Report of the 2005 FESAC Facilities Panel, Vol. 1:

**Characteristics and Contributions of the Three Major United States**

**Toroidal Magnetic Fusion Facilities**

Report of the Fusion Energy Sciences Advisory Committee

*August, 2005*

## 

#### Facilities Panel Members

> JILL DAHLBURG, Naval Research Laboratory (Chair)
>
> STEVEN L. ALLEN, Lawrence Livermore National Laboratory (Vice-Chair)
>
> RICCARDO BETTI, University of Rochester

##### STEPHEN KNOWLTON, Auburn University

> RAJESH MAINGI, Oak Ridge National Laboratory
>
> GERALD A. NAVRATIL, Columbia University
>
> STEVEN A. SABBAGH, Columbia University
>
> JOHN SHEFFIELD, University of Tennessee
>
> JAMES W. VAN DAM, University of Texas at Austin
>
> DENNIS WHYTE, University of Wisconsin at Madison

Executive Summary

The three major toroidal fusion research facilities in the U.S. have
diverse and complementary characteristics, which were developed on the
basis of evolving U.S. innovation in fusion energy sciences. Taken
together these three facilities provide the U.S. with a very effective
presence in the world program of fusion research. Their success has
enabled the U.S. to have substantial impact on the direction and
progress of the field, including world leadership in understanding
fundamental transport processes in magnetically confined plasmas, and
continuing optimization of the magnetic configuration for confinement of
high pressure plasmas.

# *Unique and complementary characteristics*

The characteristics that make each of these toroidal facilities unique
in research capability stem from their initial research motivations:

> C-Mod to understand plasma behavior at very high magnetic field with
> the plasma pressure and field appropriate for sustaining a burning
> plasma for energy production, but at a smaller size than a self-heated
> plasma.
>
> DIII-D to understand and improve plasma confinement and stability as a
> function of plasma shape and magnetic field distribution in a
> collisionless plasma.
>
> NSTX (the most recently commissioned, in 1999) to apply advances in
> understanding of magnetic field configuration to optimize both plasma
> stability and confinement in a proof-of-principle, low aspect ratio
> toroidal experiment (the spherical torus).

One standard for judging progress in fusion is the "triple product" of
plasma density, temperature, and confinement time. This may be rephrased
as the product of plasma β, the square of the magnetic field, and
confinement time. (The dimensionless quantity β is the ratio of the
plasma pressure to the magnetic field pressure.) In this rephrased form,
the triple product delineates the three approaches to fusion foci that
are embodied in the three major facilities of the U.S. Specifically,
Alcator C-Mod is the highest magnetic field, diverted tokamak in the
world; DIII-D pursues the direction of high confinement through the
development of advanced tokamak operational regimes with long
confinement times; and NSTX pursues the direction of very high β,
resulting from its extreme toroidicity (i.e., low aspect ratio). These
are three distinct dimensions that span the operational parameter space
for toroidally confined plasmas. Moreover, while these three facilities
are distinctly different in some regards, they have a valuable degree of
commonality that permits them to compare plasma behavior under similar
dimensionless --- but different dimensional --- parameters (e.g.,
magnetic field strengths from 0.3 T up to 8 T, all at high performance).

Today, each facility -- with its unique capabilities and program of
planned research -- is a leading element of the world program in
magnetic fusion research:

> C-Mod (with FY05 annual budget of \$22,038,000 U.S.) is one of two
> tokamaks tied with world\'s highest tokamak magnetic field and is the
> only facility capable of studying plasma/ wall interactions and radio
> frequency heating and current drive in ITER-like geometry with a
> divertor and magnetic field and plasma pressure characteristic of a
> burning plasma.
>
> DIII-D (with FY05 annual budget of \$55,751,000 U.S.), with its
> unparalleled plasma transport diagnostics and world-leading capability
> for plasma shaping and control of major instabilities limiting high
> pressure plasmas, has established itself as a center for developing
> long-pulse, high performance advanced tokamak operation.
>
> NSTX (with FY05 annual budget of \$34,584,000 U.S.) is the world's
> most capable spherical torus, exploring high-β plasma stability and
> confinement at extreme toroidicity (low aspect ratio) and is the major
> U.S. experiment for concept innovation in magnetic confinement now in
> operation.

While these three facilities are clearly distinct, they also have a
degree of commonality that makes them highly effective as a group. The
*commonality* is useful to allow comparisons, while the *distinctive
features* of these facilities allow extrapolations. This combination
provides the U.S. with a world leadership role in advancing the most
important frontier areas in fusion energy sciences research, recently
described in detail in the April 2005 FESAC Program Priorities Report.

Some of the most important examples of forefront research led by one or
more of these facilities are:

[Multiscale Plasma Transport:]{.underline} The U.S. is one of the
leaders of the international research effort to determine the underlying
instabilities responsible for plasma transport. The U.S. transport
program is made strong by DIII-D, which has an integrated program with
measurements over the important wavelength ranges of the turbulent
instabilities responsible for the observed transport, as well as the
most comprehensive gyrokinetic transport model in the world; and by
C-Mod and NSTX, which also contribute unique elements to this line of
research.

[Plasma-Boundary Interfaces:]{.underline} C-Mod and DIII-D have led this
research effort in the U.S. The U.S. facilities, taken together with
computational modeling, provide a world-leading capability in
plasma-boundary science. While it was a U.S. physics insight that
identified the controlling role of the plasma parameters at the top of
the H-mode pedestal, there yet exists no first-principles model to
predict the pedestal parameters for ITER. The U.S. is leading an effort
to understand pedestal physics by means of comprehensive and
complementary diagnostics and models applied across all three
facilities.

[Optimization of Magnetic Confinement:]{.underline} NSTX was designed
and built to lead the world effort to apply our improved knowledge of
magnetic field topology to a new toroidal configuration at very low
aspect ratio (the spherical torus). Enhanced shaping has been predicted
to increase global plasma stability at high plasma β, and NSTX quickly
demonstrated record levels of plasma β. It provides a unique set of
parameters complementary to the higher aspect ratios in C-Mod and
DIII-D, valuable for extending parametric studies in transport, MHD
stability, and boundary physics.

# *U.S. toroidal research as a whole*

Scientific understanding of magnetic fusion plasmas comes from a
combination of theoretical and experimental investigation. Crucial
experimental verification of theory requires facilities that can vary
plasma conditions over an extended range. No single facility can
adequately span all desired variations. The U.S. fusion program has a
complementary set of three major experimental facilities that are able
to access a very wide range of variations. The healthy synergy of joint
experiments conducted between two or more of these facilities,
worldwide, contributes to exploiting their full potential to test theory
and techniques. With their unique machine capabilities and outstanding
diagnostics, all three major U.S. facilities are carrying out
world-leading magnetic fusion plasma research.

The world program in fusion research includes major toroidal facilities
in Europe (JET, ASDEX-U, TCV, FTU, TEXTOR, Tore Supra, and the spherical
torus device MAST) and Japan (JT-60U and the superconducting stellarator
LHD). The superconducting tokamaks -- K-STAR (Korea), and China's EAST
that should have first plasma in early 2006 -- will make important
contributions to issues of long-pulse operation. The JET and JT-60U
tokamaks are each larger than any of the three major U.S. facilities and
so can perform studies closer to ITER and fusion power plant parameters;
taken together, however, the three U.S. facilities make influential
contributions to the larger world effort in key areas such as plasma
transport, MHD stability, plasma-boundary science, radio frequency
heating and current drive, and energetic particle physics. Among
spherical tori, NSTX has exceptional performance and diagnostic
capabilities. DIII-D and C-Mod use their unique and complementary
capabilities to influence and expand the world research agenda in the
areas of transport, advanced tokamak physics, and divertor development.
Backed up by the substantial U.S. effort in plasma computational
modeling, the research programs planned for the next five years on the
three major U.S. facilities emphasize these areas of strength and
promise a continued major role as the worldwide fusion effort moves
towards ITER experimental operation. Equally important, the three
experiments keep the U.S. at the forefront of fusion energy sciences --
transport, magnetohydrodynamics, etc -- which is of significance beyond
the immediate needs of tokamaks.

# *U.S. toroidal research coordination internationally, impact on ITER design*

The U.S. was a founding member (with Europe, Japan, and Russia) of the
International Tokamak Physics Activity (ITPA), established in 2000 to
coordinate world activity in tokamak research aimed primarily at burning
plasma physics. The U.S. is active in this important coordinating body,
which is currently headed by a U.S. researcher. Historically, the U.S.
facilities DIII-D and C-Mod have been important contributors to and
leaders of the international databases on tokamak confinement, plasma
disruptions, and divertor physics. In the past few years, coordinated
joint experiments on dimensional and dimensionless scaling across
devices of different size, magnetic field, and aspect ratio (under the
auspices of the IEA agreements) have engaged substantial experimental
run time on all three U.S. facilities, and plans have already been made
for significantly more of these coordinated international experiments
involving the exchange of both data and personnel. These joint
experiments are an important precursor to establishing the coordinated
experimental program necessary to operate ITER, as well as to guide a
number of ITER systems design issues not yet resolved.

The U.S. facilities DIII-D and C-Mod had a substantial impact on the
ITER design, which was finalized between 1998 and 2001. Most notable of
these impacts are the inclusion of greater shaping capability in the
plasma cross-section for advanced tokamak performance (pioneered by
research on DIII-D), the reduction of the divertor heat flux to
tolerable levels using the concept of a 'detached' plasma (developed and
diagnosed by C-Mod and DIII-D), the 'vertical target' design for the
ITER divertor in order to facilitate detached plasma operation
(initiated by C-Mod), the ITER design criteria for plasma disruption
forces (led by C-Mod measurements of current asymmetries during
disruptions), and plans for possible active control of MHD instabilities
on ITER (based on the discovery in DIII-D of the resistive wall mode as
a pressure-limiting mode and its world leadership in demonstrating
control of these modes). NSTX was under construction at the time of the
ITER design completion; it is now contributing to design improvements
and operational implementations for ITER through the ITPA.

# *Contributions to fusion science and vitality of U.S. fusion program*

The recently published FESAC Program Priorities Report describes
ten-year plans for fusion science research that show the three U.S.
major facilities as key contributors or leaders for most of the critical
issues in the Scientific Campaign areas of macroscopic plasma physics,
multi-scale transport, plasma-boundary interfaces, waves and energetic
particles, and fusion engineering science. What may not be so clear from
that report is that the three U.S. facilities are also national centers
for education and integration of national competency across the broad
range of fusion science. All are operated as "national research centers"
with substantial numbers of university collaborators. In the case of
DIII-D, more than half of the scientific staff members are
collaborators, with participants from ten major universities. NSTX has
about a half of its scientific program carried out by collaborators,
worldwide, and about one third from other U.S. institutions. C-Mod is
located at a high-profile U.S. research university, where it is the
centerpiece of fusion research. All three facilities are heavily
involved in supporting doctoral and post-doctoral training. For many
scientists at universities, participation in these three world-class
facilities provides expanded impact and drive and direction for
on-campus activity in both experiment and theory, while at the same time
university expertise and insights provide expanded capabilities, depth
of expertise, and contact with broader scientific community to the three
facilities. Another important aspect of fusion research is that there is
essentially no separation between "facility" and "user", as can be the
case for other disciplines (e.g. beam lines at a synchrotron light
facility). Fusion research requires the simultaneous contributions of
facility hardware, operation know-how and diagnostic capabilities, all
focused on the common goal of producing magnetic confinement fusion. In
other words, the evolving facility **is** part of the research. The U.S.
national fusion program is stronger overall for having these three
national research facilities, with improved capability to innovate and
advance the understanding of frontier fusion science.

# *What research opportunities would be lost*

The three major U.S. facilities---C-Mod, DIII-D, and NSTX---represent a
massive investment of talent, intellect, and finances in tackling the
key issues of toroidal confinement. Each of these facilities has made
seminal contributions to the development of toroidal plasma confinement
and the fundamental science and technology that undergird it. Their
*combined* resources allow the U.S. to continue with a major role in
answering critical scientific questions in magnetic fusion research. The
U.S. is well positioned by having these three facilities, whose
complementary capabilities enable unique research contributions. The
premature loss of any of the three facilities would significantly
undermine this U.S. strength.

Some specific examples of the losses that would be suffered by the U.S.
fusion program with premature closure for each facility, respectively,
are:

C-Mod: This is the only U.S. facility capable of developing lower hybrid
current drive for control of the plasma current near the plasma edge.
The U.S. would lose a key opportunity to study this important system now
being considered for steady-state advanced tokamak operation in ITER,
and it is implementing a major effort to do so. The leadership role of
the U.S. in the divertor and plasma-boundary area would be seriously
harmed. C-Mod features the combination of an all-(heavy)metal wall and
reactor-relevant power densities, thus providing critically needed
integrated data on material choices for ITER. As the largest fusion
experiment at a major U.S. research university, there would also be
significant damage to student recruitment and training as well as the
loss of a major point of contact for fusion science with the larger
scientific community.

DIII-D: The U.S. would lose its centerpiece capability in plasma
transport, control of MHD instabilities, divertor physics, and
development of the scientific basis for advanced tokamak operation on
ITER. In addition, as our primary national facility for maintaining and
developing core competencies for the study of burning plasmas, closure
of DIII-D would significantly damage the ability of the U.S. to
participate effectively in the ITER scientific program.

NSTX: As the newest U.S. facility, aiming at concept innovation and
anchoring our leadership in the world spherical torus program, the loss
of NSTX would seriously undermine the U.S. program in concept innovation
and eliminate a promising candidate for a Component Test Facility,
identified as an important element in U.S. Department of Energy fusion
energy development long range plan. Activity in this area would then be
turned over to Europe through MAST, although this device lacks
significant capabilities that NSTX has. Of equal importance, the loss of
the parametric breadth provided by NSTX would weaken many coordinated
studies in MHD stability, transport, divertor physics, and energetic
particles, as well as significantly narrow the scientific scope of
fusion research in the U.S.

*Recommendation*

In a February 2005 report titled *'The Knowledge Economy: is the United
States losing its competitive edge?',* the Task Force on the Future of
American Innovation "developed a set of benchmarks to assess the
international standing of the U.S. in science and technology. These
benchmarks in education, the science and engineering workforce,
scientific knowledge, innovation, investment and high-tech economic
output reveal troubling trends across the research and development
spectrum. The U.S. still leads the world in research and discovery, but
our advantage is rapidly eroding, and our global competitors may soon
overtake us."

In the particular area that is the subject of this report, that of the
scientifically interesting and broadly important topic of magnetic
fusion energy research, the Facility Panel finds that the U.S. at
present holds a position of international strength and leadership in
many areas. The next major magnetic fusion research facility will be the
offshore burning plasma experiment, ITER: the recent decision on its
site has fostered growing interest by the international community.  The
loss of any of the three of the major U.S. toroidal magnetic fusion
facilities would fundamentally jeopardize the ability of U.S.
researchers to perform relevant fusion research, and thus would
undermine the current U.S. position of international excellence.

**The Panel's recommendation is that the three major United States
toroidal magnetic fusion facilities continue operation to conduct
important unique and complementary research in support of fusion energy
sciences and ITER.**

Table of Contents

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+---+------------------------------------------------------------+----+
|   | **Characteristics and Contributions of the Three Major     |    |
|   | U.S. Toroidal Magnetic Fusion Facilities**                 |    |
+---+------------------------------------------------------------+----+
|   |                                                            | *p |
|   |                                                            | ag |
|   |                                                            | e* |
+---+------------------------------------------------------------+----+
|   | Facilities Panel Members                                   | 2  |
+---+------------------------------------------------------------+----+
|   | Executive Summary                                          | 3  |
+---+------------------------------------------------------------+----+
| 1 | Introduction and Context:                                  | 10 |
|   |                                                            |    |
|   | Toroidal Magnetic Fusion Research in the U.S.              |    |
+---+------------------------------------------------------------+----+
| 2 | Unique and Complementary Contributions to Scientific       | 16 |
|   | Campaigns                                                  |    |
+---+------------------------------------------------------------+----+
| 3 | Contributions to and Cooperation with the International    | 41 |
|   | Community                                                  |    |
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| 4 | What Would Be Lost, and Recommendation                     | 48 |
+---+------------------------------------------------------------+----+
| 5 | References                                                 | 53 |
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| A | Tables of Characteristics of Major Toroidal Magnetic       | 54 |
| 1 | Fusion Facilities                                          |    |
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| A | FESAC Facilities Charge                                    | 57 |
| 2 |                                                            |    |
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| A | The Three Major U.S. Toroidal Magnetic Fusion Facilities   | 58 |
| 3 | Joint Document                                             |    |
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| A | FESAC Facilities Panel June Meeting at the Naval Research  | 68 |
| 4 | Laboratory                                                 |    |
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| A | List of Terms and Acronyms                                 | 70 |
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|   | **Requested Information Provided by the Three Major U.S.   |    |
|   | Troidal Magnetic Fusion Facilities**                       |    |
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| 1 | Alcator C-Mod Input to the FESAC Facilities Panel          |    |
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| 2 | DIII-D National Fusion Program (submitted to the FESAC     |    |
|   | Facilities Panel)                                          |    |
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| 3 | NSTX Research and Its Role in Advancing Fusion Energy      |    |
|   | Science                                                    |    |
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|   |                                                            |    |
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1\. Introduction and Context:

######### Toroidal Magnetic Fusion Research in the U.S.

Over the past thirty-five years, toroidal magnetic fusion research in
the U.S. has developed from a collection of independent institutions and
research activities into a cohesive scientific research program that has
as its basis a complementary set of collaborative national experimental
research facilities. A consequence of this melding of intellectual vigor
and innovation from a spectrum of diverse organizations, many with
active international partners, is that fusion researchers in the U.S.
now agree that the 'knowledge base is now in hand to produce a burning
plasma -- a plasma whose high temperature is sustained predominantly by
energy from alpha particles produced by fusion reactions.'^1^ This
confidence results from a number of remarkable and important toroidal
confinement U.S. achievements in the last decade, including demonstrated
control of tokamak major disruptions, the enhancement of ion confinement
through sheared flows and transport barriers, the first-ever studies of
burning plasma behavior, and the sustained ability to operate high
performance plasmas in the MHD-stable advanced tokamak mode, with high
self-driven current, good particle and energy confinement, and a plasma
edge that enables particle and power handling. The now-sited
international experimental toroidal magnetic fusion reactor, ITER, which
is the highest-priority new facility of the U.S. Department of Energy
(DOE) Office of Science, will benefit significantly from these central
U.S. accomplishments. The vitality and vibrancy of the present U.S.
program leads to the expectation that the contributions to the world
fusion program from U.S. toroidal fusion energy sciences research will
continue to be seminal, substantive, and of high impact.

The U.S. magnetic fusion energy sciences research program evolved to its
present situation, with three major toroidal facilities, following
advice derived from a number of studies undertaken by various DOE
federal advisory committees (Magnetic Fusion Advisory Committee, Fusion
Energy Advisory Committee, Fusion Energy Sciences Advisory Committee),
panels of the National Academy of Sciences, national committees
(Toroidal Steering Committee, Ignition Physics Study Group, Transport
Task Force), and fusion community input (Coolfont Retreat, Snowmass
Summer Studies, Burning Plasma Physics Meetings). The three facilities
were deliberately planned for their respective capabilities, and they
plan their research programs jointly, e.g., through the Fusion
Facilities Coordinating Committee.

Each of the three major toroidal facilities in the current U.S.
portfolio developed from a particular thrust pioneered in the U.S.
fusion energy sciences program. In chronological order:

-   **DIII-D**: The concept that non-circularity might improve the
    performance of a toroidal plasma (tokamak) was developed in a number
    of countries, notably, Russia, France, and the U.S. In Europe, this
    thrust led to JET (U.K.) and later to TCV (Switzerland). In the
    U.S., non-circular plasmas have been developed mainly in the General
    Atomics program on D-II, D-III, and DIII-D. Following demonstration
    of the advantages of non-circularity, the ASDEX (Germany) and JT-60
    (Japan) facilities were upgraded to non-circular operation.

-   **Alcator C-Mod:** The use of high magnetic field to improve plasma
    performance and permit compact ignition was proposed at the
    Massachusetts Institute of Technology---taking advantage of their
    well-known Magnet Laboratory---and has been pursued through MIT's
    program on Alcator, Alcator-C, and Alcator C-Mod. This work has
    provided the underpinnings for all the proposed high-field burning
    plasma tokamaks (Ignitor, CIT, BPX, and FIRE).

-   **National Spherical Torus Experiment (NSTX):** The main advantage
    of a tokamak with very low aspect ratio is to provide access to
    operational regimes with extremely high relative plasma pressure
    (i.e., high β, where the dimensionless quantity β is the ratio of
    the plasma pressure to the magnetic field pressure). Successful,
    practical demonstration of this innovative confinement
    concept---also known as a spherical torus---began in the relatively
    small-size START device (U.K.) and has been followed up by the
    present MAST device (U.K.) and the present NSTX device, both of
    which are proof-of-principle class devices with plasma current on
    the order of 1 mega-ampere. The spherical torus configuration
    complements the other facilities by confirming important physics
    theories in higher aspect ratio tokamaks.

Machine and plasma parameters of these facilities are defined in
Appendix 1, Tables A and B respectively. In Appendix 1 Table C, machine
and plasma parameters of these U.S. major facilities are compared with
those of international facilities: ASDEX-U, FTU, JET, JT-60U, MAST, TCV,
and ITER. Two superconducting tokamaks, EAST and K-STAR, will start
operation soon. The tables succinctly illustrate that the U.S. toroidal
magnetic facilities collectively provide researchers with access to the
broad range of parameters necessary for unique and important,
forward-directed magnetic fusion research.

1.1. U.S. Position in Toroidal Magnetic Fusion Research

The three diverse and complementary U.S. major toroidal facilities,
derived from U.S. innovations, keep the nation at the forefront of
magnetic fusion energy sciences research worldwide. To cite a few unique
features of these three facilities (in alphabetical order):

-   **Alcator C-Mod (C-Mod),** with FY05 annual budget of \$22,038,000
    U.S., has the highest magnetic field and highest value of field
    strength to major radius (B/R) among the world's tokamaks.

-   **DIII-D,** with FY05 annual budget of \$55,751,000 U.S., has
    demonstrated significant flexibility for the creation of advanced
    operational scenarios.

-   **NSTX,** with FY05 annual budget of \$34,584,000 U.S.**,** is the
    most capable of the world's low-aspect-ratio tokamaks, with
    sufficient heating power and stabilization systems for accessing
    very high β and normalized β values.

Two major shared strengths of these facilities are exceptional
diagnostics and integration of experiments with theory and numerical
simulation. From the 1980's, the strongly coordinated community-wide
activity to understand anomalous plasma transport, under the auspices of
the U.S. Transport Task Force, has driven the development of new
diagnostics, benchmarking with theory and simulations, and targeted
experiments. The Office of Science major programmatic initiative,
Scientific Discovery through Advanced Computing (SciDAC), has more
recently enabled substantial development of U.S. capabilities for
numerical simulation of plasma behavior in specific targeted areas. This
latter effort has led to the enunciation of a new initiative, the Fusion
Simulation Project,^2^ whose goal is the fully integrated simulation of
a toroidally confined burning plasma and the predictive comparison of
simulation results to experiment by means of synthetic diagnostics. It
is expected that the Fusion Simulation Project will well utilize the
Office of Science's second-highest-priority new facility after ITER,
that of leadership-class computing.

On the international front, the three U.S. facilities are major
contributors. All three are participants in the International Tokamak
Physics Activity (ITPA), which has expanded the coordination of
international research. Results from the three U.S. facilities have
provided a significant portion of the world's database on fundamental
toroidal plasma physics. C-Mod and DIII-D have both had major impacts on
the design of the current version of ITER. (NSTX operation started in
1999, subsequent to the ITER redesign effort.) All three facilities are
involved in proposing upgrades to ITER operational scenarios. All three
also carry out joint experiments with international partners.

From the perspective of fundamental fusion science, the portfolio of
U.S. toroidal magnetic fusion facilities---with leverage from the world
program---continues to make major contributions to the science of
confined plasmas, providing detailed understanding of
magnetohydrodynamics (MHD), particle and energy transport, plasma-wall
interactions and divertors, heating and current drive, energetic
particle physics, and dynamical plasma control.

The three U.S. facilities are also paving the way for effective burning
plasma experiments on ITER. In the longer term they are expected to
provide the physics basis for a Component Test Facility (especially
NSTX), a Compact Ignition Experiment (especially C-Mod), and an improved
DEMO reactor (especially DIII-D) and ultimately a commercially viable
and environmentally benign Toroidal Fusion Power Plant (all three).

# 1.2. Importance of the Three U.S. Facilities 

The three major U.S. facilities---C-Mod, DIII-D, and NSTX---represent
massive national investments of talent, intellect, and financial
resources directed to key issues of toroidal plasma confinement. The
estimated replacement value of this portfolio is on the order of 1.2
billion dollars U.S. (see Appendix 1 Table B). Each of these facilities
has made seminal contributions to the development of toroidal plasma
confinement and the fundamental science and technology that underpin it;
as an integrated program, they provide an unparalleled test bed for
fusion energy sciences research.

One standard for judging progress in fusion research is the "triple
product" of plasma density, temperature, and confinement time. This may
be rephrased as the product of plasma β, magnetic field squared, and
confinement time. A high value of the triple product implies access to a
broad range of all relevant (dimensionless) plasma parameters and, thus
the ability to investigate fundamental plasma phenomena in regimes that
extend from laboratory fusion to wider reaches of astrophysical plasmas.
In its rephrased form, a high value of the triple product is achieved
within the U.S. toroidal portfolio. Specifically, Alcator C-Mod is the
highest field tokamak in the world; DIII-D pursues the direction of high
confinement through the development of advanced tokamak operational
regimes with long confinement times; and NSTX pursues the direction of
ultra-high β, resulting from its extreme toroidicity (i.e., low aspect
ratio). Importantly, while these three facilities are distinctly
different in some regards, they have a valuable degree of commonality
that permits them to compare plasma behavior under similar
dimensionless---but different dimensional---parameters (e.g., magnetic
field strengths from 0.3 Tesla up to 8 Tesla, all at high performance),
comparison which enables the development of fundamental predictive
understanding.

The *combined* resources of the present three major U.S facilities
provide an outstanding capability for answering critical scientific
questions in magnetic fusion research. The U.S. is well positioned by
having these three facilities, whose complementary capabilities enable
unique research contributions. All three of the U.S. facilities are
operationally very productive, with research results that have high
impact worldwide. This is evidenced by the large numbers of refereed
publications produced by the three facilities (more than one thousand
combined in the years following 1999), and by the impressive citation
counts for a number of these published journal articles. If only
articles that have appeared in print after 1998 and that have citation
counts in excess of 30 are considered, the highest citation counts are:
for C-Mod, one publications with 70 citations, one publication with more
than 50 citations, and two additional with more than 30 citations; for
DIII-D, two publications with more than 50 citations, five additional
with more than 40 citations, and two further with more than 30
citations; and, for NSTX, two publications with more than 40 citations.
(Note that ohmic operations commenced on NSTX in CY2000.)

The loss of any of the three facilities would fundamentally jeopardize
the ability of U.S. researchers to perform broadly relevant fusion
research, and thus would undermine the current U.S. position of strength
and leadership in this scientifically interesting and important
international research area, for which the next major facility will be
the offshore international experiment, ITER. To reach this conclusion,
the Panel considered the contributions^3^ from the major international
toroidal facilities---JET, JT-60U, ASDEX-U, TCV, TEXTOR, Tore Supra, and
MAST. From our understanding of the proposed international
activities,^4^ the Panel judges that the three U.S. facilities will be
making important and unique contributions to the world's magnetic fusion
program over the next five years and, furthermore, will jointly carry
out important confirmatory experiments. The U.S. will also continue to
collaborate with major international facilities through joint
experiments.

The U.S. facilities are also very important for training the next
generation of plasma physicists and fusion researchers. Their current
complements are:

-   C-Mod: 30 graduate students;

-   DIII-D: 16 graduate students;

-   NSTX: 16 graduate students.

This vital training activity is currently limited by funding constraints
and insufficient machine operation time. Moreover, these three
facilities and their host laboratories have each implemented significant
elementary and high school outreach programs.

Extensive information about prior accomplishments and expected future
contributions of the U.S. fusion energy sciences program may be found in
the FESAC Program Priorities Report.^1^

The charge to the Facilities Panel is reproduced in Appendix 2. The
Panel's responses to this charge are contained in the present document
(Volume 1). A companion document (Volume 2) contains information
submitted by each of the three major U.S. facilities.

1.3. Panel Activities Timeline and Report

The FESAC Facilities Panel (membership provided on page 2 of this
report) is comprised of ten researchers representing expertise in all
aspects of fusion energy sciences. Table 1.3-1 provides a work timeline
of the three months of Panel activity.

**Table 1.3-1. Facilities Panel 2005 Work Timeline.**

+-----------------------------------------------------------------------+
| April 5: Charge Letter is issued to FESAC                             |
|                                                                       |
| April 19: Panel membership is determined                              |
|                                                                       |
| April 28: Panel requests information from the three facilities        |
|                                                                       |
| May 28: Panel receives requested input from the three facilities      |
|                                                                       |
| June 2: Panel receives unsolicited joint facilities document          |
|                                                                       |
| June 13-17: Panel holds review meeting \[see Appendix 4 for           |
| schedule\]                                                            |
|                                                                       |
| July 11: Panel provides final report to the FESAC                     |
|                                                                       |
| July 19: Panel briefs FESAC per report contents                       |
+-----------------------------------------------------------------------+

The Panel's April 28 request for input from each of the three major
facilities included an invitation to each facility program director to
provide a document (30 pages maximum length) that addressed the panel
charge in detail. Each of the three '30-page\' documents that were
received from the facilities in response to this request is reproduced
in Volume 2 of this report, as background to the panel findings and
responses. On June 2, the Panel further received an unsolicited joint
document from the program directors of the three facilities the purpose
of which was to describe for Panel consideration the unique research
roles for the U.S. in the world program that are enabled by the combined
contributions of the three facilities. This thoughtful and detailed
document is reproduced with its cover letter in Appendix 3.

This report addresses the major toroidal magnetic facilities charge,
with focus on the four specific charge questions, reproduced in Table
1.3-2.

**Table 1.3-2. Facilities Panel Charge Questions (April 5, 2005).**

+-----------------------------------------------------------------------+
| 1.  What are the unique and complementary characteristics of each of  |
|     the major U.S. fusion facilities?                                 |
|                                                                       |
| 2.  How do the characteristics of each of the three U.S. fusion       |
|     facilities make the U.S. toroidal research program unique as a    |
|     whole in the international program?                               |
|                                                                       |
| 3.  How well do we cooperate with the international community in      |
|     coordinating research on our major facilities and how have we     |
|     exploited the special features of U.S. facilities in contributing |
|     to international fusion research, in general, and to the ITER     |
|     design specifically?                                              |
|                                                                       |
| 4.  How do these three facilities contribute to fusion science and    |
|     the vitality of the U.S. Fusion program? What research            |
|     opportunities would be lost by shutting down one of the major     |
|     facilities?                                                       |
+-----------------------------------------------------------------------+

Section 2, immediately following, provides the Panel's response to
charge Questions #1, #2, and the first part of Question #4. Charge
Question #3 is addressed in Sec. 3. In Sec. 4, the second part of
Question #4 is discussed primarily from the perspective of the Program
Priorities Report fusion energy sciences research objectives.

2\. Unique and Complementary Contributions to Scientific

Campaigns

Scientific understanding of magnetic fusion plasmas comes from a
combination of experimental and theoretical investigation. Crucial
experimental verification of theory requires facilities that can vary
plasma conditions over an extended range. No single facility can
adequately span all desired variations. The U.S. fusion program has a
complementary set of three major experimental facilities that are able
to access a very wide range of variations. The high magnetic field
approach of Alcator C-Mod, the advanced tokamak studies at intermediate
field on DIII-D, and the low-field, high-β capability in NSTX complement
each other technically, and also for collaborations. Furthermore, the
healthy synergy of similarity experiments conducted between two or more
facilities contributes to exploiting their full potential to test theory
and techniques. With their unique and complementary machine capabilities
and outstanding diagnostics, all three major U.S. facilities are
carrying out world leading magnetic fusion plasma research.

In this section, we address the charge questions about characteristics
and contributions (Questions #1, #2, and the first part of #4) in terms
of the five scientific campaign areas that describe the magnetic fusion
research program. These important campaign areas were identified in the
2005 FESAC Program Priorities Report as:

"Macroscopic plasma physics: Understand the role of magnetic structure
on plasma

confinement and the limits to plasma pressure in sustained magnetic
configurations;

> Multi-scale transport physics: Understand and control the physical
> processes that govern the confinement of heat, momentum, and particles
> in plasmas;
>
> Plasma boundary interfaces: Learn to control the interface between the
> 100-million-degree-C plasma and its room temperature surroundings;
>
> Waves and energetic particles: Learn to use waves and energetic
> particles to sustain and control high temperature plasmas;
>
> Fusion engineering science: Understand the fundamental properties of
> materials, and the engineering science of the harsh fusion
> environment."^1^

The unique and complementary characteristics and research contributions
of the three major U.S. toroidal facilities -- C-Mod, DIII-D and NSTX --
that are pertinent to these five major topical research categories are
overviewed in Secs. 2.1 -- 2.5 following. These subsections highlight
illustrative examples of ongoing research and capabilities of the three
experimental facilities and are not intended to be complete surveys.

# 2.1. Macroscopic Plasma Physics

Research in the area of macroscopic plasma physics seeks to determine
how to confine and sustain maximum plasma pressure efficiently in a
magnetic field configuration. This quest is extremely important, since
the production of fusion energy in a burning plasma facility (such as
ITER) increases with the square of the plasma pressure. The science is
complex, because plasma self-organizes into minimum energy states as it
nonlinearly interacts with the confining magnetic field, internal
electric fields, and plasma flows.

The three macroscopic plasma science topics of equilibrium, stability,
and external control are addressed individually by the first three
topical scientific questions, T1 -- T3, defined in the FESAC Program
Priorities Report. The balance of this section shows strengths of the
major U.S. magnetic fusion facilities in these areas.

####### 2.1.1. Plasma Equilibrium and Magnetic Field Structure 

The plasma pressure and current profiles, rotation, and the confining
magnetic field shape can be altered to increase the confined plasma
pressure. An appropriate dimensionless quantity used for measurement is
called β, which is the ratio of the plasma pressure to the magnetic
field pressure.

All three major U.S. facilities are able to match the ITER boundary
cross-sectional shape. Alcator C-Mod investigates plasma behavior at
high magnetic field strength. It has the highest toroidal field (8. T)
and the highest plasma pressure of any tokamak in the world. C-Mod
plasmas have intrinsic plasma rotation from torque-free heating with
radio frequency ion cyclotron waves. DIII-D has unparalleled ability to
study the scientific advantages of plasma shaping. DIII-D can closely
match the shape of most tokamaks in the world (Fig. 2.1-1), allowing
most important, inter-device similarity experiments. Advanced tokamak
studies to optimize β and extend lifetime can be performed on DIII-D
through pressure, current, and rotation profile variation, with
world-class diagnostics leading, hopefully, to conclusive comparisons to
theory. Through boundary and profile shaping, DIII-D has achieved
volume-averaged toroidal β of 12.5%, a world record for tokamaks with
conventional aspect ratios. DIII-D can vary plasma rotation, through its
co- and counter neutral beam injection capability. The low aspect ratio
geometry (high plasma toroidicity) of NSTX allows study of the many
critical macroscopic phenomena that depend on aspect ratio. Low aspect
ratio enables very high shaping factor and elongation, enabling very
high β. NSTX is the world's only tokamak to study plasmas in the range
of zero to unity (local) β: volume-averaged toroidal β in NSTX has
reached 40%, and normalized β has attained a world-record value of 7.2
through boundary shaping and pressure profile broadening. NSTX can
produce rapidly rotating plasmas with co-directional neutral beams,
allowing research at the uniquely high Alfvén Mach number (exceeding
0.45) reached in the device. Rotational effects on equilibrium are
advantageous and particularly strong at low aspect ratio, causing the
contours of constant plasma pressure to no longer be coincident with the
flux contours of the magnetic field.

![](media/image1.pct){width="5.301388888888889in"
height="4.815277777777778in"}

*2.1.2. Pressure-limiting Instabilities*

When the β value of magnetically confined plasma exceeds an upper limit,
large-scale unstable motions of the plasma can develop, leading to loss
of confinement. β-limiting macroscopic plasma instabilities include the
kink/ ballooning mode, the resistive wall mode (RWM), and the
neoclassical tearing mode (NTM). The kink/ ballooning mode can manifest
itself as an internal mode, a global mode, or an edge localized mode
(ELM), depending on the area of the plasma that is affected.
Understanding the science of the stability limits set by these modes is
an essential goal of magnetic fusion research.

High β plasmas can be stabilized by sufficient rotation. DIII-D was the
first device to show that it is possible to sustain stable plasma
operation above the ideal MHD "no-wall" limit by rotating the plasma
inside the electrically conducting vacuum vessel. Both DIII-D and NSTX
use a combination of rotation and conducting walls to passively
stabilize resistive wall modes and external kink modes. NSTX is the only
spherical torus in the world with such capability. At low rotation
values where these modes can become
![](media/image2.pct){width="5.992361111111111in"
height="2.738888888888889in"}unstable, both facilities are instrumented
to study their properties (Fig. 2.1-2).

U.S. facilities have a leadership role in the world program in
investigating how amplitude-saturated plasma modes can create drag that
slows down plasma rotation, thereby decreasing its stabilizing effect.
DIII-D and NSTX are currently performing experiments to examine drag
induced by saturated neoclassical tearing modes and edge localized
modes, by stable resistive wave modes ("resonant field amplification"),
and by the more catastrophic unstable resistive wall mode. Since the
drag theoretically depends on the aspect ratio, the safety factor, and
the toroidal mode number of the instability, NSTX (with its low aspect
ratio, *A*, and high safety factor, *q*) complements experimental
studies on DIII-D to validate theory.

Another key research topic is to understand the physics mechanism that
sets the critical plasma rotation frequency below which the resistive
wall mode goes unstable. Joint experiments between DIII-D and NSTX,
abetted by theory and numerical simulations, have recently discovered
that the critical rotation frequency normalized to the Alfvén speed
scales as 1/(*Aq^2^*). Theoretical explanations for the stabilizing
mechanism yielding the critical rotation frequency depend either on the
plasma sound speed or the Alfvén speed. The separation of the two speeds
in NSTX permits their effects to be distinguished.

DIII-D and NSTX are also studying the stability threshold for
neoclassical tearing modes, which is an area with important
contributions from ASDEX-U as well. These studies take into account the
related behavior of large internal "sawtooth" modes with toroidal mode
number *n* = 1. C-Mod plans to study issues of so-called giant sawteeth
and sawtooth control with localized ion cyclotron heating; DIII-D is
undertaking similar studies, taking advantage of its ability to
manipulate the fast ion and current profiles. NSTX is examining the
combination of rotation and kinetic stabilization effects on these
modes, since these effects are amplified at high β and Alfvén Mach
number.

DIII-D experiments, guided by theory, have led the way to establish the
kink/ ballooning mode with intermediate toroidal mode numbers as the
leading candidate to explain edge localized modes. Its comprehensive
diagnostics for the edge pedestal region allow accurate stability
predictions, while flexibility in boundary shape and in edge
collisionality (varying over two orders of magnitude) permit the
analysis of a wide range of edge plasma conditions.

Static, relatively small "error" magnetic fields pose a significant
limitation on plasma confinement by aggravating macroscopic
instabilities and also by reducing rotation. These error fields can be
amplified through resonant response of the plasma. Non-rotating "locked"
modes can grow and terminate plasma confinement. C-Mod, DIII-D, and NSTX
are all equipped with non-axisymmetric coils for complementary studies
(with JET) of the plasma response to error fields. Interestingly, C-Mod
results predict a significantly more favorable (five times higher) error
field threshold for locked modes in ITER, consistent with results in
DIII-D and JET. NSTX contributes studies of the sensitivity of plasma
rotation damping and instability threshold to the magnitude of the error
field at low aspect ratio.

A high-pressure plasma is potentially susceptible to rapid loss of its
stored energy and current if macroscopic instabilities grow unchecked.
C-Mod and DIII-D have unique capability to measure toroidal asymmetries
of the halo currents that are induced during "disruption" events. Strong
asymmetries are critical because they increase design limits on forces
that a burning plasma system must withstand and on damage caused by
high-power losses onto the plasma-facing components. Another key part in
the complementary study of disruption dynamics is control and
mitigation. DIII-D pioneered the technique of high-pressure gas
injection for mitigation of disruption damage, which is now being
implemented jointly on C-Mod for experiments at ITER-like high energy
density and pressure values. This technique of massive gas puffs will be
implemented on ITER.

####### 2.1.3. External Control and Self-organization 

Plasmas that are above instability limits can often be stabilized by
external means, with a self-organizing plasma response. Burning plasmas
will add the additional influence of strong self-heating due to alpha
particles and will provide new opportunities to study highly coupled
plasma dynamics and optimized control. All three national facilities are
involved in addressing external plasma control and self-organization by
studying the proper balance between internal and external control. The
objective is to understand the fundamental science that will allow
sustained, optimized fusion power production.

The plasma current is most efficiently sustained by non-inductive
current that is self-generated by pressure gradients; this is known as
the bootstrap current. Large fractions of the total plasma current---up
to 85% in DIII-D, and 60% in NSTX (the highest among all spherical
tori)---have been created in this way. To supplement the bootstrap
current, the U.S. facilities employ non-inductive currents that are
externally driven (typically in the outer region of the plasma) by a mix
of neutral beams and radio frequency waves. (Current drive capabilities
of the three facilities are described in Section 2.4.) DIII-D, which has
achieved up to 100% non-inductive operation, along with ASDEX-U, JET,
and JT-60, is among the world leading programs in the use of
non-inductive current to develop quasi-stationary high- β discharges and
high-performance operating scenarios that can be used in present-day and
future tokamak fusion facilities, including ITER.

![](media/image3.pct){width="5.301388888888889in"
height="3.5395833333333333in"}Sustained, self-organized plasma current
and pressure profiles must yield a stable plasma equilibrium. With
insufficient plasma flow to stabilize instabilities (such as the
resistive wall mode), active control coils can be used for this purpose.
Calculations using the present ITER design show that ITER would rotate
at one-half of the required rotation rate for RWM stabilization.
Proposed improvements of the ITER design to include active control coils
in the midplane access ports (Fig. 2.1-3) depend on demonstration of the
effectiveness of such coils in the next few years. There are existing
external and internal coils on DIII-D. NSTX has recently added a set of
external coils for error field and RWM feedback control of strongly
rotating, high β plasmas. DIII-D and NSTX will operate plasmas above the
no-wall limit by means of wall stabilization with rotation. The RWM
feedback coils inside the DIII-D vacuum vessel will be used to study
fast feedback stabilization of the RWM in plasmas with controlled low
rotation, as is expected in ITER. The rotation in DIII-D will be
controlled by toroidally balanced neutral beams. Initial results of
feedback experiments have been positive. DIII-D and NSTX plan to carry
out complementary studies that will complete the validation of active
stabilization physics, for extrapolation to the burning plasma regime of
ITER.

DIII-D is a world leader in active mode control systems for both
resistive wall modes and neoclassical tearing modes. Its resistive wall
mode control system includes both internal and external active control
coils, a high-speed digital feedback control processor, and an audio
amplifier system for wide bandwidth and reduced system delay (latency).
NSTX complements this research through studies of aspect ratio
dependency and resistive wall modes with toroidal mode numbers greater
than unity.

For active stabilization of neoclassical tearing modes -- an area first
tried on ASDEX-U -- DIII-D uses localized current drive from externally
applied electron cyclotron waves. The DIII-D integrated control system
uses real-time measurements of the pitch angle of the magnetic field
line to track the position of the rational surface where the
neoclassical tearing mode resides. With this capability, DIII-D has
demonstrated pre-emptive stabilization of neoclassical tearing modes.

The same active coils used for resistive wall mode control in DIII-D and
NSTX can also be used for control and mitigation of edge localized
modes. Their suppression, critically important for ITER, has already
been demonstrated using these coils in DIII-D, for ITER-like edge
collisionality, with the application of resonant magnetic fields.
Initial results on NSTX are not inconsistent with results from DIII-D.

Due to its high magnetic field, C-Mod generally achieves high
performance plasmas while remaining below the macroscopic instability
thresholds described above. Within the next five years, the C-Mod
advanced tokamak program enabled by the implementation of lower hybrid
current drive (see Sec. 2.4) will address issues of significant
bootstrap current generation and beta limits in higher pressure plasmas.

2.2. Multi-scale Transport Physics

In plasmas that are relatively free of large-scale instabilities, the
transport of energy, momentum, and particles along and across the
confining magnetic field lines exhibits complex dynamics and rich
phenomenology. Hence a deep understanding of the turbulence processes
that govern plasma transport has been described as one of the "grand
challenges" of fusion science and, indeed, of physics in general.
Small-scale turbulence is driven by plasma free energy or rotation. It
is the organization of this small-scale turbulence into larger
structures that can cause substantial leakage of plasma out of the
"magnetic bottle" at anomalously high rates, faster than what would be
expected from simple Coulomb collisions in toroidal geometry
(neoclassical transport).

Development of a predictive model for turbulent transport in magnetized
plasmas is, therefore, a major goal of the international fusion program,
motivated by the need to increase the confidence in extrapolating
confinement predictions to larger, more powerful facilities. With
advances in fundamental turbulence models, the invention of new
turbulence diagnostics, and the economical availability of computing
power, development of a predictive transport model is within reach for
the first time. Core plasma transport is discussed here, and edge
transport is discussed in the boundary section.

![](media/image4.pct){width="5.0625in" height="2.6840277777777777in"}The
U.S. is the leader of the international research effort to determine the
underlying instabilities responsible for plasma transport. The U.S.
program focuses on measuring plasma instabilities over the wavelength
ranges predicted by theory to be responsible for the cross-field
transport, as well as identifying the relevant processes that can damp
the turbulence and reduce the transport. The wavelength ranges of the
underlying turbulence instabilities are shown in Figure 2.2-1. The
longer wavelength range (left) corresponds to instabilities driven by
the ion temperature gradient (ITG), whereas the shorter wavelength
ranges (center and right) correspond to instabilities driven by the
trapped electron population (TEM) and the electron temperature gradient
(ETG). Spanning between the ITG and TEM wavelength ranges are
micro-tearing instabilities (μtearing), related to plasma resistivity.

![](media/image5.pct){width="5.179861111111111in"
height="2.1798611111111112in"}Each of the three U.S. facilities
contributes important elements to a vibrant domestic program. The U.S.
transport program is made strong by DIII-D, which has an integrated
program with measurements over the important wavelength ranges, as well
as the most comprehensive gyrokinetic transport model (see Fig. 2.2-2).

![](media/image6.pct){width="5.125in"
height="3.2715277777777776in"}Although the goals of the NSTX transport
program are similar to those of DIII-D, NSTX plasmas offer a different
parameter range over which to investigate these instabilities---in
particular, a wide range of β (ratio of plasma pressure to magnetic
field pressure), as shown in Fig. 2.2-3. High β enables examination of
whether electromagnetic turbulence is important, as compared to the
baseline electrostatic turbulence predicted to drive transport in lower
β plasmas.

![](media/image7.pct){width="5.549305555555556in"
height="2.796527777777778in"}In C-Mod plasmas, the studies of thermal
transport can be extended to different dimensional parameter
regimes---e.g., power density (Fig. 2.2-4) and toroidal field
B~t~---while overlap with other tokamaks is maintained in dimensionless
parameter space, e.g. β~Ν~ and ρ~\*~. In addition, C-Mod extends the
transport studies to the reactor-relevant regime for uniformly
equilibrated temperatures of electrons and ions.

The combination of the three machines improves the ability to separate
out the controlling variables through cross-machine studies of energy
confinement; in this way, for example, the importance of aspect ratio in
confinement and transport can be studied. This combination of facilities
also makes it possible to perform identity experiments (in which all
dimensionless parameters are held fixed) and similarity experiments (in
which one dimensionless parameter is scanned over a wide range) on two
or more domestic (and international) machines.

*2.2.1. Thermal Transport (Ions and Electrons)*

DIII-D pioneered the research on the role of radial electric field shear
(i.e., E×B shear) in suppressing ion turbulence and thus reducing
cross-field *ion thermal transport* to neoclassical levels. The
transport rate is not reduced everywhere across the plasma, but rather
within a distinctive region bounded by a "transport barrier," where the
density and temperature sharply fall off. A plasma with a barrier near
its edge is termed an "H-mode" plasma, whereas a barrier located farther
into the core of the plasma is called an "internal transport barrier"
(ITB). The beam emission spectroscopy diagnostic on DIII-D enables local
comparisons of the E×B shearing rate and the turbulence de-correlation
rate, which provide additional corroboration for the need to have E×B
shear in the vicinity of internal and edge transport barriers. NSTX
offers a different view of the problem. Because of the high edge
magnetic shear in NSTX endemic to low-aspect-ratio geometry, theory
predicts that ion-scale instabilities should be less important than in
usual-aspect-ratio tokamaks, leading to larger spatial regions of ion
thermal transport at the neoclassical level in H-mode plasmas.
Investigation of the effect of the ion-channel instabilities in NSTX is
facilitated by its charge-exchange recombination diagnostic, whose high
resolution can measure the ion temperature profile and flow profiles
down to the ion gyro-radius scale length. NSTX also plans to install a
new microwave turbulence imaging diagnostic with a large 2-D image for
turbulence measurements. An important issue for ion thermal transport
that is being investigated in all three facilities is the existence of
zonal flows, which can be described as large-scale poloidal rotation
along magnetic flux surfaces in narrow radial bands (similar to
planetary-scale flows observed on the surface of Jupiter). Zonal flows
are theoretically predicted to grow out of turbulence and then to
regulate that turbulence and its concomitant transport. Here, C-Mod
complements the DIII-D and NSTX data by extending the collisionality
range for a better assessment of collisional zonal flow damping of
turbulence.

*Electron thermal transport* is now a frontier for transport studies, in
part because of new and improved diagnostics that enable comprehensive
measurements. It appears that electron transport exceeds neoclassical
values even with electron transport barriers. It is not known for
certain if electron transport is dominated by short wavelength
turbulence such as electron temperature gradient (ETG) modes, which are
on the scale of the electron gyroradius, or longer wavelength turbulence
such as trapped electron modes (TEM) and μtearing modes. Domestically,
this research direction has been elevated in importance through
inter-facility cooperation promoted by the U.S. Transport Task Force
and, more recently, by the designation of electron transport as an
important research thrust in the report of the FESAC Program Priorities
Panel (2005). New backscattering and tangential scattering diagnostics
(on DIII-D and NSTX, respectively), developed primarily through
university collaborations, are adding to the phase contrast imaging
diagnostic (on C-Mod and DIII-D) in assessing fluctuations at
intermediate and short wavelengths. The three facilities can access
complementary plasma regimes since ion transport can sometimes dominate
total heat loss in high-performance C-Mod and DIII-D plasmas, whereas
electron heat transport always dominates heat loss in the highest
performance NSTX plasmas.

The damping of both ion and electron turbulence can be strongly affected
by the ratio of ion temperature to electron temperature: a high ratio is
stabilizing for ion transport gradient modes, whereas a low ratio is
stabilizing for electron temperature gradient modes. The three
facilities offer a flexible range of conditions to quantify the
importance of this ratio. Transport studies on C-Mod are conducted in
plasmas of high density (afforded by its operation at high magnetic
field), which enforces ion and electron temperatures to be everywhere
equal across the radial profile, a situation likely to occur in fusion
reactors. This strongly coupled ion-electron regime in C-Mod is
complemented by the flexibility to vary the value of the temperature
ratio by changing the fraction of neutral beam heating electron heating
via fast wave in NSTX or the fraction of ion cyclotron and electron
cyclotron radio frequency heating on DIII-D.

#### 2.2.2. Momentum and particle transport

*Momentum transport* is another important research frontier. Momentum
transport research is needed to help bridge the gap between present-day
experiments, which tend to have relatively high rotation rates, and
fusion reactors, which are expected to rotate significantly more slowly.
Recent C-Mod results demonstrate a need for basic understanding: C-Mod
ohmic and radio frequency wave-heated plasmas rotate rapidly, which is
surprising because radio frequency wave heating (in contrast to neutral
beam heating) does not impose external torque. These findings have
stimulated theoretical speculation and new experiments that suggest that
the plasma edge region plays a crucial role in core plasma flow. A new
rotation diagnostic (active charge exchange) will add to the existing
core rotation measurements by determining the importance of the edge
plasma in setting the boundary conditions for rotation of the core
plasma, and also in affecting the power level required to force the
formation of an edge H-mode transport barrier. For such studies, DIII-D
will soon have important flexibility enabled by the rotation of one
neutral beam line. The reconfigured beam line will allow up to 10 MW of
heating, while imparting zero net momentum to the plasma. An additional
10 MW of unidirectional neutral beam heating and the ability to reverse
the direction of the plasma current will give DIII-D the ability to vary
the direction of the torque and the ratio of heating power to torque
magnitude. NSTX can complement these momentum transport studies by
comparing forward and reversed plasma current in order to investigate
the origin of momentum transport with rates approaching the neoclassical
level.

Finally, a main goal of *particle transport* research is to predict the
residence time of fuel and impurities, as well as the build-up of helium
ash in a burning plasma. Historically, particle transport studies in
tokamaks have been conducted with the use of spectroscopy to track
helium or other injected impurities. Those studies yielded effective
diffusion coefficients and inward convective velocities, for comparison
with neoclassical predictions and scaling laws. Substantial progress was
made in the 1990's on important issues such as transport and exhaust of
helium and impurities, impurity screening, and density limits.
Consequently, this area of research presently receives less emphasis
than the other transport topics. New techniques to infer the turbulent
radial particle flux directly from beam emission spectroscopy data are
being pursued in DIII-D, while NSTX is focusing on impurity transport
studies, which have indicated diffusion rates comparable to neoclassical
values.

2.3. Plasma-Boundary Interfaces

![](media/image8.pct){width="5.299305555555556in"
height="2.8208333333333333in"}The science of the transitional plasma
that connects the hot (100,000,000 degrees C) thermonuclear "core" where
fusion occurs with the "terrestrial" material walls (20-1000 degrees C)
is the subject of plasma-boundary research. Confined and stable core
plasmas must still leak particles and power through this boundary layer.
This layer is only a few centimeters wide, and yet contains the largest
gradients in plasma pressure and magnetic field topology, making it a
critical region for stability and confinement. The prevailing magnetic
topology, and the design chosen for ITER, involves a "divertor" that
intentionally diverts boundary field lines to a dedicated target surface
in order to provide heat and particle exhaust. These field lines are
termed "open" because they strike material surfaces before having the
chance to close upon themselves. The dividing surface between the open
field lines and "closed" field lines in the core is called the magnetic
separatrix and the region outside of the separatrix is called the
scrape-off layer (Fig.2.3-1). The characteristics of the scrape-off
layer plasma are determined by the balance between transport along the
open field lines to the divertor target and transport across the open
field lines to the wall.

Plasmas with edge transport barriers, i.e., H-mode plasmas, exhibit
steep density and temperature radial gradients within centimeters of the
separatrix. The end of the steep gradients as one moves further into the
core plasma is called the H-mode pedestal. In this plasma-boundary
interface section, we discuss plasma transport from the pedestal into
the scrape-off layer, the divertor and the first wall, including the
topics of edge plasma optimization and wall material selection.

Plasma transport to wall surfaces, both along and across magnetic field
lines, represents a complex interaction between multiple states of
matter (gas, liquid, solid) and the highly turbulent plasma. Severe and
simultaneous demands are placed on the plasma-facing component (PFC)
materials for heat removal, plasma cleanliness, and compatibility with
fusion hydrogen fuels. Small changes in the highly non-linear
equilibrium of the boundary plasma can dramatically affect the overall
performance of a burning plasma. Hence, the study of plasma-boundary
interactions aims to answer the following scientific question: *How can
a 100-million-degree burning plasma be interfaced to its room
temperature surroundings?*

Understanding the physics of the boundary region with sufficient clarity
to reliably predict the behavior of a burning plasma experiment presents
one of the greatest challenges to the fusion community.

## 

## *2.3.1. Boundary Characteristics of U.S. Devices*

Prediction of boundary plasma behavior is more complex than for the core
plasma. The added complexities of the boundary plasma, such as atomic
physics and gas-plasma interaction, require extra dimensionless
parameters that cannot be matched simultaneously. Hence, there is a
strong necessity to have access to a wide range of boundary plasma
parameters in present fusion facilities for extrapolation to future
devices. Simultaneously, significant cross-device comparisons occur
between facilities due to both their common use of divertor topology (as
in ITER) and their strong links to a comprehensive suite of U.S. edge
models. The high-field C-Mod features local plasma parameters of power
and particle density nearly identical to those in the ITER divertor.
This allows for unique tests of radiation trapping and neutral/ gas
physics critical to ITER divertor operations. DIII-D features upper and
lower divertors that operate at lower density due to efficient in-vessel
pumping. This allows for unique control of plasma density in highly
shaped plasmas in an integrated manner with advanced tokamak research.
NSTX features the unique aspects of low-aspect-ratio edge plasmas
(including high expansion ratios of the poloidal flux to broaden the
heat footprint), short magnetic field connection lengths and high heat
flux to the divertor target. In combination the three U.S. facilities
have a strong focus on divertor and edge plasma diagnostics, which
provides a detailed set of measurements for model benchmarking. Taken
together, the U.S. facilities and computational modeling effort thus
provide an effective capability in edge and divertor science.

The U.S. facilities have made seminal contributions to the current ITER
boundary design. An operational regime known as detachment was
identified which greatly reduces divertor plasma temperature with the
goal of controlling the localized heat flux to the divertor targets.
ITER adopted the radiative divertor concept from C-Mod and DIII-D that
facilitates detachment. DIII-D's divertor Thomson scattering diagnostic
to measure the electron temperature verified the ultra-low temperature
plasma. C-Mod showed that the plasma recombines into neutral states
before it reaches the wall, in a sense demonstrating "detaching" the
plasma by suspending it off of the material surface. Divertor detachment
is the standard ITER operation scenario. A continuing unique U.S.
contribution in detachment physics is carried out on C-Mod on the effect
of line radiation trapping, which can make detachment more difficult to
be obtained.

*2.3.2. Plasma-facing Component Materials*

The design choice of the ITER wall material is strongly interlinked with
all the other boundary research topics, and remains an issue that
requires timely contributions in the next few years for ITER
construction. The existing ITER design uses a mixture of plasma-facing
component materials: beryllium metal main chamber walls, tungsten metal
near the divertor entrance, and carbon divertor plates (see Figure
2.3-1). These choices are based on the expected compatibility of each
material in different locations: beryllium to ensure only
low-atomic-number (low-Z) impurity in the core, high-Z tungsten for its
low sputtering yield, and carbon (which sublimates rather than melts)
for its resilience to transient heating. Present boundary plasma
research is identifying questions about the validity of these
trade-offs. In particular, carbon is favored for heating resilience, but
tritium retention and recovery from carbon could impact its operational
availability. Significant research is required in the next five years to
ensure an optimized design of the complex and expensive water-cooled
ITER wall components.

Due to their present diversity of wall materials, the U.S. fusion
facilities are well positioned to make key contributions to ITER. C-Mod
is the only divertor tokamak in the world with an all-metal, high-Z
plasma-facing component wall. With significant operational experience in
this area, C-Mod is playing a central role internationally in assessing
the effects of its molybdenum/ tungsten walls on high-Z impurity
radiation in the core, on surface melting, and on mitigation of tritium
retention without carbon. DIII-D and NSTX presently have all-carbon
plasma-facing component walls, allowing them to study carbon tritium
retention and recovery issues without the complicating aspects of
metals. This portfolio is well integrated with other facilities: JT-60U
and JET have carbon walls, with JET installing beryllium/ tungsten walls
in four years, and ASDEX-U is gradually switching from carbon to
tungsten-coated graphite.

## 

## 

## *2.3.3. Pedestal Physics* 

A high level of energy confinement is required for ITER to reach its
goal of fusion gain Q=10. A U.S. physics insight had identified the
controlling role of the plasma parameters at the top of the H-mode
pedestal. The pedestal temperature sets the global energy confinement
due to critical temperature-gradient transport in the core; however
there presently exists no first-principles model to predict the pedestal
parameters for ITER. The U.S. is leading an effort to understand
pedestal physics with comprehensive and complementary diagnostics and
theoretical models applied across the three facilities. C-Mod, DIII-D,
and NSTX all have spatially detailed measurements of the electron
density and temperature and ionization in the pedestal, along with
measurements of fluctuations that can drive transport. DIII-D has a
unique diagnostic for the edge plasma current. Measurements from all
facilities are currently being compared with common stability models
(e.g., ELITE) and edge turbulence models (e.g., BOUT). Importantly, the
three facilities span a large range of edge parameters: C-Mod operates
with high density and neutral opacity and is adding cryopumping for edge
density control for the FY06 campaign; DIII-D can lower the edge density
with cryopumps; and NSTX has very large magnetic shear. The U.S. goal in
the next five years is to provide the first predictive model of the edge
pedestal in a tokamak. These activities are focused and coordinated
through the Transport Task Force in the U.S. and the International
Tokamak Physics Activity (ITPA) internationally. Pedestal physics is a
high-leverage burning plasma science issue, and the U.S. is well
positioned to be a leader in this area during the next five years.

## *2.3.4. Edge Localized Modes*

As a consequence of the need for steep pressure gradients at the
pedestal, most high confinement regimes are characterized by edge
localized modes, which are MHD instabilities that cause repetitive,
sudden pulses of heat and particles through the edge plasma to material
surfaces. Baseline ITER operation is an H-mode plasma with edge
localized modes, the size of which are presently predicted to
transiently heat divertor materials past their melting and ablation
limits, risking unacceptable lifetime for the material targets. The U.S.
has a concerted cross-facility U.S. effort to: (1) identify the
modification to the edge stability that maintains good confinement and
high edge gradients without large edge localized modes; and, (2)
identify the processes responsible for residual steady-state particle
and heat transport. DIII-D and NSTX have high-resolution measurements of
edge localized modes, which are being compared in detail with results
from stability codes. C-Mod routinely accesses a high-collisionality
small ELM regime. DIII-D has identified a low-collisionality ELM-free
regime and uses an internal coil set for suppression of edge localized
modes. NSTX has identified a different small edge localized mode regime
that is compatible with very high β. Advances in understanding and
controlling edge localized modes is a high-leverage issue for the U.S.
relevant to the success of ITER.

## *2.3.5. Tritium Retention and Plasma-facing Components* 

Tritium retention is a critical issue for ITER construction, nuclear
licensing and operations. The radioactive tritium fuel used in ITER (or
any burning plasma) can be trapped in films caused by erosion of and
deposition on the walls by the boundary plasma. Extrapolation and
modeling suggest the ITER tritium in-vessel limit, mainly in carbon film
deposits, may be reached in only one to five days of operation,
necessitating tritium removal intervention that may compromise the
scientific productivity of ITER. However this prediction is uncertain
since it is based on data from all-carbon tokamaks, and modeling is
hampered by a lack of fundamental understanding for plasma particle
transport. To address tritium retention the U.S. facilities are
examining different approaches that are complementary to each other, and
to efforts in other tokamaks in the world. C-Mod has all-metal walls
(molybdenum/ tungsten) without any carbon surfaces. C-Mod has shown that
molybdenum walls have significantly less tritium retention (by factors
of ten) than tokamaks with carbon plasma-facing components. Largely
based on these C-Mod results, other facilities (ASDEX-U and JET) are
investigating whether to switch from carbon as a plasma-facing component
in order to solve the tritium retention problem. However molybdenum and
tungsten have other issues, namely melting during edge localized modes
and disruptions, and strong radiation in the core plasma, which affects
core plasma performance. C-Mod has also observed that frequent
application of low-Z films on top of the high-Z metal may be necessary
to achieve good plasma performance. C-Mod, with its ITER-relevant edge
power density, should continue to make vital contributions on these
topics in the next five years. DIII-D has all-carbon walls and is
therefore an ideal facility to pursue a strategy of understanding
tritium retention in order to maximize recovery of tritium from carbon
films. DIII-D is exploiting the unusually high toroidal symmetry of its
wall surface to carefully diagnose and account for carbon-13 isotope
tracers injected into the plasma, with the goal of understanding the
underlying carbon sources and transport that sets tritium retention.
This is complementary to similar studies on JET and ASDEX-U. DIII-D will
further explore the efficiency of oxygen baking to recover tritium by
laboratory studies and possibly in-vessel baking. Since oxygen baking is
presently the only tritium recovery technique planned for ITER,
controlled tests of its efficiency are critically required in the next
five years. NSTX has carbon walls and has presently installed a unique
in-situ diagnostic to measure the presence of dust containing carbon.
Tritium-laden dust is a safety concern in ITER. NSTX is also proposing
to test a liquid lithium divertor, which would provide both particle
pumping and heat flux control. Because it is liquid, lithium can form a
"renewable" divertor, and the tritium can be removed by lithium
recycling. However, liquid lithium may also experience MHD forces; NSTX
will test the stability of the liquid metal against such forces, an
activity that complements melt-layer studies of high-Z metals on C-Mod.

## 

## *2.3.6. Boundary Transport*

Reliable predictions of the boundary plasma and plasma-facing component
behavior require an understanding of the controlling mechanisms for the
complex turbulent transport of plasma to the surfaces. The U.S.
facilities have recently carried out several experimental and
theoretical research efforts on this subject. All three facilities have
documented the existence, first reported on C-Mod, of intermittent
plasma ejections by convective transport across magnetic fields. The
three facilities feature common diagnostic techniques, such as rapid
imaging and probes, across a large span of edge plasma parameters, thus
providing critical tests for the strong U.S. modeling effort on
turbulent transport. The presence of near-sonic plasma flow along field
lines is another topic of interest that is commonly studied across the
three facilities. This edge flow appears to be one of the explanations
for the prevalent retention of deposited films with tritium at the
inboard divertor, but it has eluded theoretical explanation to-date.
C-Mod research has recently suggested that the flow is caused by
ballooning transport at the outboard side of the plasma, and NSTX is
planning to image the 2-D flow patterns at different poloidal locations.
Toroidally symmetric injection of carbon-13 into DIII-D has confirmed
the link between the inboard-directed flow and carbon deposits at the
inner divertor. Taken together, the three U.S. facilities will play a
leading role in developing a better understanding of edge plasma
transport, which will lead to better predictability for ITER.

2.4. Waves and Energetic Particles

There is a large variety of electromagnetic waves that propagate in
magnetized plasmas, and their behavior is important for achieving
high-temperature plasmas. Waves, particularly in the radio frequency
range, that are applied to a plasma from the outside can accelerate
plasma ions and electrons through resonant interactions, thereby heating
the plasma and/ or driving non-inductive current. The applications of
such waves are becoming increasing important and precise tools for
stable high performance plasmas, and are essential for achieving
steady-state, advanced tokamak scenarios. Externally launched waves are
also useful for diagnosing plasma properties. Waves that arise
spontaneously inside plasmas can develop into eigenmodes that may cause
large-scale unstable disturbances in plasma motion. A "sea" of
turbulently fluctuating waves may enhance plasma transport, reducing
confinement. Thus, understanding and manipulating wave-particle
interactions are important thrusts in the U.S. and international fusion
research programs and will remain so during the next five years.

In addition to thermal ions and electrons, plasmas often contain a
special population of highly energetic (supra-thermal) particles, which
can strongly influence plasma behavior. The energetic particles may be
fast ions injected via neutral beams or generated via radio frequency
applied waves. Runaway electrons are another example. In burning plasmas
there arises an important class of energetic particles, viz., alpha
particles (helium-4 ions) born from thermonuclear reactions. These alpha
particles transfer their energy mostly to the thermal electrons, which
then through collisions heat the plasma ions that are the fusion "fuel."
Since this process represents the main heating mechanism in a
self-sustained fusion plasma, maintaining good confinement of the alpha
particles is vital. However, non-Maxwellian energetic ions and alpha
particles can excite unstable plasma waves that may degrade their
confinement; also, fast particles can affect the macro-stability of the
bulk plasma. Hence, understanding the behavior of fast particles will be
critical for successful operation of ITER and also is necessary for
fusion research today, particularly as studies of plasma confinement
push into new regimes in which the plasma pressure is high, the magnetic
field shear is reversed, and rotation is strong.

The three major U.S. facilities contribute significant and distinct
efforts toward advancing our knowledge about energetic particle behavior
and application of wave heating and current drive in fusion-grade
plasmas.

# *2.4.1. Wave Heating and Current Drive*

The specific techniques for radio frequency wave heating and
non-inductive current drive are different in each of the facilities, due
to the different properties of their plasmas and to the different tasks
for the waves (Table 2.4-1). Some of these techniques have been
investigated since the early years of controlled fusion research and are
now reliable and flexible, while others are relatively new. The wide
range of frequencies---from 30 MHz (characteristic of ion cyclotron
motion) up to 120 GHz (characteristic of electron cyclotron
motion)---means that quite different types of antennas must be used to
couple high-power radio waves into the plasma from its edge.

**Table 2.4-1.** Wave heating and current drive techniques on the three
facilities.

+-------+-------------+-----------+-------------+---------------------+
|       | **C-Mod**   | *         | **NSTX**    | **ITER**            |
|       |             | *DIII-D** |             |                     |
+-------+-------------+-----------+-------------+---------------------+
| ## *  | Ion         | Neutral   | Neutral     | Neutral beam        |
| *Heat | cyclotron   | beam      | beam        | injection (33 MW)   |
| ing** | resonant    | injection | injection   |                     |
|       | heating     |           |             | Ion cyclotron       |
|       | (minority   | Fast wave | Fast wave   | resonant heating    |
|       | ion & mode  | heating   |             | (20 MW)             |
|       | conversion) |           |             |                     |
|       |             | Electron  |             | Electron cyclotron  |
|       | Lower       | cyclotron |             | heating             |
|       | hybrid      | heating   |             |                     |
+-------+-------------+-----------+-------------+---------------------+
| **Cu  | Lower       | Neutral   | Neutral     | Electron cyclotron  |
| rrent | hybrid      | beam      | beam        | current drive (20   |
| Dr    |             | injection | injection   | MW)                 |
| ive** | Fast wave   |           |             |                     |
|       | current     | Electron  | Electron    | Lower hybrid        |
|       | drive       | cyclotron | Bernstein   | (option--20 MW)     |
|       |             | current   | waves       |                     |
|       | Mode        | drive     |             | Neutral beam        |
|       | conversion  |           | Coaxial     | injection (33 MW)   |
|       | current     | Fast wave | helicity    |                     |
|       | drive       | current   | injection   | Negative ion        |
|       |             | drive     |             | neutral beam        |
|       |             |           | Fast wave   | current drive       |
|       |             |           | current     |                     |
|       |             |           | drive       |                     |
+-------+-------------+-----------+-------------+---------------------+

DIII-D is the only U.S. facility to use electron cyclotron wave heating,
as a supplement to neutral beam injection, its primary heating method.
It also uses electron cyclotron waves to drive non-inductive current for
two purposes: (1) broadening the radial profile of the current, in order
to access high-gain, steady-state advanced tokamak operating modes; and
(2) suppressing resistive instabilities known as neoclassical tearing
modes at targeted radial locations. The theory of electron cyclotron
current drive has been extensively validated by experiments. These
instabilities are known to limit plasma β, and hence fusion performance,
if they grow to large amplitude, and may even cause major disruptions.
DIII-D is a leader in the worldwide effort to control neoclassical
tearing modes. Following initial international work, experiments on
DIII-D showed that electron cyclotron current drive can successfully
suppress neoclassical tearing modes. Over the next five years DIII-D
plans to double the power of its electron cyclotron waves, up to 6 MW,
making it the most powerful such system in the world. The efficiency of
neoclassical tearing mode suppression will be improved through power
modulation techniques and full integration of the electron cyclotron
current drive into the DIII-D advanced plasma control system; this will
permit the method to be validated for use on ITER, which will have 20 MW
of electron cyclotron current drive and heating. This is an area in
which ASDEX-U and TCV are also expected to make important contributions.

DIII-D also employs fast waves in the ion cyclotron frequency range to
drive non-inductive, steady-state current in the core of the plasma (in
contrast to electron cyclotron wave-driven current, which is mainly
off-axis). It is believed that extended-discharge scenarios will benefit
from control of the central current profile, and DIII-D will explore
this as an element of their integrated advanced tokamak research thrust.

Spherical tori, due to their innovative configuration, have limited
capability to produce purely inductively driven long-pulse discharges,
and therefore NSTX is exploring non-inductive current drive for current
sustainment and current profile control (the latter to achieve optimally
stable current and pressure profiles). However, the high density and
relatively low magnetic field in NSTX prevents electromagnetic waves at
the electron cyclotron frequency from penetrating the plasma, rendering
current drive by electron cyclotron waves unfeasible. Within the next
five years, NSTX will initiate current drive via electron Bernstein
waves: possibly the only method suitable for off-axis current driven in
over-dense plasmas. Electron Bernstein wave heating and current drive
were demonstrated on the W7-AS stellarator and the COMPASS-D tokamak.
Presently, NSTX is collaborating with the MAST facility (U.K.) to
optimize plans for its own studies. Because of their higher magnetic
fields, neither of the other two major U.S. facilities proposes---or
needs---to implement this scheme. Also, NSTX is unique in investigating
coaxial helicity injection, a non-inductive method (albeit not based on
radio frequency waves) to initiate plasma current in spherical tori.
Successful non-inductive localized current drive on NSTX will supplement
the bootstrap current to complete the current drive requirement for
steady-state advanced tokamak operation at high values for the plasma β.
Finally, like DIII-D, NSTX presently uses neutral beam heating and
current drive and is currently exploring the use of fast wave current
drive for central heating and non-inductive current generation in the
plasma core.

Alcator C-Mod is unique in that the entirety of its auxiliary heating
and current drive is obtained from the application of radio frequency
waves. Unlike DIII-D, NSTX, and major international facilities, C-Mod
does not use neutral beam heating. Instead, it exclusively employs ion
cyclotron resonant heating and has succeeded in making this method a
reliable heating tool, with a heating efficiency comparable to that of
neutral beams. This is a notable achievement that follows from decades
of multi-institutional work to understand the coupling, propagation, and
absorption of these waves and their antenna technology. Active research
on this subject during the next five years will continue to be a
priority in the U.S. program because ion cyclotron resonant heating is
more flexible than neutral beams in providing localized heating (and
current drive); this is particularly important for large fusion-grade
plasmas in which neutral beams cannot penetrate to the core. Moreover,
the ion cyclotron heating deposition profile is independent of density,
and can be tailored by the use of multiple sources with different
frequencies to make it extremely useful in confinement studies and
optimizing the generation of transport barriers. C-Mod will also use ion
cyclotron waves to explore novel methods of local current drive and flow
drive (i.e., momentum transfer to the plasma), involving the mode
conversion of the incident fast waves into ion Bernstein and ion
cyclotron branches. Since ion cyclotron resonant heating will be the
main radio frequency heating method on ITER, these studies on C-Mod are
crucial to U.S. participation in the development of key elements for
burning plasma experiments.

C-Mod has also recently begun non-inductive current drive experiments
with lower-hybrid waves that are intermediate in frequency between ion
cyclotron and electron cyclotron frequencies. Lower-hybrid waves have
the highest efficiency of all available (and proposed) methods of
wave-driven current drive, and they are particularly suited to produce
the current profiles needed for long-pulse, moderate β~N~ advanced
tokamak scenarios that need significant non-central current. On ITER,
lower hybrid is being held in reserve as the primary method for
auxiliary current drive in future advanced tokamak studies. Among the
three major U.S. facilities, lower hybrid current drive is practical
only for the high-density, moderate- β plasmas achievable in C-Mod. Its
3 MW lower-hybrid system will be key to this facility's contributions
over the next five years to the knowledge base for high-power-density
lower-hybrid current drive at ITER-relevant densities, made accessible
by the uniquely high magnetic field capability of C-Mod.

A particular strength of U.S. fusion research that compares favorably
with the international program is the capability to compare detailed
experimental measurements of wave propagation and absorption with
comprehensive full-wave modeling of radio frequency heating and current
drive scenarios. In C-Mod studies, excellent agreement of density
fluctuations associated with ion cyclotron heating, measured with the
novel phase contrast imaging diagnostic, is obtained with numerical
modeling from a state-of-the-art SciDAC code simulation of the launched
wave spectrum. Similarly advanced modeling and experimental validation
capability will available for the lower hybrid current drive studies on
C-Mod.

As a whole, the diversity of research on wave heating and current drive
being carried out on these three facilities solidifies U.S.
participation in the international burning plasma program and is also
essential for the development of innovative advanced tokamak confinement
scenarios, in which effort the U.S. fusion science program is a world
leader.

# *2.4.2. Energetic Particles*

![](media/image9.pct){width="5.25in" height="3.792361111111111in"}All
three major U.S. facilities pursue fast particle investigations with
different emphases, related to their characteristic plasma properties
and machine capabilities. To investigate the behavior of Alfvén-type
eigenmodes destabilized by energetic particles, a potential threat to
confinement in burning plasmas, the *fast ion velocity* should be
comparable to the plasma Alfvén velocity. NSTX, due to its low magnetic
field, can access values for the fast ion velocity that are several
times above Alfvénic, allowing excitation of high-frequency Alfvén
instabilities. The *fast ion β* (normalized pressure) measures the
strength of the excitation by fast particles; DIII-D, with its multiple
neutral beam ion sources, achieves high values for this parameter. The
*thermal plasma β* value is also relevant to fast particle modes, since
it skews the spectral gap where the modes occur and also affects their
damping rate. With NSTX (high β), DIII-D (moderate β), and C-Mod (low β
at high performance), a large variation in plasma β is available to the
U.S. program for such studies. As shown in Figure 2.4-1, the U.S.
facilities have fast ion to Alfvén velocity ratios (V~fast~/V~A~) and
fast ion to plasma β ratios (β~fast~/β~plasma~) relevant to the
projected operating space for ITER.

With regard to other key energetic particle parameters, the experimental
conditions in the three facilities are different from those expected in
ITER. A burning plasma like ITER will contain energetic alpha particles,
whose velocity distribution is isotropic, whereas fast ions created by
neutral beam injection (DIII-D and NSTX) or by ion cyclotron wave
heating (C-Mod) are anisotropic, with fast ion energy primarily in the
direction of the toroidal magnetic field for neutral beams or in the
cross-field direction for wave heating.

The *degree of anisotropy* can affect the strength of the Alfvén
eigenmode instability; fortunately, for equivalent energies, the
prediction for isotropic alpha particles in ITER will probably be more
optimistic than the result for either limit of anisotropy. Another key
parameter is the ratio of the *fast ion gyro-radius* to the plasma minor
radius (ρ∗~fast~ = ρ ~fast~ /*a*); this determines which are the most
unstable mode numbers. C-Mod performs studies at ρ∗~fast~ values that
are smaller than in NSTX and DIII-D and closer to the values expected in
ITER.

In other respects, the three U.S. facilities go beyond the standard ITER
operating space and explore features of energetic particle-excited
instabilities in broader regimes of operation. The regime of ultra-high
β is accessible in NSTX, as is also the issue of strong poloidal
coupling among Alfvén eigenmodes, thanks to NSTX's low aspect ratio
configuration. All three facilities can access the advanced tokamak
regime of flat and reversed magnetic field shear; in particular, they
plan to study how the known nonlinear properties of Alfvén cascade waves
can serve as an indirect diagnostic of the internal magnetic field.
Also, the stability of fast ion-excited instabilities can be examined in
regimes of rapid plasma rotation (NSTX and DIII-D) and low rotation
(C-Mod, DIII-D and NSTX), to see if fast ion losses are thereby
minimized.

A characteristic of all three U.S. facilities is that they have very
good diagnostics for energetic particle physics studies. C-Mod is unique
within the U.S. program for having active external antennae, by means of
which the damping of Alfvén eigenmodes can be measured in the absence of
fast ions; these studies are collaborative with JET, where this method
originated. C-Mod is using its new phase contrast imaging diagnostic
(which it has also installed on DIII-D) to observe the modes deep inside
the plasma. DIII-D is the only machine using beam emission spectroscopy
to see the internal modes. DIII-D is developing a new diagnostic, based
on D~α~ emission, to measure the distribution of fast ions. NSTX has an
especially good set of fast ion diagnostics, including a scanning
particle analyzer, lost ion probes, and plans a neutron collimator. All
three machines either now or will soon have the powerful motional Stark
effect diagnostic to measure the internal magnetic field profile, which
is sensitive to localized beam current drive and on which the Alfvén
eigenmodes have a sensitive dependence. The NSTX is unique in being able
to measure q(r,t) at B fields where V~beam~ \> V~Alfven~.

In the world arena, the three U.S. facilities are able to make
significant contributions to energetic physics research. Since the
shutdown of the U.S. tokamak TFTR, experimental leadership in this area
has belonged to the world's largest tokamaks, JET (in Europe) and JT-60U
(in Japan). JET produces fast ions with either neutral beams or ion
cyclotron waves; it has also carried out deuterium-tritium campaigns
with alpha particles (following the initial such investigations on
TFTR); it also injects helium ions to mimic alpha particles. JT-60U
likewise creates fast ions with both beams and waves; in particular, it
has implemented the technology of neutral beam injection for negatively
charged ions, with much increased beam ion energy. To diagnose the fast
ions, both JET and JT-60U use neutron measurements, and JET also has
two-dimensional gamma-ray tomography. Healthy collaborative links have
been established by the U.S. facilities with these energetic particle
efforts abroad, with U.S. scientists installing new diagnostics and
running experiments.

The three U.S. facilities provide key, distinct capabilities for the
international effort in this research area. Flexibility in manipulating
plasma shape and profiles is a major strength. DIII-D, for example, can
be configured as a quarter-size ITER. Shape flexibility permits
cross-machine similarity experiments; a recent example was the
confirmation on DIII-D and NSTX of the predicted scaling with fast ion
gyro-radius for toroidal Alfvén eigenmodes driven by beam ions (as
demonstrated on NSTX). Flexibility also extends to controlling the fast
particles. With its co- and counter-directional neutral beam injection
ability and also fast wave coupling to beam ions, DIII-D can change the
isotropy, energy, and pressure of the fast particle population, enabling
the facility to access a wide variety of energetic ion phenomena.
Another strength is that energetic particle experiments on the three
major facilities benefit from close interaction with theory and
numerical simulations---a traditional characteristic of the entire U.S.
fusion science program.

Since the behavior of burning plasmas will be dominated by a large
population of fusion-product energetic alpha particles, as well as beam-
and wave-generated fast ions, understanding their effects is vitally
important for the successful operation of ITER. The three U.S.
facilities are poised to make useful contributions to this knowledge
base.

2.5. Fusion Engineering Science

The FESAC Program Priorities Report^1^ states that the objective of
fusion engineering science is to "Understand the fundamental properties
of materials and the engineering science in the harsh fusion
environment." Much of the research activity in this area is outside the
scope of the experimental programs of the three major U.S. toroidal
facilities; hence this section will be brief. (A related discussion can
be found in Sec.3).

The aspects of fusion engineering science that are relevant to the three
facilities are primarily concerned with providing the knowledge base in
plasma technologies for heating, current drive, fueling, and
plasma-facing materials (particle and impurity control). In addition,
there is the goal of supporting the construction and safe operation of
ITER, including disruption mitigation. This research, overall, is highly
collaborative and productive; see Fig. 2.5-1 for examples of fusion
engineering science capabilities
![](media/image10.pct){width="5.68125in"
height="2.0944444444444446in"}developed by collaborating institutions.

The wide range of complementary wave heating schemes used in the three
U.S. facilities are important for both profile and instability control:
C-Mod uses ion cyclotron and lower hybrid waves; DIII-D used electron
cyclotron and ion cyclotron waves; and, NSTX uses electron Bernstein and
plans to use electron Bernstein waves. Figure 2.5-1(a) shows the
steerable electron cyclotron wave launcher built for DIII-D by PPPL.

NSTX is studying coaxial helicity injection in order to enable
solenoid-free startup, which is important for future embodiments of the
spherical torus concept.

The choice of plasma materials for ITER remains an issue. Presently ITER
has plans to use beryllium walls and carbon fiber composites and
tungsten brushes for the divertor. JET is undertaking tests of beryllium
and will also test tungsten. C-Mod has molybdenum walls and in-situ
boron coatings and is testing tungsten brushes; the tungsten divertor
wall element built for C-Mod by Sandia National Laboratories is shown in
Fig. 2.5-1(b). DIII-D uses all carbon and may test hydrogen recovery
with oxygen baking. NSTX uses carbon; it plans to test liquid lithium
for pumping and as a divertor target. NSTX is also using lithium pellets
and plans to use a lithium evaporator. Both C-Mod and NSTX have
ITER-level divertor power densities.

For density control in H-modes, DIII-D uses divertor cryopumping. C-Mod
plans to install a cryopump and the neutrals will be in a fluid regime,
as compared with the kinetic regime in DIII-D. C-Mod operates at
ITER-level power density and opacity in the scrape-off layer. A variety
of fueling schemes is in use, including pellet injection on all three
facilities . NSTX plans to install a compact torus injector.

The detection, avoidance, and mitigation of disruptions in modes of
operation important for ITER are being studied in C-Mod and DIII-D.

Significant contributions have been and will continue to be made by the
U.S. facilities in demonstrating integrated operational scenarios that
maximize performance -- β, transport, impurity minimization, steady
state -- while minimizing undesirable wall and divertor interactions.
The Multiple Input Multiple Output dynamic control system in DIII-D uses
inputs from profiles and instability mode behavior; this work is very
important for ITER.

The MDSplus data management system developed at C-Mod is now used at 30
sites internationally, as well as for ITER databases and for remote
access to JET. This system could be very useful for international
operation of the ITER facility.

The SciDAC-funded Fusion Collaboratory tools have been implemented on
all three facilities to enable remote participation and serve as a
prototype for U.S. participation in ITER operation.

There are areas in which the U.S. facilities will not make a
contribution. Key are:

-   Operation with burning deuterium-tritium plasmas. Following the
    shut-down of TFTR, JET remains the only experiment with this
    capability.

-   Operation with ultra-long to "steady state" plasmas. Tore Supra
    operates for 100's of seconds with a limiter plasma. The
    superconducting divertor tokamaks -- EAST and K-STAR -- will make
    important contributions after they start-up and have had time to
    optimize their complement of heating, current-drive, and diagnostic
    system.

3\. Contributions to and Cooperation with

> the International Community

This section presents an overview of the research contributions made by
the U.S. facilities to the international program in burning plasma
research and of their future plans with emphasis on the near term (i.e.,
ITER). The research on the U.S. facilities will be placed in the context
of the cooperative research effort carried out with international
facilities. This section, therefore, addresses Question #3 of the charge
to the Panel.

As background to this section, key plasma parameters and machine
parameters of C-Mod, DIII-D and NSTX are provided in Appendix 1 Tables
A1-1 and A1-2 respectively. Table A1-3 of Appendix 1 provides the same
key plasma and machine parameters for a subset of operating
international facilities (ASDEX-U, FTU, JET, JT-60U, MAST, TCV), and for
ITER.

The current framework for coordinating much of the international tokamak
research is the International Tokamak Physics Activity (ITPA), which
aims at cooperation in development of the physics basis for burning
tokamak plasmas. The ITPA includes researchers from several countries
(including the European Union, Russia, U.S., and Japan) and covers
several toroidal devices (viz., tokamak, stellarator, and spherical
torus). Over 50 U.S. scientists are members of the various ITPA working
groups, and the overall head of the ITPA is also a U.S. scientist.

Each year, the ITPA agrees on a list of experiments, analyses, and
modeling efforts that will be completed on U.S. and international
facilities in the following areas: confinement database and modeling;
transport; pedestal and edge physics; divertor and scrape-off layer; MHD
disruptions and control; steady-state operation; and diagnostics. It is
important to note that the U.S. research for ITPA comprises a strong
subset of the fundamental science and technology for a robust fusion
energy *sciences* program, independent of any particular confinement
approach.

The combined contributions of the three major U.S. facilities constitute
an important part of nearly all of the ITPA tasks. At a recent meeting,
the ITPA enumerated several advancements in the tokamak physics basis.
The U.S. facilities played important roles in each of these
advancements:

-   Achievement of good H-mode confinement at ITER plasma density;

-   Development of alternative H-mode regimes with ITER-required
    performance at lower plasma current and higher edge safety factor;

-   Demonstration of enhanced helium exhaust by elastic scattering;

-   Improvements in theory-based confinement validation and projection;

-   Confirming scaling-based predictions of fusion amplification factor;

-   Progress in elucidating the coupling between plasma edge conditions
    and core confinement;

-   Improved understanding of edge localized mode (ELM) physics,
    developments in ELM mitigation techniques, and alternative ELM
    regimes with high confinement;

-   Demonstration of active feedback control of neoclassical tearing
    modes by electron cyclotron wave current wave and resultant
    improvement in the sustainable β values;

-   Demonstration of increased β-limit by stabilization of resistive
    wall modes;

-   Demonstration of disruption mitigation by strong impurity puffing.

To draw from the FESAC Program Priorities Report, burning plasma
research encompasses the two themes of "create a star on earth" and
"develop the science and technology to realize fusion energy." It
involves the *integration* of the fundamental science and technology
issues that were discussed in Section 2. In the near term, the research
will have application to ITER, in terms of qualifying its existing
design and motivating possible improvements. Research for the longer
term will include a broad range of generic burning plasma issues, which
will help validate the physics of a demonstration power plant (DEMO).

The ITER schedule envisions first plasma operation ten years from the
start of construction. During those ten years, the U.S. will contribute
about ten percent of the funds for construction of ITER, building
several of its major systems. At the same time, coordinated research on
the U.S. facilities will prepare us for full and effective participation
in ITER operation and research by: (1) increasing confidence in the
current ITER design, e.g., the choice of wall material; (2) providing
information for design decisions not yet finalized, e.g., details of
heating systems; (3) suggesting possible improvements to the baseline
design, e.g. magnetic control coils for reducing field errors and
stabilizing MHD mode; (4) developing new measurement techniques,
diagnostics, and control systems controls, e.g., feedback sensing and
mitigation of plasma disruptions; and (5) enhancing theory and
integrated modeling, so that advanced simulation capability will be
available to design experiments on ITER. More general research for a
burning plasma experiment to provide scientific understanding and
concept improvements will also be pursued (e.g., advanced scenarios in
highly shaped plasmas, high toroidal magnetic fields, different aspect
ratios), even though not immediately applicable to the present ITER
design.

The strength of the three U.S. facilities *in combination* will be
demonstrated by means of a few illustrative examples of how they have
directly contributed to fusion science understanding, with impact on the
ITER design.

*3.1. Transport Basis for ITER Operating Scenarios*

The U.S. leads the world in transport research, including the invention
of new diagnostics to measure the underlying turbulence at the important
wavelength scales and the development of comprehensive first-principles
transport models. This work has focused around the U.S. Transport Task
Force since the 1980's and, more recently, around the ITPA. C-Mod
examines confinement at high plasma density, with equilibrated electron
and ion temperatures, and has novel diagnostics, including the phase
contrast imaging system. DIII-D has a comprehensive diagnostic set for
measuring fluctuations over a wide range of wavelengths, and has
developed the most comprehensive transport code in the world for
detailed comparisons of experiment with theory. NSTX examines how
transport scales with aspect ratio at high β, and is commissioning
fluctuation diagnostics that cover the relevant wavelength regimes.

The integrated theory and diagnostic programs of the U.S. have
identified important issues not evident from multi-machine scalings. In
the mid-1990's, core transport models such as IFS-PPPL and GLF23
demonstrated that ITER's performance projections are very sensitive to
the assumed electron temperature at the top of the H-mode pedestal,
prompting a world-wide emphasis on pedestal research. C-Mod (in June,
1997) and ASDEX-U (in July 1997) were the first devices to publish the
confirmation of a strong correlation between the pedestal electron
pressure and the confinement enhancement factor. Subsequent similarity
experiments among C-Mod , DIII-D and ASDEX-U and JET have sought to
isolate the dependence of the density pedestal height and width on the
neutral penetration depth, and a recent experiment among NSTX, MAST, and
DIII-D seeks to isolate the effect of aspect ratio on the pedestal
parameters.

Despite the progress in core and pedestal research, more work remains to
be done in transport science. For example, the physics of electron and
momentum transport is not well understood and represents a leading
frontier in transport research, with each machine making unique
contributions. Still, the U.S. transport program has improved---and will
continue to improve---the reliability of such transport predictions for
ITER.

The projections for ITER concerning its plasma performance and access to
H-mode confinement are based on an international database, for which
data from C-Mod, DIII-D, and NSTX are important for an accurate
determination of the relevant scaling variables. In combination with the
larger international machines such as ASDEX-U, JET and JT-60U, C-Mod and
DIII-D provide data crucial for an accurate size extrapolation to ITER.
In addition, NSTX data facilitates an evaluation of the effect of aspect
ratio differences among machines in scaling up to ITER.

*3.2. Wall Lifetime with Bursts of Heat and Particles*

The heat loads from transient phenomena such as edge localized modes
(ELMs) are predicted to cause ablation of plasma-facing components in
ITER. This has stimulated world-wide research on high-performance
regimes with small or no edge localized modes. The U.S. facilities have
(1) developed and benchmarked models of the edge localized mode, (2)
developed operational regimes with tolerable edge localized modes, and
(3) pioneered the use of edge resonant magnetic perturbations to avoid
edge localized modes altogether.

-   For (1), all three machines have world-class diagnostics for
    characterizing edge localized modes and measuring intermittent,
    cross-field transport.

-   In (2), DIII-D developed a high-performance, low-collisionality
    "quiescent H-mode" (QH), which has no edge localized modes but
    achieves an acceptable cross-field particle transport rate through
    edge MHD activity. The quiescent H-mode has been subsequently
    reproduced in several international machines, and its
    extrapolability to ITER is being assessed. The direction of neutral
    beam injection is important in this mode, and JT-60 has done
    balanced injection, i.e., with no torque input. One of the DIII-D
    beam lines is presently being reversed, in order to extend the
    JT-60U experiments. C-Mod discovered the "enhanced D-alpha H-mode"
    with no edge localized modes, which also relies on edge MHD activity
    for particle control. In addition, NSTX found a new small-ELM regime
    (Type V edge localized modes), which allows sufficient particle
    control with minimal thermal transients; the similarity between the
    NSTX regime and small-ELM regimes in C-Mod and MAST is being studied
    through the ITPA.

-   In (3), DIII-D pioneered the use of internal coils to apply an n=3
    magnetic perturbation, which suppresses edge localized modes but
    enhances cross-field diffusive particle transport. Similar
    experiments are being conducted in NSTX, with edge localized mode
    mitigation sometimes observed. The DIII-D success with resonant
    magnetic perturbations has led to the planned installation of edge
    localized mode control coils in JET. The U.S. is also developing the
    physics basis to propose this system for ITER. Just as international
    researchers will participate in experiments on DIII-D, U.S.
    researchers are involved in upcoming JET experiments. An
    international database is being developed, which could eventually be
    incorporated into the ITER design.

*3.3. Machine Damage due to Sudden Disruptions*

The plasma in ITER will contain a large amount of stored energy. If MHD
instabilities grow, the plasma pressure can -- either locally or
globally -- exceed stable operating limits of the confining magnetic
fields. The plasma can be quickly lost ("disruption"), and this can
result in large forces on the tokamak mechanical structure and also
large heat loads on the walls. The U.S. has been the leader in
experiments and modeling to detect an impending disruption and then
apply a large puff of gas to safely extinguish the disrupting plasma.
DIII-D did the first such mitigation experiments, along with detailed
measurements and modeling of the physical processes. This work has now
been expanded to C-Mod, which provides a critical test of the gas puff
technique due to its high absolute plasma pressure. The effort to
determine the scaling and distribution of disruption forces has been led
by C-Mod. The ITPA has a joint program of disruption experiments among
C-Mod, DIII-D, and JET; these machines span the parameter range from
high absolute plasma pressure in C-Mod to large plasma size in JET. NSTX
has contributed in a somewhat unexpected way: because spherical tori
exhibit moderate resilience against disruptions, understanding this
property may allow application in higher aspect ratio devices. Reducing
the severity of disruptions in ITER is important for allowing a safety
margin in the engineering design.

*3.4. Tritium Inventory and Choice of Wall Materials*

The present ITER design calls for beryllium main walls and a combination
of tungsten and carbon walls in the divertor region. Most experience on
international tokamaks has been with carbon surfaces at high heat and
particle fluxes. Carbon is fairly forgiving with respect to large
transient heat loads, such as disruptions, and it does not radiate
strongly if it migrates to the core plasma. However, carbon that is
eroded from the wall can co-deposit with the tritium fuel on the vessel
walls, resulting in an undesirably large inventory of trapped
radioactive tritium. The U.S. facilities, along with international
tokamaks, are attacking various parts of this problem with two different
and complementary approaches: namely, (1) to further characterize carbon
erosion and re-deposition, carbon transport, and surface film and dust
generation, as well as to study tritium removal techniques; and (2) to
examine other materials, such as molybdenum and tungsten, which do not
lead to tritium accumulation, but which can melt with disruptions and
can radiate strongly if they penetrate into the core plasma. DIII-D is
an all-carbon machine and is focusing on carbon transport experiments to
determine where the tritium would be co-deposited and on in-situ
techniques to remove the tritium. NSTX is mostly carbon at present;
since it plans to add lithium to the divertor region for an integrated
power and particle control solution, it will provide a good test of
mixed materials. C-Mod is an all-metal molybdenum machine. It is adding
tungsten in a unique "brush" structure, similar to the ITER design, to
address the melting issues. The U.S. machines can address a
comprehensive set of issues relevant to the selection of the wall
material for ITER. They also complement international facilities:
ASDEX-U (gradually converting to tungsten-coated carbon), JT-60U
(carbon), and JET (beryllium and carbon, with plans in the next few
years for beryllium walls with a tungsten divertor). In as much as the
ITER divertor is a modular "cassette," there is still time to influence
its design and choice of materials. It may also be the case that
different wall materials will be used during the startup phase (without
tritium) and the later operational phase of ITER.

*3.5. MHD Stability, Plasma Rotation, and Feedback Control with Magnetic
Coils*

The U.S. has been the leader in research on the characterization and
control of macroscopic MHD instabilities, particularly the resistive
wall mode and the neoclassical tearing mode. DIII-D and NSTX have
demonstrated that the resistive wall mode can be stabilized with
sufficient plasma rotation. NSTX revealed the safety factor dependence
of the rotation required for resistive wall mode stabilization, and
joint work between NSTX DIII-D, guided by theory, is revealing the
aspect ratio dependence. Some uncertainty remains concerning the level
of rotation for mode stabilization in ITER. Research to be conducted on
DIII-D and NSTX in the next five years should clarify the physics
mechanism for resistive wall mode stabilization, thus allowing more
accurate calculation of the critical rotation speed required for
stabilization. C-Mod was the first to discover plasma rotation with
radio frequency wave heating, even though it provides no external
momentum input; understanding this phenomenon could be a key element in
determining the level of rotation in ITER. DIII-D and NSTX will conduct
experiments in slowly rotating target plasmas with external control
coils in order to study the requirements for stabilizing the resistive
wall mode if ITER does not have sufficient rotation. DIII-D has
demonstrated feedback stabilization of the mode; the installation of a
counter-directed neutral beam will allow more variation in rotation and
further exploration of feedback stabilization scenarios.

Stabilization of neoclassical tearing modes by means of localized
current driven with radio frequency waves was pioneered by ASDEX-U and
DIII-D. DIII-D has recently demonstrated closed-loop feedback
stabilization of this mode, resulting in higher β plasmas; planned
modulation experiments will attempt to reduce the amount of radio
frequency wave power that would be required in ITER. Scaling studies
among DIII-D, JET, and ASDEX-U have been used to understand the
underlying physics mechanism and to extrapolate the results to ITER.
NSTX will use a different form of radio frequency current drive (based
on electron Bernstein waves) for neoclassical tearing mode
stabilization.

Departures from axisymmetry in the structure of the tokamak magnetic
field can destabilize non-rotating tearing modes ("locked modes") that
may significantly impact plasma operation. Locked modes induced by such
error fields can lead to disruptions, as well as degradation of
confinement. The impact of these modes on the operation of future
burning plasma experiments, such as ITER, has been a matter of concern;
key issues are the prediction of the error-field sensitivity (threshold
perturbation) in such devices and the requirements for corrective
measures. Joint experiments through the ITPA that make use of
non-axisymmetric coils to apply known field perturbations are being
undertaken among C-Mod, DIII-D, and JET in order to provide a basis for
extrapolative prediction (in field and size) of the symmetry
requirements for ITER. These experiments are conducted with matching
shape and normalized plasma parameters in the three devices, spanning a
range of size of nearly a factor of five. Preliminary results comparing
C-Mod and JET indicate that the required symmetry is on the order of one
part in 10^4^, which is within the capability of the planned ITER
correction coil system. Earlier projections had indicated a more
stringent requirement, in the range of one part in 10^5^. NSTX has also
conducted similar experiments, using the new trim control coil
capability of the device. JT-60U, which is installing iron inserts in
the toroidal field coils to better symmetrize its toroidal magnetic
field, will also add to this database.

*3.6. Energetic Particles, Heating, and Current Drive*

The combined capabilities of the U.S. facilities lead the world in the
ability to vary the fast particle energy over a wide parameter range, in
unique diagnostics, and in theoretical modeling. The normalized
gyro-radius of the fast particles (ρ∗~fast~) varies from low values on
C-Mod to very high on NSTX.  In addition, NSTX has a very wide operating
space in terms of V~fast~/V~Alfven~ and β~fast~/β~tot,~ with a
comprehensive diagnostic set including the motional Stark effect
operating in the low magnetic field regime; its operating space contains
the predicted ITER operating range in terms of these two parameters.
C-Mod has active MHD spectroscopy diagnostics, and DIII-D has a
comprehensive diagnostic set, including a new fast ion diagnostic and
the ability to vary the fast ion population. As a combined package, the
U.S. machines provide important information for ITER.

In most cases, the density and energy of fast particles are related to
the auxiliary heating and current drive scenarios. The U.S. facilities
can examine mixed regimes with neutral beam and radio frequency waves
(DIII-D and NSTX) and with radio frequency waves only (C-Mod).
Furthermore, the U.S. has a strong capability for modeling these results
and developing validated, predictive models for ITER, particularly with
electron cyclotron current drive in DIII-D and fast wave current drive
in DIII-D and NSTX, and both ion cyclotron heating and lower hybrid
current drive on C-Mod. Adding to the physics understanding of radio
frequency heating and current drive, NSTX will also carry out
experiments with electron Bernstein waves, and C-Mod with lower hybrid
waves.

*3.7. Advanced and Hybrid Operational Modes*

The U.S. is a major contributor, with ASDEX-U, JET, and JT-60, in the
development of advanced tokamak operational modes. DIII-D has obtained
highly shaped, double-null diverted plasmas at high normalized pressure
β~n~ and high confinement (H-factor). C-Mod has achieved the highest
absolute pressure plasmas in the world, and NSTX has the highest
normalized pressure (i.e., relative to the magnetic field pressure) with
its strong plasma shaping and broad pressure profile. All of these
advanced operational modes are possible directions for improving the
tokamak concept. The U.S. program has emphasized tokamak innovations and
has already made improvements in the baseline ITER design. Several of
the advanced operational modes are being developed to a sufficient
extent that they will eventually be part of the ITER design. In
addition, DIII-D and ASDEX-U have led the effort to obtain "hybrid"
discharges, which rely less strongly on inductive current and have
parameters that lie between the conventional ITER and an advanced
tokamak. If realized, these operating modes would provide easier access
to high-Q plasmas in ITER with lower plasma current.

4\. What Would Be Lost, and Recommendation

The three U.S. experimental programs have a superb track record of
collaborative discovery and innovation within the U.S. and within the
international program. They have the most extensive set of diagnostics,
heating and fueling systems, and control systems, as well as a strong
connection to theory and superb modeling capability. Alcator C-Mod is
the world's only diverted, high field tokamak, featuring all radio
frequency heating and high power flux to a unique metal wall/ divertor.
DIII-D is the best equipped and most flexible tokamak in the world. NSTX
has the most capabilities of any spherical torus facility, and it can
study electromagnetic effects at ultra-high β values. In total, they
have a unique set of capabilities for a coordinated effort to explore
advanced plasma science relevant to fusion in the range of all key
parameters -- except machine size and alpha heating power. Moreover,
each program plays a major role in the education of the next generation
of fusion energy scientists and engineers.

The charge Question #4 (part 2) -- "What research opportunities would be
lost by shutting down one of the major facilities \[during the next five
years\]?" -- is addressed in terms of the U.S. fusion energy sciences
program. In Sec. 4.1, this Question is treated from the perspective of
the generation of essential knowledge for fusion energy sciences, in
terms of goals and accomplishments that are expected for scientific
campaigns. In Sec. 4.2 this question is treated from the perspective of
each of the three major U.S. facilities. In conclusion, Sec. 4.3
presents the Panel's central recommendation. The Panel notes that it is
important for the U.S. to be viewed as a major player in the world
program in order for the U.S. to effect the science on and reap full
benefit from ITER, and that only a folio of good research results buys a
seat at that sort of decision table.

4.1. Scientific Campaign Goals and Expected Accomplishments

For each of the scientific campaigns identified in the 2005 Program
Priorities Report^1^ we summarize the principal elements related to
accomplishments that are expected during the next five years, for a
constant level of effort. In some of these elements, good progress has
been made already. Cancellation of one or more facilities
(parenthetically identified by name, with "All" meaning all three of the
facilities) would lead to either the loss of an element or the weakening
of a unique coordinated effort.

[Macroscopic Plasma Physics]{.underline} *(cf. this report Sec. 2.1):*

Substantial progress will be achieved in understanding the role of
magnetic field structure on plasma confinement, both for fundamental
science and in direct preparation for ITER \[**All**\]. Experiments to
understand pressure limits in rotating plasmas with resistive walls
should be completed, and studies of active stabilization without
rotation will be underway \[**DIII-D, NSTX**\]. The understanding of
neoclassical tearing instabilities and edge localized modes will be
complete and documented, including methods of control and suppression
\[**All**\]. An understanding will be developed of mode stabilization
due to plasma rotation, as well as the physical processes responsible
for increased drag due to plasma modes **\[DIII-D, NSTX\]**. Studies
will be performed of different types of plasma self-organization and
their interaction with external control methods \[**All**\]. Methods
will be developed to sustain the plasma duration, including "hybrid
operating modes" for ITER **\[All\].** Scientific issues associated with
the integration of high plasma pressure, good confinement, and efficient
sustainment of plasmas, including bootstrap current, will be
investigated \[**All**\]. The mechanisms for relaxation and reconnection
of the magnetic field will be identified \[**DIII-D, NSTX**\].

[Multi-scale Transport Physics]{.underline} *(cf. this report Sec.
2.2):*

The goal of developing a predictive capability for ion thermal transport
is within reach \[**All**\]. Turbulence-driven electron cross-field
thermal transport remains to be understood at the fundamental level. As
with ion transport, studies over a parameter range broader than that
accessible by one machine will be required. The three U.S. facilities
are sufficiently distinct to provide this range, and are experienced in
carrying out the necessary comparisons. Small-scale turbulence
diagnostics are being implemented to explore the role of turbulence on
electron thermal transport, and within ten years the full wavelength and
frequency ranges of the dominant turbulence should be identified. Loss
of any of the three facilities will severely compromise this research
area \[**All**\]. Particle and momentum (rotation) transport studies are
less mature, and there are known surprises that must be understood;
within five years, some convergence can be expected on which physical
mechanisms are most important \[**All**\]. A more complete understanding
of the conditions and thresholds for the formation of edge and core
transport barriers, and their dynamics, will be obtained. The
expectation is that transport barriers can then be used as a tool to
control the levels of core transport \[**All**\]. Within ten years it
will be possible to compare theoretical and numerical models with
experimental data on zonal flows and their effects \[**All**\]. Large
scale flows in plasmas without dominant external flow drive will be
documented and analysis should enable predictions for ITER \[**All**\].

[Plasma-Boundary Interfaces]{.underline} *(cf. this report Sec. 2.3):*

First-principles models of the edge pedestal will be available to
reproduce many of the measured characteristics of the plasma boundary in
one or more fusion experiments \[**All**\]. Experiments will be carried
out to validate these plasma-boundary models for ITER \[**All**\]. The
physics of edge localized modes and their impact on the scrape-off layer
will be largely understood, enabling the development of new mitigation
techniques and operational regimes to minimize their impact \[**All**\].
The fundamental characteristics of scrape-off layer turbulence in
high-confinement plasmas will be determined \[**All**\]. There should be
sufficient data to quantify the major impurity sources, including
divertor and first wall erosion and deposition rates \[**All**\]. The
distribution of tritium that is expected to be trapped in carbon-based
plasma-facing components will be largely understood in terms of plasma
conditions observed in present experiments. Candidate techniques for
tritium removal will be developed in carbon-based components \[**DIII-D,
NSTX**\]. Operation with potentially low tritium retention will be
performed with a purely metallic wall \[**C-Mod**\]. Alternative
high-heat flux components such as liquid lithium divertors will be
qualified \[**NSTX**\]; tungsten brush divertor components and all-metal
walls will be utilized \[**C-Mod**\].

[Waves and Energetic Particles]{.underline} *(cf. this report Sec.
2.4):*

Externally launched waves are the only controllable means of providing
core plasma heating and the necessary precision control of the current
profile in advanced tokamak scenarios in ITER and in future experiments.
Considerable advances are expected in developing the fundamental physics
models for wave coupling, propagation, absorption, and plasma responses
that will enable coupling on critical time scales with plasma stability
and transport models. Experiments with lower hybrid and ion cyclotron
waves \[**C-Mod**\], electron and ion cyclotron waves \[**DIII-D**\],
fast waves and electron Bernstein waves \[**NSTX**\] will validate these
models. Improved understanding of radio frequency sheath effects and
antenna loading will be achieved \[**All**\], and with an ITER-like
scrape-off layer \[**C-Mod**\]. Wave heating and current drive
simulations will be benchmarked in support of ITER \[**C-Mod,
DIII-D**\]. Prior to the advent of full burning plasma operation in
ITER, progress is expected in understanding the behavior of energetic
particles (generated by auxiliary heating) and of the unstable waves
that can be excited by them with the use of existing facilities, and in
improving theory and simulations for energetic particle dynamics
\[**All**\].

[Fusion Engineering Science]{.underline} *(cf. this report Sec. 2.5):*

Considerable progress is expected in developing the knowledge base
required to determine performance limits and identify innovative
solutions for the plasma-facing components and divertor materials
\[**All**\]. Advances will be made in fueling, wave heating, current
drive, and plasma control systems \[**All**\]. Support will be provided
for the construction and reliable operation of ITER, including
disruption mitigation \[**C-Mod, DIII-D**\] and support for a candidate
Component Test Facility \[**NSTX**\].

[Burning Plasmas]{.underline} *(cf. this report Sec. 3):*

The three U.S. machines taken together provide a wide set of parameters
to determine relevant physics parameters for ITER and future burning
plasma devices \[**All**\]. The physics of the edge pedestal is critical
for ITER, and the three machines all contribute to this ongoing effort
\[**All**\]. Control of exhaust heat and particles is being approached
with different wall materials in the three machines \[**All**\].
Experiments on the understanding and control of ELMs, a short burst of
heat flux, will be undertaken on all three machines, along with
coordinated scaling experiments on international machines \[**All**\].
The physics of plasma rotation and its role in MHD stability will be
investigated; this is important in determining whether ITER will be
stable to various modes such as the resistive wall mode \[**DIII-D,
NSTX**\]. Experiments on facilities with no apparent torque input when
only radio frequency waves are used for heating and current drive
\[**C-MOD**\] will be compared with experiments in facilities in which
the torque can be varied \[**DIII-D, NSTX**\].

4.2. What Research Opportunities Would Be Lost

Here we address the charge Question #4b specifically in reference to
each of the three major U.S. facilities.

**[C-Mod]{.underline}** is distinguished by the following salient
characteristics: (1) it operates at higher magnetic fields than any
other existing divertor tokamak, and over a range that spans the ITER
magnetic field value; (2) it has all-metallic (high atomic number)
tungsten and molybdenum wall armor and divertor plates; and (3) its
non-inductive heating and current drive are supplied entirely by
flexible radio frequency techniques. C-Mod is also distinguished by its
location at a major research university and, as such, hosts the largest
number of graduate and undergraduate students.

*The loss of C-Mod would:*

-   Eliminate studies of tritium fuel retention and power flow to the
    wall and divertor in ITER-relevant edge conditions, which have
    potentially serious consequences for ITER.

-   Eliminate ITER-relevant tests of thermal, particle and momentum
    transport with radio frequency heating and current-drive techniques
    in plasmas with the ITER-like characteristics of ions and electrons
    coupled at the same temperature through collisions, no core external
    momentum-drive, and high-pressure edge conditions.

-   Compromise development of ion cyclotron and lower hybrid RF
    capabilities for controlled plasma heating and edge current drive
    essential for advanced tokamak scenarios in ITER and future burning
    plasma facilities.

**[DIII-D]{.underline}** is the best equipped -- in terms of
diagnostics, heating and control systems -- and most flexible tokamak
--in terms of shaping capability -- in the world. It has an outstanding
record of contributing to plasma science and technology, and to the
development of advanced scenarios for tokamak operation that offer the
possibility for ITER to achieve its goals at reduced plasma current. Its
capabilities have been enhanced, and this excellent program is expected
to continue producing centrally important information for the
advancement of science and the magnetic fusion program.

*The loss of DIII-D would:*

-   Eliminate a world-class program that contributes to understanding
    turbulent transport, fast ion instabilities, pedestal and divertor
    physics, and mode stabilization.

-   Greatly reduce U.S. leadership in the world-wide development of a
    high β, high bootstrap current, "steady-state" plasma and of hybrid
    operating scenarios for ITER.

-   Cede the U.S. position of strength in electron cyclotron current
    drive for current profile and MHD stability control.

**[NSTX]{.underline}** has exceptional performance and diagnostic
capability among spherical tori, with highly flexible shaping capability
and the ability to access ultra-high plasma β values. The facility
provides an operational regime for key science theories to be tested and
confirmed, providing understanding for tokamak fusion devices in
general.

*The loss of NSTX would:*

-   Eliminate not only the U.S. leadership position in the world for
    device capability and research on spherical tokamaks, but also
    eliminates the U.S. ability to guide and contribute to spherical
    torus research at the "proof-of-principle" level.

-   Eliminate numerous important experiments such as Bernstein wave
    current drive, plasma start-up in a proof-of-principle scale
    experiment, ultra-high β operation, and fast ion measurements.

-   Eliminate U.S. leadership in high β spherical tokamak research, and
    the U.S. independence to pave the path to the future strategic
    option of constructing a Component Test Facility based on the
    spherical torus.

4.3. Recommendation

In a February 2005 report titled *'The Knowledge Economy: is the United
States losing its competitive edge?,'* the Task Force on the Future of
American Innovation "developed a set of benchmarks to assess the
international standing of the U.S. in science and technology. These
benchmarks in education, the science and engineering workforce,
scientific knowledge, innovation, investment and high-tech economic
output reveal troubling trends across the research and development
spectrum. The U.S. still leads the world in research and discovery, but
our advantage is rapidly eroding, and our global competitors may soon
overtake us."^5^

In the particular area that is the subject of this report, that of the
scientifically interesting and broadly important topic of magnetic
fusion energy research, the Facility Panel finds that the U.S at present
holds a position of international strength and leadership. The next
major magnetic fusion research facility will be the offshore burning
plasma experiment, ITER.  Fostered by the recent decision over its ITER
site, interest in ITER is rapidly growing in the international research
community. The loss of any of the three major U.S. toroidal fusion
facilities would fundamentally jeopardize the ability of U.S.
researchers to perform relevant fusion research, and thus would
undermine the current U.S. position of international excellence.

The three major U.S. magnetic fusion facilities -- C-Mod, DIII-D, and
NSTX -- represent a massive investment of talent, intellect, and
finances in tackling the key issues of toroidal confinement. Each of
these facilities has made seminal contributions to the development of
toroidal confinement and to the fundamental science and technology that
undergird it. The wealth of discoveries and the generation of knowledge
made possible by the three coordinated U.S. facilities has enabled the
U.S. to be an effective presence among the larger foreign programs
involved in ITER. Premature closure of one of these three major
facilities would seriously compromise the effectiveness of the U.S.
fusion program internationally and also the U.S. ability to advocate
future proposals for advanced performance scenarios that could lead to a
more economically competitive high-power-density fusion system.

**The Panel's recommendation is that the three major United States
toroidal magnetic fusion facilities continue operation to conduct
important unique and complementary research in support of fusion energy
sciences and ITER.**

5\. References

1\. *Scientific Challenges, Opportunities, and Priorities for the U.S.
Fusion Energy*

*Sciences Program: A Report to the Fusion Energy Sciences Advisory
Committee,*

(Apr 2005);
<http://www.ofes.fusion.doe.gov/more_html/FESAC/PP_Rpt_Apr05R.pdf> .

2\. *Fusion Simulation Project: Integrated Simulation & Optimization of
Fusion Systems,*

(Dec 2002);
<http://www.ofes.fusion.doe.gov/News/FESAC/FSP_report_Dec9.pdf> .

3\. Presentations at the 20th IAEA Fusion Energy Conference, Vilamoura,
Portugal,

(Nov 1-6, 2004); http://
[www.cfn.ist.utl.pt/20IAEAConf/Announcement.htm](http://www.cfn.ist.utl.pt/20IAEAConf/Announcement.htm)
.

4\. *Device schedules/Program elements in 2005*, compiled by O. Gruber.
3rd Joint

Workshop of IEA Implementing Agreement on "Implementation of ITPA

Coordinating Research Recommendations," Eynsham Hall, U.K. (Dec 8-10,
2004).

5\. *The Knowledge Economy: is the United States Losing its Competitive
Edge?*

*Benchmarks of our Innovation Future*, Task Force on the Future of
American

Innovation (Feb 2005):
[http://www.futureofinnovation.org](http://www.futureofinnovation.org/)
.

Appendix 1. Tables of Characteristics of the Three Major

U.S. Toroidal Magnetic Fusion Facilities

In this Appendix are provided tabulated values of the characteristic
plasma (Table A1-1) and machine (Table A1-2) parameters for the three
major U.S. toroidal fusion facilities. In Table A1-3, parameters for a
subset of international facilities are compared with those for ITER
\[http://www.iter.org\].

**Table A1-1. Plasma Parameters \[present {future}\].**

+-----------------+-----------------+---------------+------------------+
| Plasma          | Alcator C-Mod   | DIII-D        | ######## NSTX    |
| Parameters      |                 |               |                  |
+-----------------+-----------------+---------------+------------------+
| \<n\> ~max~     | 6.0             | 3.0           | 0.7              |
| (x10^20^ m^-3^) |                 |               |                  |
+-----------------+-----------------+---------------+------------------+
| T~i~(o) ~max~   | 5.6             | 27            | 2.5              |
| (keV)           |                 |               |                  |
+-----------------+-----------------+---------------+------------------+
| T~e~(o) ~max~   | 6.0             | 16            | 4.1              |
| (keV)           |                 |               |                  |
+-----------------+-----------------+---------------+------------------+
| W~tot~^max^     | .25             | 4.2           | 0.43             |
| (MJ)            |                 |               |                  |
+-----------------+-----------------+---------------+------------------+
| n/n~GW~         | 0.7             | 1.4           | 1.1              |
| (H89[           |                 |               |                  |
| \>]{.underline} |                 |               |                  |
| 1.6)            |                 |               |                  |
+-----------------+-----------------+---------------+------------------+
| Max H89P for    | 2.1             | 3.2           | 2.5              |
| 5τ~E~           |                 |               |                  |
+-----------------+-----------------+---------------+------------------+
| T~i~/T~e~       | 0.5 -- 1.9      | 0.15-5        | 0.6 -- 1.7       |
| *\[range\]*     |                 |               |                  |
+-----------------+-----------------+---------------+------------------+
| τ~Ε~∗ν~ei~      | 0.5 - 100       | 0.2-200       | 4 - 14           |
| *\[range\]*     |                 |               |                  |
+-----------------+-----------------+---------------+------------------+
| ν\* *\[vol. av. | 0.007-23;       | 0.001-3;      | 0.1              |
| value;*         |                 |               | \[ρ=0--0.6\];    |
|                 | 0.5-20          | 0.02-20       |                  |
| *pedestal top   |                 |               | 0.5              |
| value\]*        |                 |               |                  |
+-----------------+-----------------+---------------+------------------+
| 1/ρ\* *\[vol.   | 170-500;        | 50-600;       | LFS avg. 50 --   |
| av. value;*     |                 |               | 100;             |
|                 | 250-1000        | 120-650       |                  |
| *pedestal top   |                 |               | 50               |
| value\]*        |                 |               |                  |
+-----------------+-----------------+---------------+------------------+
| \<β\> % max     | 1.4             | 12.5          | 40 {50}          |
+-----------------+-----------------+---------------+------------------+
| β~N~ (max)      | 1.78            | 6.0           | 7.2 {8}          |
+-----------------+-----------------+---------------+------------------+
| β~N~H~89~ (for  | 3.1             | 9 {12}        | 12               |
| 1 τ~R~)         |                 |               |                  |
+-----------------+-----------------+---------------+------------------+
| εβ~pol~^max^    | .77             | 1.8           | 1.6              |
+-----------------+-----------------+---------------+------------------+
| f~BS~^max^      | .22 {.7}        | 0.85          | 0.6              |
+-----------------+-----------------+---------------+------------------+
| β~fast~/β~tot~  | 0.07-0.73       | 0 -- 0.6      | 0 -- 0.75        |
| *\[range\]*     |                 |               |                  |
|                 |                 |               | (β~fast~ 0 --    |
| (volume         |                 |               | 25%)             |
| averaged β)     |                 |               |                  |
+-----------------+-----------------+---------------+------------------+
| V~              | 0.2-0.8         | 2.5           | 1-5              |
| fast~/V~Alfven~ |                 |               |                  |
+-----------------+-----------------+---------------+------------------+
| \<Pressure\>    | 0.18            | 0.125         | 0.022            |
| MPa             |                 |               |                  |
+-----------------+-----------------+---------------+------------------+
| *f~NI~ %;*      | {100 %, 10}     | 100 %; 9      | 70 %; 4 {100%,   |
| (τ~pulse~) /    |                 |               | 10}              |
| (τ~R~)\]        |                 |               |                  |
+-----------------+-----------------+---------------+------------------+
| SOL: density;   | 1x10^18^ to     | 0.4-3.4       | 0.1 -- 1 x       |
|                 | 1x10^20^ m^-3^; | x10^20^       | 10^19^ cm^-3^;   |
| collisionality  |                 | m^-3^;        |                  |
| range           | 10-600          |               | 2 - 60           |
|                 |                 | 0.1-60        |                  |
+-----------------+-----------------+---------------+------------------+
| P               | Up to 80%       | 70%           | \~ 30%           |
| rad~div~/P~SOL~ |                 |               |                  |
| (H89[           |                 |               |                  |
| \>]{.underline} |                 |               |                  |
| 1.6)            |                 |               |                  |
+-----------------+-----------------+---------------+------------------+

**Table A1-2.** **Machine Parameters \[present {future}\].**

+----------------+-----------------+----------------+-----------------+
| Machine        | Alcator C-Mod   | DIII-D         | NSTX            |
| Parameters     |                 |                |                 |
+----------------+-----------------+----------------+-----------------+
| R (meters);    | 0.61-0.74;      | 1.49-1.88;     | 0.8-1.0;        |
| R/a            | 2.8-3.6         | 2.5-4.5        | 1.27-1.6        |
| *\[range\]*    |                 |                |                 |
+----------------+-----------------+----------------+-----------------+
| κ;             | 0.9 -- 1.85;    | 1.15-2.6;      | 1.5-2.7;        |
| δ *\[range\]*  | 0 - .84         | -0.1-1.0       | 0.2-0.8         |
+----------------+-----------------+----------------+-----------------+
| B              | 2-8             | 0.5-2.2        | 0.3-0.45 {0.6}  |
| (Tes           |                 |                |                 |
| la)*\[range\]* |                 |                |                 |
+----------------+-----------------+----------------+-----------------+
| I (MA)         | 2.05            | 3.0            | 1.5             |
| *\[max\]*      |                 |                |                 |
+----------------+-----------------+----------------+-----------------+
| Plasma Vol.    | 1.0             | 24             | 14              |
| (max, m^3^)    |                 |                |                 |
+----------------+-----------------+----------------+-----------------+
| τ~pulse~ (sec) | 1 @ 8 Tesla;    | 6 {10} @ full  | 1.5 @ B~T~ =    |
| *\[max at full |                 | field;         | 0.45T;          |
| field;*        | 3 {5} @ 5.3     |                |                 |
|                | Tesla           | 15 {25} @ 75%  | 3.5 @ B~T~ =    |
| *at 75%        |                 | nominal        | 0.34T           |
| nominal\]*     |                 |                |                 |
+----------------+-----------------+----------------+-----------------+
| Heating:       | ICRF - 8 MW;    | ^+^NBI 17.5    | NBI 7.5MW;      |
|                |                 | {20};          |                 |
| -   Type       | LHRF -- 3 {4}   |                | HHFW 6MW;       |
|                | MW              | ^+^NBI 12.25   |                 |
| -   MW         |                 | {14} @ 5 s;    | {EBW 4MW}       |
|                |                 |                |                 |
|                |                 | ^+^ECH 4.5     |                 |
|                |                 | {6.75};        |                 |
|                |                 |                |                 |
|                |                 | ^+^FW 3 {6}    |                 |
+----------------+-----------------+----------------+-----------------+
| Current Drive: | LHCD -- 3 {4}   | ^+^NBCD 8.75   | NBI 7.5MW;      |
|                | MW;             | {10.5}         |                 |
| -   Type       |                 |                | HHFW 6MW;       |
|                | MCCD -- 8 MW;   | co / 3.5 ctr @ |                 |
| -   MW         |                 | 5 s;           | CHI 0.4MA       |
|                | FWCD -- 8 MW;   |                | {0.5MA};        |
|                |                 | +ECCD 4.5      |                 |
|                | ICCD -- 8 MW;   | {6.75}         | {EBW 4MW}       |
|                |                 |                |                 |
|                |                 | co/ctr;        |                 |
|                |                 |                |                 |
|                |                 | ^+^FWCD 3 {6}  |                 |
|                |                 | co/ctr         |                 |
+----------------+-----------------+----------------+-----------------+
| Nearby         | Inner           | Yes            | Yes,            |
| Conducting     | (high-field     |                | w/a~plasma~ \~  |
| wall           | side): close;   |                | 1.3 -1.5        |
|                |                 |                |                 |
| *\[yes or      | Outer           |                |                 |
| no\]*          | (low-field      |                |                 |
|                | side): far;     |                |                 |
|                |                 |                |                 |
|                | Effects of      |                |                 |
|                | outboard        |                |                 |
|                | structures      |                |                 |
|                | (limiters,      |                |                 |
|                | antennas):      |                |                 |
|                | {TBD}           |                |                 |
+----------------+-----------------+----------------+-----------------+
| Error Field    | Yes             | Yes            | Yes             |
| Correction     |                 |                |                 |
| coils          |                 |                |                 |
|                |                 |                |                 |
| *\[yes or      |                 |                |                 |
| no\]*          |                 |                |                 |
+----------------+-----------------+----------------+-----------------+
| RWM feedback   | No              | Internal and   | External        |
| coils          |                 | External       |                 |
|                |                 |                |                 |
| *\[internal,   |                 |                |                 |
| external, or   |                 |                |                 |
| none\]*        |                 |                |                 |
+----------------+-----------------+----------------+-----------------+
| NTM            | MCCD, ICCD, LH  | ECCD: proven   | {EBW}           |
| stabilization  |                 | stabili-zation |                 |
| (e.g. EC, LH,  |                 | of 3/2 & 2/1   |                 |
| EBW)           |                 | NTM            |                 |
+----------------+-----------------+----------------+-----------------+
| First Wall and | molydenum       | Carbon,        | CFC/Graphite,   |
|                | {tungsten}      | periodic       | Li coating      |
| Divertor       |                 | boronization   |                 |
| Material       |                 |                | {Lithium        |
|                |                 |                | target}         |
+----------------+-----------------+----------------+-----------------+
| Divertor       | LSN, USN, DN,   | LSN (pumped),  | Double null;    |
|                | inner-wall      | USN (pumped),  |                 |
| *\[shape,      | limited, lower  | DN (pumped)    | upper, lower    |
| pumped or      | nose limited    |                | single null     |
| not\]*         | {pumped, 2005}  |                |                 |
|                |                 |                | {lith           |
|                |                 |                | ium}{Cryo-pump} |
+----------------+-----------------+----------------+-----------------+
| Power/R (MW/m) | 9.0 {13.4}      | 14.7 {19.3}    | 16 {20}         |
+----------------+-----------------+----------------+-----------------+
| P              | 0.86 {1.3}      | 0.46 {0.60}    | 0.4 {0.5}       |
| ower/Surf.Area |                 |                |                 |
| (MW/m^2^)      |                 |                |                 |
+----------------+-----------------+----------------+-----------------+
| Power/Vol.     | 6.7 {10.0}      | 1.3 {1.7}      | 1.1 {1.3}       |
| (MW/m^3^)      |                 |                |                 |
+----------------+-----------------+----------------+-----------------+
| Fueling        | Low pressure    | Pellets:       | Gas injectors:  |
|                | gas; high       | outer, top,    | LFS, HFS,       |
| *\[list of     | pressure gas;   | inner          | divertor        |
| options\]*     | D~2~ pellets;   | (radial),      |                 |
|                | Li pellets      | inner (45      | Supersonic Gas  |
|                |                 | deg), top (ELM | Injector        |
|                |                 | pacemaking);   |                 |
|                |                 |                | NBI             |
|                |                 | Gas: divertor, |                 |
|                |                 | inner wall,    | Solid pellet    |
|                |                 | outer, through | injector (Li,   |
|                |                 | divertor       | B, C, {doped})  |
|                |                 | baffles (top & |                 |
|                |                 | bottom,        | {deuterium      |
|                |                 | toroidally     | pellet          |
|                |                 | symmetric)     | injector}       |
|                |                 |                |                 |
|                |                 |                | {CT injector}   |
+----------------+-----------------+----------------+-----------------+
| Facility       | \$180 M         | \$694 M        | \$ 464 M        |
| Replacement    |                 |                |                 |
| Value, in      |                 |                |                 |
| current \$ US  |                 |                |                 |
+----------------+-----------------+----------------+-----------------+

^+^ Power launched into plasma

**Table A1-3. Comparison with International Machines.**

+-------+--------+-------+--------+--------+-------+--------+-------+
| Param | A      | FTU   | JET    | JT-60U | MAST  | TCV    | ITER  |
| eters | SDEX-U |       |        |        |       |        |       |
+-------+--------+-------+--------+--------+-------+--------+-------+
| R     | 1.65;  | 0     | 2.9;   | 3.2;   | 0.85; | 0.89;  | 6.2;  |
| (m);  | 3.2    | .935; | 3.1    | 4.0    | ≥ 1.3 | 3.6    | 3.1   |
| R/a   |        | 3.2   |        |        |       |        |       |
+-------+--------+-------+--------+--------+-------+--------+-------+
| κ; δ  | Up to: | 1     | 1.9;   | 1.8;   | 2.5;  | 1-2.8; | 1.7;  |
|       | 1.80;  | .05,; | 0.5    | 0.4    | 0.5   |        | 0.35  |
|       |        | 0     |        |        |       | -0     |       |
|       | 0.55   |       |        |        |       | .6-0.6 |       |
+-------+--------+-------+--------+--------+-------+--------+-------+
| B     | 1.5-3  | 8     | 3.8    | 4.8    | 0.52T | 1.43 @ | 5.3   |
| (T    |        |       |        |        | @ R   | R      |       |
| esla) |        |       |        |        |       |        |       |
+-------+--------+-------+--------+--------+-------+--------+-------+
| I     | Up to: | 1.6   | 4.5    | 5.0    | 1.4   | 0.8    | 15    |
| (MA)  | 1.4    |       |        |        |       |        |       |
+-------+--------+-------+--------+--------+-------+--------+-------+
| W     | Up to: | 0.14  | 17 MJ  | 11 MJ  | \~    | 40 kJ  | 350   |
| ~tot~ | 1.5 MJ | MJ    |        |        | 160   |        | MJ    |
|       |        |       |        |        | kJ    |        |       |
+-------+--------+-------+--------+--------+-------+--------+-------+
| He    | NBI,   | LH,   | NBI,   | NBI,   | NBI,  | X-     | NB,   |
| ating | ICRH,  | ECRH, | LH,    | LH,    | {EBW} | 2^nd^, | ICH,  |
| Type  | ECRH   | IBW   | ICH    | ECH    |       | X      | ECH,  |
|       | {LH?}  |       |        |        |       | -3^rd^ | LH    |
|       |        |       |        |        |       | ECRH   |       |
+-------+--------+-------+--------+--------+-------+--------+-------+
| Cu    | NBCD,  | LH (8 | NBCD,  | NBCD,  | NBCD, | ECCD   | NB,   |
| rrent | ECCD   | GHz)  | LHCD,  | LHC    | {E    |        | ECH,  |
| Drive | {      |       | ICCD   | D,ECCD | BWCD} |        | LH    |
|       | LHCD?} |       |        |        |       |        |       |
+-------+--------+-------+--------+--------+-------+--------+-------+
| First | W; C   | Inco  | C; Be  | C,     | C     | C (92% | Be    |
| Wall; |        | nel + |        | pe     |       | cov    | wall; |
|       |        | Mo    |        | riodic |       | erage) |       |
| Div.  |        | lim   |        | boroni |       |        | CFC,  |
| Mat.  |        | iters |        | zation |       |        | W div |
+-------+--------+-------+--------+--------+-------+--------+-------+
| RWM   | {Pl    |       | No     | No     | No    | No     | No    |
| co    | anned} |       |        |        |       |        | {Yes} |
| ntrol |        |       |        |        |       |        |       |
+-------+--------+-------+--------+--------+-------+--------+-------+
| ν\*   | 0.04   | 0     | 0      | 0      | 0     | 30 --  | 0.02  |
| range | -- 3   | .2-10 | .001-2 | .001-3 | .02 - | 0.08   |       |
|       | (Vol.  |       |        |        | 0.2   |        |       |
|       | ave    |       |        |        |       | (T~e~= |       |
|       | raged) |       |        |        |       | 0.5    |       |
|       |        |       |        |        |       | -5keV, |       |
|       |        |       |        |        |       |        |       |
|       |        |       |        |        |       | n~e~   |       |
|       |        |       |        |        |       | =4-1x1 |       |
|       |        |       |        |        |       | 0^13^/ |       |
|       |        |       |        |        |       | cm^3^) |       |
+-------+--------+-------+--------+--------+-------+--------+-------+
| 1/ρ\* | 100    | 450   | 350    | 400    | \~ 60 | ≥80    | 800   |
| ty    |        |       |        |        |       |        |       |
| pical |        |       |        |        |       | (Ti ≤  |       |
|       |        |       |        |        |       | 1keV)  |       |
+-------+--------+-------+--------+--------+-------+--------+-------+
| \<β\> | Up to  | 0.3   | 4.3    | 2.7    | 16%   | 2.5    | 2.5   |
| % max | 5 in   |       |        |        |       |        | {3.9} |
|       | I~p    | (v    |        |        |       |        |       |
|       | ~-ramp | olume |        |        |       |        |       |
|       | down;  | av.)  |        |        |       |        |       |
|       | \~3.5  |       |        |        |       |        |       |
|       | SS     |       |        |        |       |        |       |
+-------+--------+-------+--------+--------+-------+--------+-------+
| β~N,~ | Up to  | 0.9   | 3.9    | 4.8    | 5.3   | 3.2    | 1.8   |
| max   | 5 in   |       |        |        |       |        | {2.8} |
|       | I~p    |       |        |        |       |        |       |
|       | ~-ramp |       |        |        |       |        |       |
|       | down;  |       |        |        |       |        |       |
|       | \~3.5  |       |        |        |       |        |       |
|       | SS     |       |        |        |       |        |       |
+-------+--------+-------+--------+--------+-------+--------+-------+
| β~fa  | 0.0    | 0.0   | 0.2    | 0.4    | \~ 15 | ≤ 25   | 0.1   |
| st~/β | 5-0.3; |       |        |        |       |        |       |
| ~tot~ | typ.   |       |        |        |       | (fast  |       |
| (typ  | impr.  |       |        |        |       | elec   |       |
| ical) | H-mode |       |        |        |       | trons) |       |
|       | with   |       |        |        |       |        |       |
|       | 0      |       |        |        |       |        |       |
|       | .1-0.2 |       |        |        |       |        |       |
+-------+--------+-------+--------+--------+-------+--------+-------+
| V~    | 0.5-1  | 0.0   | 1.0    | 2.0    | \~    | N/A    | 1.8   |
| fast~ | for    |       |        |        | 0.8   |        |       |
| /V~Al | NBI    |       |        |        | --    |        |       |
| fven~ | par    |       |        |        | 2.5   |        |       |
|       | ticles |       |        |        |       |        |       |
+-------+--------+-------+--------+--------+-------+--------+-------+
| Div   | Pol.   | N/A   | LSN    | LSN    | LSND, | SNL,   | LSN   |
| ertor | div    |       | pumped | pumped | USND, | SNU,   | p     |
| *(s   | with   |       |        |        | DND,  | DND,   | umped |
| hapes | cryo-p |       |        |        | {pum  | No     |       |
| wi    | umping |       |        |        | ping} |        |       |
| th/wo | in lwr |       |        |        |       |        |       |
| pump  | SN ∀   |       |        |        |       |        |       |
| ing)* | s      |       |        |        |       |        |       |
|       | hapes, |       |        |        |       |        |       |
|       | upper  |       |        |        |       |        |       |
|       | SN     |       |        |        |       |        |       |
|       | w      |       |        |        |       |        |       |
|       | ithout |       |        |        |       |        |       |
|       | p      |       |        |        |       |        |       |
|       | umping |       |        |        |       |        |       |
+-------+--------+-------+--------+--------+-------+--------+-------+
| P~S   | Up to: | 2.6   | 9      | 12     | \~ 6  | 5.1    | 1     |
| EP~/R | 15     |       |        |        |       |        | 2{14} |
|       |        |       |        |        |       |        |       |
| (     |        |       |        |        |       |        |       |
| MW/m) |        |       |        |        |       |        |       |
+-------+--------+-------+--------+--------+-------+--------+-------+
| P~S   | Up to  | 0.22  | 0.2    | 0.3    | \~    | 0.26   | 0.11{ |
| EP~/S | 0.6    |       |        |        | 0.2   |        | 0.12} |
|       |        |       |        |        |       |        |       |
| (MW/  |        |       |        |        |       |        |       |
| m^2^) |        |       |        |        |       |        |       |
+-------+--------+-------+--------+--------+-------+--------+-------+
| P~CO  | Up to  | 1.55  | 0.3    | 0.4    | \~    | 5.6 -- | 0.6   |
| RE~/V | 2      |       |        |        | 0.5   | 10     | {0.8} |
| (MW/  |        |       |        |        |       |        |       |
| m^3^) |        |       |        |        |       |        |       |
+-------+--------+-------+--------+--------+-------+--------+-------+

Appendix 2. FESAC Facilities Assessment Charge Letter

> April 5, 2005

Professor Richard D. Hazeltine, Chair

Fusion Energy Sciences Advisory Committee

The University of Texas at Austin

Institute for Fusion Studies

1 University Station, C 1500

Austin, TX 78712-0262

Dear Professor Hazeltine:

Funding U.S. participation in the ITER project will place considerable
pressure on the Office of Science budget in future years. I am committed
to a strong base fusion research program during ITER construction but
want to be sure that the program is cost-effective. In particular, I do
not believe that we should continue running existing facilities just
because we can do more research on them but because we can do unique and
important research on them. Accordingly, I would like you to assess
characteristics of the three major U.S. toroidal magnetic fusion
facilities in the context of the international fusion programs and
determine what contributions they can make to fusion science and the
fundamental vitality of the U.S. Fusion Program during the next five
years.

I would like you to consider the following issues in your assessment:

1\. What are the unique and complementary characteristics of each of the
major U.S. fusion facilities?

2\. How do the characteristics of each of the three U.S. fusion
facilities make the U.S. toroidal research program unique as a whole in
the international program?

3\. How well do we cooperate with the international community in
coordinating research on our major facilities and how have we exploited
the special features of U.S. facilities in contributing to international
fusion research, in general, and to the ITER design specifically?

4\. How do these three facilities contribute to fusion science and the
vitality of the U.S. Fusion program? What research opportunities would
be lost by shutting down one of the major facilities?

I would like to have your assessment by the end of August.

> Sincerely,
>
> /s/
>
> Raymond L. Orbach
>
> Director

Office of Science

Appendix 3. The Three Facilities Joint Document

> \[submitted to the FESAC Facilities Panel on June 2, 2005\]

A3.1. Cover Letter

![](media/image11.pct){width="5.298611111111111in"
height="7.102777777777778in"}

A3.2. Joint Document

######## Unique Research Roles for the United States in the World Fusion Program 

**Enabled by the Combined Contributions of the Three U.S. Facilities:**

**Alcator C-Mod, DIII-D, and NSTX.**

This report identifies a set of research areas in which the
[combined]{.underline} resources of the three U.S facilities provide
world leadership in answering key scientific questions. The United
States is well positioned by having three facilities whose complementary
capabilities enable unique research contributions. Examples of those
research areas are:

**1. The role of non-axisymmetric fields in plasma performance.**

**2. The role of plasma facing components.**

**3. The physics of plasma rotation and momentum transport.**

**4. Advanced Tokamak / ST performance through off-axis non-inductive
current drive.**

**5. The role of density/collisionality in edge plasma turbulence and
divertor physics.**

**6. The physics of energetic particle and Alfvén eigenmodes.**

**7. The dependence of transport on beta**

**1. The role of non-axisymmetric fields in plasma performance.**

[Locked Modes]{.underline}

Departures from axisymmetry in the tokamak magnetic field structure
produced by the as-built coil set can destabilize non-rotating tearing
modes (locked modes) which can significantly impact plasma operation.
Such error-field induced modes can lead to degradation of confinement
and even major disruptions. The impact of these modes on the operation
of future burning plasma experiments such as ITER has been a matter of
concern; key issues are the prediction of the error-field sensitivity
(threshold perturbation) in such devices and the requirements for
corrective measures.

The U.S. program is making key contributions in this area. All three
U.S. facilities have undertaken studies of their intrinsic error fields
resulting from imperfections in fabrication and assembly of their
as-built coil sets. Each has installed non-axisymmetric coils capable of
correcting these asymmetries. Joint experiments making use of the
non-axisymmetric coils to apply known field perturbations are being
undertaken among C-Mod, DIII-D, NSTX, and JET in order to provide a
basis for extrapolative prediction (in field and size) of the symmetry
requirements for ITER. These experiments are conducted with matching
shape and normalized plasma parameters, spanning a range of size of
nearly a factor of five and a range of field of a factor of 20.
Preliminary results indicate that the required symmetry will be in the
range of a part in 10^4^, which is within the capability of the planned
ITER correction coil system. Earlier projections had indicated a more
stringent requirement, in the range of a part in 10^5^.

Ongoing experiments are continuing to investigate the origin and impact
of non-axisymmetric error fields, and the physics of locked modes. An
important issue is the extent to which correction of the
toroidally-coupled "sidebands" of the dominant mode (usually 2/1) is
required. Experiments on DIII-D have indicated that these terms have a
non-negligible impact, while C-Mod has succeeded in greatly expanding
its available operating space by correcting only the 2/1 component of
the intrinsic error field. It might be expected that sideband correction
would be even more important in the low aspect ratio ST configuration,
the investigation of which on NSTX is planned for 2005. Additional
experiments have been proposed for C-Mod and DIII-D to study the effect
of non-resonant 3D perturbations on magnetic surfaces near the
separatrix, as well as possible effects on the H-mode pedestal.

[Resistive Wall Mode]{.underline}

The pressure limit in the tokamak is set by the ideal kink mode, the
lowest order plasma driven departure from axisymmetry. This limit is of
primary importance since the fusion power is proportional to the square
of the plasma pressure. In the presence of a conducting wall, the kink
mode manifests itself as the Resistive Wall Mode (RWM) which rotates
slowly and grows on the flux diffusion time through the wall. While
transient operation above the no-wall limit was demonstrated in DIII-D
in 1994, in a major breakthrough in 2001, DIII-D showed that it was
possible to sustain operation of a tokamak above the no-wall beta limit
with a plasma rotating inside a close fitting conducting shell. The key
physics discovery was that above the no-wall beta limit, the RWM
amplified intrinsic asymmetries in the magnetic field structure of the
same helicity as the RWM. A set of coils external to the vacuum chamber
counteracted this resonant field amplification, maintained the plasma
rotation, and thus stabilized the RWM. The new wall-stabilized
normalized beta limit is found in some cases to be twice the no-wall
limit, implying possibly 4 times the fusion power and 2 times the
bootstrap current fraction.

The RWM coils on DIII-D and NSTX afford the U.S a unique opportunity to
study both the no-wall and with-wall beta limits over a range of aspect
ratio and plasma rotation speed, normalized to the intrinsic sound and
Alfvén wave speeds. ITER is seriously looking into whether such trim
coils can be added to the ITER design. Evidence of rotational
stabilization of the RWM has been seen since 2002 on NSTX, at very high
toroidal beta values. NSTX has recently added a set of external coils
for error field and resistive wall mode feedback control of the strongly
rotating, very high beta plasmas. This coil set and the existing
external and internal coil sets on DIII-D will enable both NSTX and
DIII-D to operate plasmas above the no-wall limit by means of wall
stabilization with rotation. DIII-D has a set of RWM feedback coils
inside its vacuum vessel to study direct fast feedback stabilization of
the RWM in plasmas with low rotation, as is expected in ITER. The
rotation in DIII-D will be nulled by toroidally balanced neutral beam
injection. Initial results of feedback experiments have been positive.
The coils in NSTX, similar in configuration to what is being considered
for ITER, will permit studies of the mode-plasma coupling physics and
its dependence on sound vs. Alfvén wave physics. With the planned
ability in 2005 of resonant error field correction using the newly
commissioned control coils, NSTX plans to determine the effects of
rotational stabilization of the RWM for durations longer than the plasma
current relaxation time at beta values significantly exceeding the
no-wall limit. No other tokamaks in the world have such trim coil sets
to enable exploration of the tokamak operating space above the no-wall
limit. While C-Mod has not yet extended its operating regime to beta
values above the no-wall limit, the AT program on that device has as a
long-term goal the study of optimized configurations sustained for times
much longer than the current diffusion time at high plasma temperature.
In this context, installation of RWM coils, based on the designs
perfected on DIII-D and NSTX, may prove necessary.

**2. The role of plasma facing component materials**

Current projections to ITER based on carbon Plasma Facing Component
(PFC) tokamaks indicate that ITER could be limited to tens to hundreds
of discharges before being forced to stop and remove the co-deposited
tritium from the PFC surfaces owing to a tritium inventory restriction.
Carbon divertor materials are the ITER baseline design owing to their
proven ability to handle both high steady-state and pulsed heat fluxes.
Such a limitation on ITER operation is a very important problem.

The US facility complement has a very valuable diversity of first wall
materials that is being brought to bear on the choice of ITER plasma
facing component materials. The most obvious expectation is that use of
high-Z PFCs (e.g. tungsten) would reduce the tritium retention in ITER
substantially as the theoretical limit is 0.4 T for each carbon and
\<0.01T for each tungsten in the lattice. The ITER project, while
considering such a change, is primarily concerned about the minimal
existing operational capability (e.g. compatibility with a high
performance core plasma) of such materials in the presence of high heat
loads and possible melting. Alcator C-Mod is addressing these concerns
by utilizing molybdenum as the PFC material for all in-vessel armor. To
address the need for more tungsten-specific experience C-Mod is
currently testing tungsten "brush" divertor tiles and will convert the
entire high heat flux region to tungsten. These studies will also shed
light on the concentration of high-Z impurities that can be tolerated in
the plasma core, and the approaches to maintain H-mode plasmas in the
presence of such impurities.

NSTX and DIII-D operate with carbon walls, concentrating their efforts
on reducing the T retention in C and developing T removal methods that
can be effective on the time scales needed. Possible solutions to the
carbon problem are being pursued by research on boundary plasma flows to
understand where in the divertor tokamak the codeposited layers will
form, by investigating methods of periodic removal of the codeposits
(principally baking in an oxygen environment) from those limited areas,
and by pursuit of methods to desorb the vacuum chamber wall generally
(e.g. by use of the radiation flash from mitigated disruptions). Lastly
NSTX is exploring the more speculative possibility of switching to a
liquid lithium divertor target, which would be resistant to disruption
damage, and could in principle be cleansed of tritium outside the
device. The peak divertor heat flux on NSTX can equal that predicted for
normal ITER operation, providing an excellent test-bed for such
techniques. Key questions to be answered in these tests include the heat
flux capability and the MHD stability of the liquid during ELM's and
disruptions.

There are also questions of what role PFCs and wall conditioning (e.g.
boronization) play in affecting core and divertor performance.
Boronization was introduced at least a dozen years ago and is now
applied to essentially all magnetic fusion research devices. Its
application is generally found to improve plasma operation although the
reasons are not well understood. However, it is not yet clear if
boronization can be applied in a meaningful way in ITER, casting doubt
on achieving the best confinement in next-step devices. DIII-D and NSTX
are exploring this for boron applied over carbon plasma facing
components while C-Mod is providing the equivalent experience with
high-Z. NSTX plans also to investigate the effects of lithium-coating on
carbon. All three facilities have programs in place to characterize dust
formation and transport.

Lastly, the choice of PFC material affects divertor and core
performance. When C is present in the plasma in sufficient quantities
from the PFCs less extrinsic impurities are needed to increase divertor
radiation so as to dissipate the core power efflux before it reaches the
divertor plates. C-Mod is required to add low-Z impurities to increase
the divertor radiation. In all cases the question is how to control the
level of such impurities that are needed for power dissipation without
adversely diluting the core reactivity and increasing radiation there.
The three experiments are exploring this question and the effects of
their different divertor geometries and materials, thus providing
complementary approaches to solving this problem in ways applicable to
ITER.

**3. The physics of plasma rotation and momentum transport**

The rotational stabilization of the RWM sparked increased interest in
the physics of why plasmas spin toroidally even without apparent applied
torque. In the conventional picture, the toroidal rotation is the result
of toroidal torque from tangential neutral beams balancing a momentum
loss through diffusion. Because ITER has high energy negative ion beams
with low momentum content, this conventional picture predicts rotation
speeds in ITER below but possibly near the threshold value for RWM
stabilization. However the underlying physics has not been fully
clarified. Hence it is important to study RWM stabilization with and
without rotation and to understand the connection of rotation damping to
the plasma sound and Alfvén speeds.

Other motivations for studying momentum transport and rotation include
the role of rotation in producing sheared ExB flows that stabilize
turbulence, the role of plasma flows in edge physics, and the role of
rotation in stabilizing internal plasma modes. The discovery of strongly
rotating plasmas (up to 130 km/s, Mach number 0.3) at high densities on
Alcator C-Mod with no momentum input (purely ohmic and ICRF heated)
shows that the conventional picture misses some important element of the
physics. DIII-D has also found toroidal rotation with EC heating only.
NSTX has observed strong edge flows associated directly with RF heating.
If spontaneous rotation could be understood and controlled, this might
allow for rotational RWM stabilization in ITER.

DIII-D has studied both toroidal and poloidal rotation because of the
role rotation plays in forming the radial electric field whose shear
forms both internal transport barriers and the H-mode transport barrier
(pedestal) at the edge. Such flows and transport barriers are also
studied in Alcator C-Mod and NSTX. From studies of the edge plasma
flows, Alcator C-Mod has advanced the hypothesis that the physics on the
open field lines in the SOL determines the edge rotation which sets the
boundary condition for the core plasma rotation. The role of intrinsic
plasma rotation in the H-mode power threshold in different magnetic
configurations (with the X point towards and away from the gradB drift
direction) has been demonstrated in Alcator C-Mod plasmas with no
external momentum input. Edge plasma flows from drifts and ExB flows are
studied for all three machines using the BOUT edge turbulence code and
the UEDGE fluid code.

NSTX has strong co-rotation from its neutral beam with comparable values
of Mach and Alfvén Mach numbers (due to very high beta). This capability
is being utilized to study the stabilizing effect of rotation on
internal MHD instabilities. The ability to measure the plasma rotation
speed with spatial resolution equal to the ion gyroradius on NSTX is
permitting accurate studies of the relationship of internal structures
to rotation damping. This will help untangle possible electrostatic and
electromagnetic mechanisms that drive core momentum diffusion. Soon
DIII-D will be equipped with both co- and counter-beam injection
capability, which in cooperation with the near perpendicular injection
capability and ripple loss on JT-60U, should allow the physics of
rotation/electric field formation from momentum input and orbit loss to
be disentangled. Plasmas with no applied torque can be compared using
balanced beams in DIII-D versus RF only in Alcator C-Mod, DIII-D, and
NSTX. Momentum drag can be altered with the RWM coils in DIII-D and NSTX
and the A-coil set in Alcator C-Mod.

The physics of rotation is important and contains many unsolved physics
problems. The study of rotation has become a U.S focus area in which all
three facilities combined will make important contributions.

**4. Advanced Tokamak performance through off-axis non-inductive current
drive.**

High fusion gain plasmas with high bootstrap current fraction for
steady-state operation at and above the no-wall beta limit, require
broad, if not hollow, current profiles. High bootstrap operation tends
naturally toward hollow current profiles. Hence the use of non-inductive
current drive off-axis at large radii is crucial to achieving such
plasmas. DIII-D has pioneered off-axis current drive using electron
cyclotron waves. The magnetic field strength in DIII-D is well suited to
the use of gyrotrons for EC power. The basic physics of the off-axis
ECCD process has been verified. Initial experiments have shown the
ability to control the current profile in plasma operating above the
no-wall beta limit. Upgrades to the long pulse EC power are underway to
control the current profile on current diffusion time scales in high
beta, high bootstrap fraction plasmas. Stationary hybrid mode plasmas
for ITER have been operated for nine current diffusion times.

At the high magnetic field in Alcator C-Mod, the optimal off-axis
current drive technique is lower hybrid current drive (LHCD). Also,
because LH waves damp efficiently at relatively low electron
temperature, they are particularly well suited for driving current in
the outer plasma. The LHCD system on Alcator C-Mod has recently become
operational, and low power coupling experiments have been completed to
assess launcher performance. The LHCD experiments on C-Mod will be
performed at magnetic field strength (5 Tesla) and plasma density
(1-1.5x10^20^ m^-3^) identical to those required for current profile
control in ITER. The C-Mod facility will also provide critical data for
divertor performance at SOL densities and power flows required for AT
operation in ITER. Finally, the C-Mod experiments will be carried out
with fully relaxed current density profiles (τ~pulse~ ≥τ~L/R~) near the
no-wall beta limit (β~N~ ≅ 3), where AT operation is planned for ITER.

For over-dense plasmas, NSTX is developing Electron Bernstein Wave (EBW)
current profile control. A unique aspect of these experiments will be
use of the Ohkawa effect to drive off-axis current at low aspect ratio,
where toroidal trapping effects are significant. Radiometer measurements
and modeling have verified efficient emission of EBW at low multiple
harmonics from the over-dense plasma core, and conversion to the X and
then the O mode ECW near the upper hybrid layer usually located in the
SOL. This has established a reference approach for EBW heating and
current drive at the MW power level, which is planned to be implemented
in the next 3 years. The theoretical efficiency of this new current
drive technique is very high, and techniques of this type will be
required in future high-beta, over-dense devices.

It is very important to study all of these off-axis current drive
techniques because they are the key to the Advanced Tokamak and ST
approaches to high gain steady-state plasmas for future fusion power
systems, including advanced performance modes in ITER and a possible
Component Test Facility. ITER currently has a high power ECCD system
planned, and the option to add comparable amounts of LH power.

**5. The role of density/collisionality in edge plasma turbulence and
divertor physics.**

The three U.S facilities combined afford a unique opportunity to address
another crucial element of a successful ITER program, the ability of the
SOL and divertor plasma to exhaust impurities and the high power
produced in the ITER plasma. Central to this issue is the development of
reliable physics-based prescriptions for heat and particle transport in
the edge and divertor regions, which can extrapolate to ITER conditions.
ITER's edge and divertor will operate with a unique combination of
parameters that are not simultaneously accessible to any one device
except ITER: high absolute plasma densities and temperatures at the
separatrix, high parallel heat fluxes into the divertor region, and high
density, low temperature (partially detached plasmas) near the divertor
target plates. Therefore, the understanding of SOL/divertor physics must
be built in terms of the dimensionless parameters that control it, using
present experiments to collectively access a sufficiently wide parameter
range. Two dimensionless parameters that are thought to influence SOL
transport physics are: plasma collisionality and the transparency of the
edge to neutral penetration, (i.e., neutral mean free path divided by
machine size), that scales with poloidal field. Additional parameters
that are believed important for divertor physics include the opacity of
the divertor to Lyman-alpha radiation and neutral-neutral mean-free path
compared to divertor size.

The U.S is fortunate to have facilities that cover this entire range in
these parameters anticipated for ITER. The machine parameter with most
leverage in this regard is plasma density, which scales roughly as
plasma current divided by machine size squared. In terms of absolute
density, ITER is projected to operate from the middle of the Alcator
C-Mod density range down to the middle of the DIII-D and NSTX density
range. Alcator C-Mod's high field and compact size allows it to explore
the high density end of the spectrum, producing SOL and divertor plasmas
with high collisionality and low neutral transparencies.
Correspondingly, divertor regimes with high Lyman-alpha opacity and
fluid-like neutrals are readily accessible in C-Mod. DIII-D and NSTX
provide essential access to lower density regimes: edge collisionalities
that are comparable to ITER, a SOL more transparent to neutrals, and
divertor regimes with kinetic neutrals and low levels of Lyman-alpha
trapping.

Some important differences in edge plasma transport behavior are
detected across the US facilities, with important implications for ITER.
At medium to high densities in Alcator C-Mod with a SOL that is opaque
to neutral penetration, a high level of cross-field transport is
evident: clumps of plasma are transported deep into the scrape-off
layer; SOL profiles develop a shoulder; plasma dominantly refuels from
main chamber surfaces instead of the divertor. At the high end of its
density range, DIII-D sees a similar edge turbulence behavior. But at
the low density range, corresponding to ITER's collisionality, the
density shoulder in DIII-D is much less pronounced, the edge turbulence
is much less clump-like, and the plasma refuels from the divertor. NSTX
also sees very strong clump-like behavior in L-mode, but much quieter
edge activity in the H-mode. The range of physics behaviors that are
possibilities in ITER may be contained within the broad range of edge
parameters afforded by the three facilities.

The edge plasma turbulence has a strong in-out asymmetry. In this
regard, NSTX adds a unique capability since the in-out asymmetry of
turbulence is maximized at low aspect ratio. All three facilities have
observed the bursty, clumped plasma edge turbulence and NSTX, compared
to DIII-D and Alcator C-Mod provides a test of the effects of large
gyroradius (low field) as well as large in-out asymmetry. The range in B
encompassed by the three devices provides a unique opportunity to
understand the underlying physics of this behavior. The turbulence
behavior of the plasma edge is being studied for all three machines with
the BOUT code. Particular challenges to the physics in this code are
posed by ExB drifts (DIII-D), edge plasma flows connected to the grad-B
drift direction (Alcator C-Mod), and the large in-out asymmetry at low
aspect ratio (NSTX). The uniquely extended parameter range afforded by
these three facilities, integrated through modeling, creates a unique
opportunity for the United States to contribute to projecting edge
physics to ITER.

**6. The physics of energetic particles and Alfvén eigenmodes.**

The physics of energetic particle-wave interactions is an area where the
three facilities have complementary capabilities. Between them,
theoretical models of Alfvén eigenmode damping and drive mechanisms can
be tested under a wide range of conditions and with a wide range of
diagnostics, to provide a better understanding of energetic
particle-wave interactions to extrapolate to a burning plasma.

Gaps in the Alfvén continuum, and hence the structure, spectrum and
damping of Alfvén eigenmodes, are predicted to depend strongly on the
plasma shape, including the aspect ratio. DIII-D and C-Mod have aspect
ratios similar to ITER while NSTX, with a/R almost twice as large,
provides a test of the predicted aspect ratio dependence, and thence the
underlying physics. The tighter aspect ratio in NSTX also creates a
larger population of trapped particles, facilitating the study of
energetic particle modes driven by a resonance with the trapped particle
bounce frequency. Through similarity experiments on the three devices
with the same plasma shape, except aspect ratio, measured differences in
the drive and damping could isolate the underlying physics.

Models for the fast ion driving term can be tested with the varying fast
ion velocity distributions and ratios of fast ion velocity to Alfvén
velocity that are found in the three machines. Due to the low magnetic
field, the neutral beam ions in NSTX are super-Alfvénic, with ratios of
beam ion velocity to Alfvén velocity between 1 and 4. Strong ICRF
heating in C-Mod, provides super-Alfvénic ions with a velocity
distribution quite different from the neutral beam slowing-down spectrum
and with equilibrated thermal ion and electron temperatures. In DIII-D,
variation of the toroidal field can make the neutral beam ions sub- or
super-Alfvénic: V~f~/V~A~ can vary from about 0.3 to 2, the latter at
reduced magnetic field. The new capability of mixed co and counter beam
injection in DIII-D provides variation in the velocity distribution. RF
heating capabilities in DIII-D and NSTX also provide additional
flexibility in modifying the fast ion distribution, including
acceleration of beam-injected ions. The different heating schemes also
allow tests of the dependence of the Alfvén mode spectrum and stability
on plasma rotation and rotational shear. Beam-heated plasmas in NSTX
rotate strongly, up to a significant fraction of the Alfvén velocity,
while rf-heated plasmas in C-Mod have lower rotation, as expected in
ITER. Beam-driven rotation in DIII-D provides an intermediate case or
moderate rotation, and mixed co and counter beam injection in DIII-D
will allow variation down to near-zero rotation.

The three machines also have complementary diagnostics. C-Mod has the
key capability to study Alfvén eigenmodes with active MHD spectroscopy;
this allows detailed measurements of the frequency and damping rate of
stable modes, and thus separate measurements of the driving and damping
terms. A wide range of core fluctuation diagnostics including phase
contrast imaging (C-Mod), reflectometry (NSTX, C-Mod and DIII-D), beam
emission spectroscopy (C-Mod and DIII-D), laser interferometry and far
infrared scattering (DIII-D), and far infrared tangential interferometry
and polarimetry as well as imaging microwave reflectometry (NSTX) can be
used to investigate the newly discovered core-localized Alfvén modes.
DIII-D and NSTX's capabilities for measurement of the q-profile
(motional Stark effect) and the fast ion distribution (D-alpha
spectroscopy and scanning neutral particle analyzer, respectively) allow
study of the q-profile dependence of mode stability and the effects of
instabilities on fast ion transport. The MSE system on NSTX enables
detailed measurements of the q-profile in regimes when the beam ions are
super-Alfvénic.

The complementary characteristics of the three machines are already
being exploited: comparisons of toroidal Alfvén eigenmodes and Alfvén
cascades in all three devices have indicated similarities in the drive
mechanisms but with differences in the detailed time evolutions of the
mode structures. Such comparisons between machines, under similar and
widely varying conditions and with a broad range of diagnostics, provide
a much more powerful confirmation of theories than would be possible in
one machine alone.

**7. The dependence of transport on beta**

There is substantial theoretical and programmatic interest in the
dependence of turbulent transport on β. The standard ITER confinement
scalings imply confinement degrades like 1/β, a scaling that severely
punishes the thrust toward higher beta plasmas. On the other hand,
carefully constructed dimensionless parameter scaling experiments
between DIII-D and JET showed confinement was independent of beta.
Recent re-analysis of the ITER confinement database has increased the
uncertainty in the β-exponent of the confinement time scaling
expression, due to the relatively larger errors in the dependence on the
heating power inferred from data analysis compared to other plasma and
machine parameters. More recently, NSTX confinement time data across a
wide range of β~T~ values (10-30%) have been accumulated and analyzed.
Preliminary results suggest a positive β-exponent in the confinement
time scaling. Zero or positive β-dependence in the confinement time
scaling may allow significant performance improvement in the higher
beta-N operation scenarios estimated for the ITER hybrid and
steady-state modes.  

Nonlinear turbulence modeling has not advanced sufficiently to resolve
these questions, as the electromagnetic effects are expected to emerge
as β is increased. The combined contributions of Alcator C-Mod, DIII-D,
and NSTX can significantly expand the range of the ITER confinement
database in field (0.3-8T), beta (1-40%), and aspect ratio (1.3-3.0).
This indicates a unique capability of the combined U.S. facilities to
reduce this uncertainty in the beta-exponent, through a systematic
database analysis, that accounts for the errors in the variables. These
three machines are also well positioned to perform dimensionless
parameter scaling experiments varying magnetic field and size at the
same shape and aspect ratio (Alcator C-Mod and DIII-D), varying aspect
ratio and field at the same shape and size (DIII-D and NSTX), and
varying aspect ratio and field at the same shape but different size
(Alcator C-Mod and NSTX). The results could shed additional light on the
effects of the increasing electromagnetic effects, due to the increasing
beta, on the underlying turbulence and transport mechanisms, an
important scientific issue for all high beta confinement configurations.

Appendix 4. FESAC Facilities Panel June Meeting at the

U.S. Naval Research Laboratory (June 13-17, 2005)

All members of the FESAC Facilities Panel (see page 2 of this report)
met for five days to hear testimony from representatives of each of the
three major U.S. toroidal magnetic fusion facilities, and to deliberate
about panel findings and responses. The meeting agenda is reproduced in
A4.1 below. The list of meeting attendees is provided in A4.2.

####### A4.1. FESAC Facilities Panel Meeting Presentations

# *Monday, June 13:*

Executive Session \[08:45 -- 09:45\]

Alcator C-Mod Presentations, and Q&A \[10:15 -- 18:00\]

Introduction, *E. Marmar*

ITER/ Burning Plasma Supporting Research, *M. Greenwald*

Wave-Plasma and Advanced Tokamak, *A. Hubbard*

Transport, *M. Greenwald*

Plasma Boundary, *B. Lipschultz*

Macroscopic Stability, *E. Marmar*

Facility Capabilities, *E. Marmar*

Summary, *E. Marmar*

# *Tuesday, June 14:*

Executive Session \[08:45 -- 09:45\]

DIII-D Presentations, and Q&A \[10:15 -- 18:30\]

DIII-D Program, Staff and Facility, *R. Stambaugh*

University Research Using DIII-D, *E. Doyle*

Advanced Tokamak and ITER Research, *M. Wade*

Transport Research, *K. Burrell*

Stability Research, *E. Strait*

Alfven Mode Research, *R. Nazikian*

Pedestal and Divertor Research, *M. Fenstermacher*

Answers to the \[Charge\] Questions -- Summary, *T. Taylor*

# *Wednesday, June 15:*

Executive Session \[08:45 -- 09:45\]

NSTX Presentations, and Q&A \[10:00 -- 15:15\]

Macroscopic Plasma Physics and Solenoid-free Start-up

\+ HHFW and EBW, *J. Menard*

Multi-scale Transport Physics, *S. Kaye*

Energetic Particles, *D. Gates / N. Gorelenkov*

Plasma Boundary Interfaces, *J. Boedo*

Long-term Vision of ST as an Option for CTF, *M. Peng*

General Discussion \[15:15 -- 15:45\]

NSTX Presentations, and Q&A, continued \[15:45 -- 18:30\]

Science Overview, *E. Synakowski*

Facility Diagnostic Capabilities, *M. Ono*

# *Thursday, June 16*: Executive Session \[08:45 -- 18:00\]

# *Friday, June 17:* Executive Session \[08:45 -- 18:00\]

####### A4.2. FESAC Facilities Panel Meeting Attendees

# *U.S. Department of Energy Attendees:*

Erol Oktay, DOE (June 13-15)

Stephen Eckstrand, DOE (June 13-15)

Adam Rosenberg, DOE (June 13-15)

# *Alcator C-Mod Attendees:*

Earl Marmar, MIT (June 13-15)

Martin Greenwald, MIT (June 13-15)

Miklos Porkolab, MIT (June 13-15)

Ronald Parker, MIT (June 13)

Amanda Hubbard, MIT (June 13)

Steven Scott, PPPL (June 13)

Robert Granetz, MIT (June 13)

Paul Bonoli, MIT (June 13)

Steven Wukitch, MIT (June 13)

Bruce Lipschultz, MIT (June 13)

# *DIII-D Attendees:*

David Baldwin, GA (June 13-15)

Ronald Stambaugh, GA (June 13-15)

Tony Taylor, GA (June 13-15)

Vincent Chan, GA (June 14)

Mickey Wade, GA (June 14)

Keith Burrell, GA (June 14)

Edward Strait, GA (June 14)

Max Fenstermacher, GA (June 14)

Raffi Nazikian, PPPL (June 14)

Edward Doyle, UCLA (June 14)

# *NSTX Attendees:*

Rob Goldston, PPPL (June 13-15)

Richard Hawryluk, PPPL (June 15)

Edmund Synakowski, PPPL (June 13-15)

Martin Peng, ORNL (June 13-15)

Masayuki Ono, PPPL (June 15)

Stan Kaye, PPPL (June 15)

Jon Menard, PPPL (June 15)

Jose Boedo, UCSD (June 15)

David Gates, PPPL (June 15)

Stewart Zweben, PPPL (June 15)

Appendix 5. List of Terms and Acronyms

**Advanced Tokamak (AT) Mode:** An integrated-system MHD-stable
operating scenario for a tokamak that has high self-driven (bootstrap)
current, good particle and energy confinement, and a plasma edge that
enables particle and power handling.

**Alcator C-Mod:** A compact high-field tokamak experiment, also denoted
C-Mod.

**Alfvén wave:** A plasma wave, which involves bending or compression of
the magnetic field. The Alfvén time scale is the interval for an Alfven
wave to traverse the plasma.

**Alpha heating:** In a fusion power plant, energetic alpha particles
and neutrons are created by the fusing of deuterium and tritium nuclei.
As a charged particle, the alpha particle is unable to cross the
confining magnetic field and gives up its energy as heat to the plasma.

**Aspect ratio:** In a toroidal device, the ratio of the major radius to
the minor radius.

**Ballooning mode**: A pressure-driven MHD instability that "balloons"
out with larger amplitude in the larger major radius (lower toroidal
field) portion of a toroidal plasma, and has smaller amplitude at
smaller major radius.

**Beta:** The dimensionless ratio of plasma pressure to magnetic
pressure. Often denoted by the symbol β.

**Bootstrap current:** Currents driven by collisional transport effects
in toroidal plasmas.

**Confinement:** Property of magnetic fields in preventing loss of
energy and particles; degree or measure of this property, e.g., as in
'confinement time.'

**Detached plasma:** Cold dense plasma in the divertor chamber,
separated from the walls by neutral gas. This is characteristic of a
particular mode of divertor operation.

**DIII-D:** A medium scale, strongly shaped tokamak experiment.

**Disruption:** The abrupt termination of a tokamak plasma through the
growth of large amplitude MHD instabilities. Control of disruptions is a
critical problem for this type of confinement device.

**Divertor:** Region outside the plasma core with open field lines
leading to a chamber some distance from the plasma.

**DOE:** U.S. Department of Energy

**ECRF:** Electron Cyclotron Range of Frequencies -- refers to radio
frequency heating and current drive using waves close to the electron
cyclotron frequency.

**Edge:** Plasma region between the hot core plasma and the walls.

**ELM:** Edge Localized Mode -- an MHD mode that relaxes the pressure
gradient of a plasma device operating in H or High confinement mode.

**Equilibrium:** Usually refers to MHD equilibrium -- a steady state
solution of the MHD equations.

**ETG:** Electron Thermal Gradient (modes) -- a 'micro' plasma
instability that may be responsible for turbulent transport by very
short wavelength (electron cyclotron radius) fluctuations.

**FESAC:** Fusion Energy Sciences Advisory Committee -- Federal panel
that provides independent advice to the Director of the DOE Office of
Science on the scientific and technological issues that arise in the
planning, implementation, and management of the OFES program.

**Fusion gain (Q)**: The ratio of fusion power produced in a system to
the amount of externally applied plasma heating power.

**GA:** General Atomics -- a privately held research and development
corporation in San Diego, CA; the site of DIII-D.

**Gyroradius:** The radius at which a particle gyrates about the
magnetic field.

**H-Mode:** High (confinement) Mode -- an experimental regime with a
transport barrier or pedestal at the edge of the plasma.

**IBW:** Ion Bernstein Wave -- short wavelength electrostatic waves in
the ion cyclotron range of frequencies. Generated in the plasma by mode
conversion.

**ICRF:** Ion Cyclotron Range of Frequencies -- refers to radio
frequency heating, current and flow drive using waves close to the ion
cyclotron frequency.

**ITB:** Internal Transport Barrier -- a regime of strongly reduced
transport over some part of the core plasma.

**ITER:** A proposed international experimental fusion reactor based on
the tokamak concept.

**ITPA:** International Tokamak Physics Activity -- an activity that
fosters corporative research that aims to advance the understanding of
the physics of burning plasmas.

**JET:** Joint European Torus -- a large tokamak experiment run jointly
by a European Community consortium.

**Kink Instability:** A macroscopic (large scale) instability driven by
plasma currents.

**LH:** Lower Hybrid -- RF waves of frequency intermediate between the
electron and ion gryofrequencies and used for heating and current drive.

**L-Mode:** Low (confinement Mode) -- the baseline for confinement in
magnetic fusion devices, dominated by strong turbulent transport.

**Magnetic flux:** The integral over a surface area of the magnetic
field perpendicular to the surface. Examples are poloidal flux and
toroidal flux.

**Major radius:** Distance from the center-line of the torus to the
magnetic axis or center of the plasma cross section.

**MHD:** MagnetoHydroDynamics -- a fluid description of magnetized
plasmas.

**MIT:** Massachusetts Institute of Technology -- a major university in
Cambridge, MA; the site of C-Mod.

**Minor radius:** Distance from the center of the plasma cross section
to the plasma edge.

**Mode:** (1) a plasma wave that has, at least approximately, a constant
wavenumber and frequency or can otherwise be characterized as a normal
mode of some wave equation; (2) a mode of operation of a device, such as
AT-mode or H-mode.

**ν^\*^e:** Pronounced "nu-star-E" -- electron collisionality parameter.

**NBI:** Neutral Beam Injection -- a common method of plasma heating
which employs intense beams of neutral atoms.

**Neoclassical:** The theory of collisional transport in toroidal
geometry.

**Normalized beta:** A characterization of plasma beta (multiplied by
toroidal field and minor radius, divided by plasma current) typically
used to express the stabilization quality of a system. Initial high beta
tokamak experiments reached normalized beta values of approximately 3.

**NSTX:** National Spherical Torus Experiment -- an experimental fusion
device based on the spherical tokamak concept.

**NTM:** Neoclassical Tearing Modes -- resistive MHD modes driven
unstable by bootstrap currents.

**OFES:** Office of Fusion Energy Sciences -- the DOE Office of Science
program office that sponsors research at C-Mod, DIII-D, and NSTX.

**Ohmic:** Characteristic of dissipation due to plasma resistivity;
Ohmic heating, in which a current is driven inductively in the plasma.

**Pedestal:** Region of steep gradients at the plasma edge.

**Poloidal:** The short way around in a torus.

**PPPL:** Princeton Plasma Physics Laboratory -- a U.S. DOE fusion
energy and plasma physics research laboratory in Princeton, NJ; the site
of NSTX.

**RF:** Radio Frequency -- refers to methods of heating, flow or current
drive depending on the interaction of plasmas with externally launched
waves.

**ρ\*:** Pronounced 'rho-star', the ratio of the ion gyroradius to the
minor radius.

**RWM:** Resistive Wall Mode -- an MHD mode driven unstable by finite
wall resistivity.

**Safety factor (q):** The ratio of the number of times a magnetic field
line wraps toroidally (the long way around a toroidal device) to the
number of times it wraps poloidally (the short way around). Values less
than one typically suffer internal instability, and larger values are
typically more stable.

**Sawtooth:** a cyclical minor disruption in which core field lines
reconnect flattening the temperature profile, followed by a slow reheat
of the central plasma.

**Separatrix:** The boundary surface between regions of a plasma on
closed and open field lines.

**SOL:** Scrape-Off Layer -- region of plasma existing on open field
lines outside of a separatrix.

**Tearing modes:** Resistive MHD modes driven by plasma currents.

**Tokamak:** A magnetic confinement device relying on a large toroidal
current carried by the plasma itself. (The acronym is from the Russian
phrase Toroidalnaya Kamera I Magnitaya Katushka, meaning toroidal
chamber and magnetic coil.)

**Transport Barrier:** Region of strongly reduced transport -- typically
a bifurcated state.

**Transport flux:** The total energy or particle flow per unit area
passing through a surface element.

**Zonal flows:** Flows that are constant on flux surfaces but with
strong radial variation. These flows are self-generated by plasma
turbulence for which it is also an important non-linear saturation
mechanism.