figure
16 shows contour plots of the electron temperature measured
at a 500 kHz sampling rate by the two electron cyclotron
emission (ECE) grating polychromators (GPC's) separated
by 126 0 in the toroidal direction. The ballooning character ofthis mode is observed as a poloidal asymmetry on the magnetic loops signals, the signal is five times larger on theoutside than the inside. The simultaneous presence of theballooning mode on one GPC , and its absence on the second
clearly demonstrates the toroidal localization of the mode.
The ratio of the frequency of the ballooning mode and then = 1 kink indicates that the ballooning mode has a toroidal
wave number of about 10-15 (assuming only toroidal rotation). The radial structure of the kink mode suggests coupling of a predominantly internal kink to a weaker external
kink. While PEsr9 predicts that the n= 1 kink is unstable forthis disrupting plasma, it also in general predicts that most2184

Phys. Plasmas, Vol. 2, No.6, June 1995

supershot plasmas are similarly unstable, as q(O) is typicallyless than unity30 and the plasma pressure is sufficient to drive
an ideal mode.
The kink mode can locally decrease the magnetic shear
and increase the local pressure gradient so that the ballooning mode is locally destabilized. The thermal quench phase
may result from destruction of flux surfaces by the nonlinear
growth of the n= 1 kink, possibly aided by the presence ofthe ballooning modes. There is no evidence for a global magnetic reconnection as is seen in high density disruptions. The
electron temperature collapses on a time scale of several
hundred microseconds with no local flat spots, indicating thatthe magnetic geometry is destroyed uniformly over theplasma cross section. The thermal quench phase is typically
preceded by a large nonthermal ECE burst. The burst is atleast 10 to 20 times larger in amplitude than is predicted by
the fast compression of electrons by a rapidly growing internal kink displacement.
In both D and D-T experiments, MHD activity with low
toroidal and poloidal mode numbers is observed to increase
the loss of fusion products. Both minor and major disruptions produce substantial losses of alpha particles. In a major
disruption, ~20% of the alpha stored energy is observed to
be lost in ~2 ms during the thermal quench phase, while theplasma current is still unchanged. The loss is preferentially tothe bottom of the vessel to the 900 detector only (90° withrespect to the midplane), which is in the ion VB-drift direction, as opposed to locations such as 20°, 45°, or 60° below
the midplane where the other alpha particle detectors arelocated. The design of in-vessel components in a reactor will
have to accommodate the localized heat flux from alpha particles during a disruption.
McGuire et al.

,'p=2.0 MA
90 Q detecto rPfusion = 7.5

2.0

/
~

N

J:

~

"!" 1.0
0

0'1

2
345
6
Peak Fusion Power (MW)7

FIG. 17. Alpha loss does not increase with fusion power on TFTR during
D-T. The variation of lost alpha fraction with fusion power is consistent
with the first-orbit loss model.

-....

Idf

0.0 '----'-"---"'----"'----"'----~

o

100

200

300

400

500

Frequency (kHz)C. Toroidal Alfven eigenmodes studies
Experiments on TFfR31 and DIII_D32 have demonstrated

that it is possible to dest~bilize the toroidal Alfven eigenmode (TAE) with neutral beams and ICRF tail ions. In bothcases, there is some loss of energetic beam particles and tailparticles. Two of the most important physics questions arewhether alpha-induced instabilities are present and where thepredicted thresholds are in agreement with the experiment.
The highest fusion power shots on TFfR have produced
fast a populations with some dimensionless alpha P3J~m­
eters, such as RV f3lX' which are comparable to those for the
projected fast a populations for ITER. In typical TFfR D:-T
supershots, the thermal and beam ion Landau damping arestronger than the fusion a drive for TAE modes. Experiments
were done successfully to reduce the thermal ion Landau
damping; however, the a drive was still not sufficient toovercome the beam ion Landau damping.33,34
At fusion power levels of 7.5 MW, fluctuations at thetoroidal Alfven eigenmode frequency were observed withmagnetic diagnostics to increase. However, no additional alpha loss due to the fluctuations was observed. Figure 17
shows that the fraction of alpha particles that are lost is independent of the fusion power, indicating that additional loss
does not occur at high power up to 9.3 MW.
'
The threshold for instability is determined by a balance
between drive and damping terms.·Recent experiments haveinvestigated modifying the relationship to test the theory
quantitatively. For TFfR parameters, electron and ion Landau damping can be important. In one series of experiments
at relatively high fusion power (5 MW), the ion temperature
was suddenly decreased by employing a He gas puff, or injection of a Li or D2 pellet. This rapidly decreased the central
ion temperature from 22 keY to 6 keY. Despite the change inelectron and ion Landau damping, the mode was not destabilized. A more detailed analysis is in progress to compare
theory and experiment.
Experimentally the search for a-driven TAE activity in
D-T plasmas has been complicated by the presence of amode near the expected TAE frequency in both D-D and
D-T NBI-heated plasmas. This mode has a relatively broad
peak in frequency, with a spectral width of about 50 kHz at
Phys. Plasmas, Vol. 2, No.6, June 1995

FIG. 18. Spectrum of magnetic fluctuation for D-T plasmas generating 7.5
MW and 6.4 MW of fusion power and a D-only plasma.

300 kHz. This mode may represent a "thermal" level of
excitation or be driven by fast beam ions. For these plasmas
the beam ion velocity is one-third to one-fifth the Alfven
velocity. 35
In Fig. 18 is shown the spectrum of the edge magnetic
fluctuations for a D-T'shot with 7.5 MW'of fusion power
and for a similar shofat 6.5 MW and a D-only shot. The
mode amplitude has increased by a factor of 2-3 in the 7.5
MW shot. The NOVA-K code36 finds n=5 and n=6 corelocalized TAE activity in the region where q<l in thisphisma. 37 The localization of the mode near the plasma core
increases the coupling of the fusion a's which makes themode unstable. The calculated TAE mode frequency fromthe NOVA code was about 250 kHz, lower than the experimental frequency of 300 kHz. In this experiment the toroidal
mode number was not measured.

x. ICRF HEATING IN 0-T
In preparation for D-T operations, the TFfR ion cyclotron range of frequencies (ICRF) heating system has been
upgraded. The positions of the antennas' can be controlled
remotely to maximize coupling to the plasma in' different
regimes. Phasing of the antennas at 0°, 180°, and 90° has
been establishedifl both deuterium majority and 4He plasmas
to allow for both heating and current drive studies. The antennas have operated successfully during D-T plasmas. The
increased radiation field- from D-T neutrons, as well as the f3
decay from tritium, has not affected antenna performance.
ICRF wave physics in deuterium-tritium plasmas iscomplicated by the possibility of mUltiple, spatially separated resonances and by alpha damping which can competewith electron absorption in' the fast wave current drive regime. A promising scenario for heating D-T plasmas is fast
wave absorption at the second harmonic of the tritium cyclotron frequency, which is degenerate with the 3He fundamental. By selectively heating a majority ion species rather thana minority ion species, potential difficulties with instabilities
McGuire et af.

2185

40 r-r-'-";=====;"I

a:
w

~

a.
u.

a:

5
a.
2S
u.

o

:z

o
o
«
a:
u.
i=

TRITIUM BEAM POWER FRACTIONo~~~~~~~~

2.5

3.0

3.5

MAJOR RADIUS (m)FIG. 19. Comparison of (a) ion and (b) electron temperature profiles for two
D-T plasmas. The discharge indicated by the bold solid line had 5.5 MW oflCRF heating.

(e.g., TAE modes) excited by fast ions may be avoided.
Though the core damping is predicted to be acceptable, offaxis absorption near the deuterium fundamental and n 2 = S
layers can compete with the second harmonic tritium coredamping in tokamaks with moderate aspect ratio. The TFfR
supershot plasmas, with the second harmonic tritium heating(20T) layer coincident with the Shafranov-shifted axis at
2.82 m, the second harmonic deuterium and fundamental hydrogen heating (20ofOH) layer is out of the plasma on the
low field side, but the fundamental deuterium heating (OD)
layer is in the plasma on the high field side at R-2.1 m.
Experiments have been performed utilizing combined
ICRF heating and neutral beam injection in deuteriumtritium plasmas. The initial experiments have focused on the
RF physics associated with D-T plasmas. Second harmonictritium heating with a 2% 3He minority at a power of -5.5
MW in a plasma with 23.5 MW of neutral beam injection(60% in T) has resulted in an increase of the ion temperaturefrom 26 to 36 ke V. The electron temperature increased from
8.5 to 10.5 ke V due to direct electron damping and 3He minority tail heating (Fig. 19). Similar heating was measured indischarges in which no 3He was added. These favorable results indicate that ICRF can be used to heat a D-T plasma
with core second harmonic tritium damping. The observed
second harmonic tritium damping is consistent with 2-D
code predictions (Fig. 20), but the observed off-axis fundamental D damping is much less than the 2-D code estimate.
Further analysis of the power deposition in both deuterium
and deuterium-tritium experiments is in progress.
Experiments using a 3He, 4He, D Ohmic target plasma
have shown localized electron heating near the mode conversion surface where excitation of the ion Bernstein wave(lBW) is predicted. Localized off-axis power deposition centered at rla<O.25 with half-width Arla<O. 17 has also
been observed. Up to 75% of the radio frequency power isobserved to be deposited on electrons. Numerical modeling2186

Phys. Plasmas, Vol. 2, No.6, June 1995

FIG. 20. Measured power fraction to ions and electrons (P. + Pi) and toelectrons only (Pe) are shown by the solid lines plotted against tritium beam
power fraction. Data is compared to RF code calculations shown by thedashed lines.

indicates that 74%-91 % of the power should be damped onelectrons,. with the remainder damped on ions. 38
Majeski et a1. 39 have suggested that this mode-converted
ion Bernstein wave excited at the n 2 = S mode conversion
layer in a multiple ion species plasma (such as D-T) could
be used for electron heating or to drive localized electron
currents. Preliminary experiments to investigate mode conversion current drive (MCCD) in 3He_ 4He plasmas and fast
wave current drive (FWCD) in 4He (H) plasmas have begun.
Initial results from the MCCD experiments show a strong
dependence of the electron response on the direction of thelaunched waves. In this experiment, comparisons were madebetween similar shots with the waves launched either parallelor antiparallel to the plasma current. For these comparison
shots the loop voltage, electron density, and visible bremsstrahlung were identical. In contrast, for the FWCD case, theloop voltage depends on the direction of the RF wave propagation. The change in loop voltage during FWCD corresponds to =40 kA of RF-driven currents. FWCD or MCCD
techniques could play a crucial role in advanced tokamak
regimes for TPX and ITER.
XI. ADVANCED TOKAMAK OPERATION IN D-T

Experiments on TFfR have demonstrated the benefits ofadvanced tokamak regimes characterized by a strongly
peaked electron density profile, n e (0)/(n e )",-;3.4 (central
density divided by the volume averaged density), and bylarge ratios of TEITt; mode) ",-; 4.5 along with T/Te <4, re~sulting in high fusion power. As shown in Fig. 2, the 2.0 MA
L-mode discharges produce much lower fusion power thancomparable supers hots or high-!3p discharges. Both the supershot and high-!3p regimes have large bootstrap current
fractions which reduce the requirements for current drive, anattractive feature for reactors. Values ranging up to 70%80% have already been achieved in deuterium plasmas on
TFTR. To take full advantage of these advanced-tokamak
features calls for control of the current and pressure profiles
so as to optimize the plasma reactivity while also maintainMcGuire et al.

ing plasma stability. The current profile was modified in boththe high-,Bp experiments and in experiments that produce a
region of reversed shear near the plasma core, res1J.lting infavorable global and central confinement. These e-xperiments
also result in increased values of q(O), which could destabilize the TAE in D-T experiments;40 however, at sufficiently
high values of f3 the TAE instabilities are expected to become stable again. 34 Thus, the path to a desirable D-T tokamak reactor regime may turn out to be rather complex; notonly must the plasma maintain overall stability but also stability against TAE modes.
XII. SUMMARY

Tritium and neutron activation are being handled safely,
thereby allowing TFTR to operate a full schedule and to be
maintained routinely. TFTR presently plans to run most of
FY95, based on present funding, and expects to stop operation in September 1 9 9 5 . "The D-T experiments on TFTR have allowed the dire~t
examination of many critical issued of physics and technology for ITER, and has clarified what new technical features
could be most helpful to an advanced tokamak reactor. The
confinement in a D-T plasma in TFTR is better than in a
D-D plasma, allowing significantly improved nrT fusion
performance in D-T. The alpha particles are found to be
confined as expected and hints of alpha heating have alfeady
been seen. ICRF heating of both electrons and ions has been
demonstrated at modest ICRF power levels. Initial measurements of alpha ash suggest that helium transport in theplasma core will not be a fundamental limiting factor forhelium exhaust in a reactor with supershot-like transport.
The high power deuterium and tritium experiments on TFTR
have provided the first look at the effect on MHD activity ofa fusion a population similar to that expected on ITER. In
the TFTR D-T experiments to date there is no evidence fora loss due to a-driven instabilities. Theoretical calculations
suggest that presently achieved a parameters in TFTR areclose to the stability threshold for a-driven TAE modes. The
limiter H modes in D-T plasmas have better confinement
than those in D, but have larger ELM's. The presence ofmore virulent ELM's may increase the challenge of plasmawall interactions and divertor problems on ITER.
Since the f3 limit is now providing a fundamental limi-·tation on the achievable D-T performance in TFTR, the toroidal magnetic field was recently increased from 5.2 T to
5.6 T and a further increase to 6.0 T is planned for a limited
number of pulses. This allows stable operation at highplasma current. Such an increase could raise the central
plasma energy density at the ,B limit by up to 30% and, if thepresent scaling is maintained, the achievable D-T fusion
power by up to 60%.
ACKNOWLEDGMENTS

The authors express their appreciation to the technical,
engineering, and scientific staff who enabled the execution ofthese important D-T experiments, to our collaborators on
whose expertise we relied, and to Dr. R. C. Davidson and Dr.
P. Rutherford for their support and encouragement.
Phys. Plasmas, Vol. 2, No.6, June 1995

This work was supported by U. S. Department of Energy
Contract No. DE-AC02-76-CH03073.
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Proceedings of the 13th International Conference, Washington, DC, 1990
(International Atomic Energy Agency, Vienna, 1991), VoL 3, p. 239.
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