SNOWMASS
HOT TOPIC
SUBGROUP
2: PLASMA SUPPORT TECHNOLOGY
Question
1
Topic
Leaders: D. Swain and R. Temkin
We
are working to gather a group of people to discuss and answer a particular
question in the Plasma Support Technology field, Heating, Current Drive, and
Fueling Subgroup (H/CD/F) at Snowmass. The question is:
1. What
is the potential for and what advances will be required in profile control
technologies (plasma heating, current drive and fueling) to enable present,
near term, and next step devices to meet their performance goals and ultimate
research potential?
In
addition, this subgroup is supposed to supply responses for our area to the
following questions that are common to all Plasma Support Technology Subgroups:
A. What
are the most important contributions that technology can make over the next 10
years to improve the vision for an attractive and competitive fusion product?
B. What
are the new technology issues that must be solved to allow the continued
exploration of "Advanced Tokamaks" and to enable full development of the
recently initiated or planned innovative confinement concepts and next step
devices?
C. What
constitutes engineering proof of principle and engineering proof of performance
for fusion energy systems?
This
document is intended to start this process, providing an outline for the topics
to be discussed in order to answer these questions.
Background:
Heating
and current drive technologies are essential for heating plasma to
fusion-relevant betas and temperatures and manipulating plasma properties to
access advanced operating scenarios (reversed shear, MHD stabilization,
turbulence suppression). Significant progress has been made in developing and
deploying high-power gyrotrons in the ~1-MW level at 110 GHz and the
development of 170-GHz prototype units for electron cyclotron heating/current
drive (ECH/ECCD) and fast-wave (FW) antenna arrays in the >1-MW unit size
for Ion Cyclotron Heating (ICH) and current drive (via direct electron
heating). Progress is also being made in other countries on the development of
negative-ion based, high power neutral beams (0.5–1.0 MeV). With the
present program emphasis on increasing plasma performance and reducing
next-step option costs, the emphasis of the development of these heating and
current drive technologies will concentrate on improving power density (higher
voltage limits for ICRF launchers), higher gyrotron unit power (2 to 3 MW),
increased efficiency gyrotrons featuring multistage depressed collectors, ICRF
tuning and matching systems that are tolerant to rapid load changes, and
steady-state gyrotrons and actively cooled ICRF launchers for
long-pulse/burning-plasma, next-step options.
Fueling
is another technology that is essential for the achieving fusion-relevant
plasma parameters and manipulating plasma parameters to achieve improved
performance (peaking of the density profile for higher reactivity and reducing
transport via turbulence suppression). Recent successes include sustained
operation above the density limit on DIII-D, high-field side launch with
improved density profile peaking, internal transport barrier generation, the
development of steady-state pellet injectors operating in the 1.5-km/s speed
range, and the demonstration of core fueling in proof-of-principle experiments
using accelerated compact toroids (CTs). Pellet fueling technology has also
been used recently to ameliorate the effects of major disruptions in tokamaks
by delivering massive amounts of low- and high-Z material that rapidly quench
the current in vertically unstable plasmas. It has been estimated that
eliminating disruptions in tokamaks in the fusion energy development class
would increase the lifetime of divertor plasma facing components by a factor of
two. Reducing the severity of disruptions could allow the advanced tokamak to
operate nearer its ultimate
β
potential. A critical issue for fueling in next-step device plasma regimes is
the degree to which profile peaking is needed (for higher density operation and
improved reactivity and confinement) and the technological requirements to meet
that need (pellet speed, CT density and the physics of CT deposition).
Needs
for profile control technologies:
The
ability to control the heating, fueling, and driven current profiles will lead
to the next level of plasma understanding and (it is hoped) improvement in
performance of present-day and next-step fusion devices. There are several
potential uses of H/CD/F technologies; these are described briefly in the
following paragraphs.
Plasma
profile control to improve confinement
The
ability to manipulate plasma profiles (pressure, density, current, and/or
electric field profiles)
and
sustain them in the correct state for long periods
is believed essential for achieving improved confinement and stability (e.g.,
the “advanced tokamak” modes in tokamaks). In addition, the
achievement of high (VH-mode) confinement will require the ability to develop
and sustain a transport barrier near the edge of the plasma. This is generally
believed to be possible by manipulating the electric field and/or plasma
rotation velocity profiles in the plasma.
The
profiles needed are not known, although theoretical work in this field is a hot
area of investigation. Furthermore, the
mechanisms
by which the “good confinement” regimes are set up are not well
understood. Therefore, it is important (at least at present) that H&CD
systems must be designed to allow
flexibility
in the heating deposition and driven-current profiles that they can generate.
Non-inductive
current drive
Long-pulse
or steady-state operation in tokamaks will require the full plasma current to
be driven non-inductively using some combination of current-drive techniques
and bootstrap current. The requirement (for tokamaks) that steady-state
operation be achievable is in addition to the requirement of plasma profile
control described above.
Control
of burning plasma
While
some methods of burning-plasma modes are calculated to be stable, others are
not. In this case, stabilization of the power output of the fusion power may
require feedback control of the heating power, fueling rate, or plasma
confinement. Schemes to do this have been proposed using heating and fueling.
Plasma-wall
interaction/mitigation
It
is generally recognized that for future reactor-scale tokamaks (and perhaps
stellarators), the use of “simple” divertors will cause heat loads
on the divertors and/or first walls that will be difficult to handle at best,
and possibly impossible to withstand for long periods of operation. Therefore,
some mechanism for spreading the heat load more evenly over the plasma wall is
highly desirable. The use of fueling and/or heating techniques to form (for
example) detached divertor operation or radiative plasma boundaries may be
possible.
Disruption
avoidance/mitigation
The
requirement that plasma-facing components (PFC’s) must withstand
disruptions is one of the major engineering drivers for the design of these
components in tokamaks. The ability to avoid disruptions, or at least to modify
their properties so that they will offer less damage potential to the
PFC’s, would greatly expand the ability to design a more cost-effective,
reliable fusion device.
[Are
there other uses that should be included here?-DWS]
Technologies
for consideration:
Possible
candidates for discussion are:
• Neutral
beam heating and current drive
• Electron
cyclotron heating and current drive
• Lower
hybrid heating and current drive
• Ion
cyclotron heating and current drive
• Bootstrap
current generation and profiles
• Pellet
core fueling
• Impurity
pellet injection
• Gas
impurity injection
• Compact
torus injection
• Helicity
injection
• ...
[Should
other topics be added?-DWS]
Specific
questions:
We
propose to ask (and as much as possible to answer) a common set of questions
for each of the technologies listed in the previous section:
• What
are the
needs
of present-day and future devices (heating, current-drive, and fueling profiles)?
• What
is the
presently
demonstrated
technological
status
of each technique [power/unit, pulse length, frequency (if applicable), power
flux, reliability, rep. rate, fueling ability...]?
• What
is the
presently
demonstrated
scientific
status of each technique [heating and current-drive ability (both on- and
off-axis), fueling profile, effect on plasma properties (e.g., barrier
formation)]?
• What
is the
present
ability
of the technologies that now exist to fulfill these needs?
• What
development
is needed, and how difficult does the development appear to be?
[Are
these the right questions? Should other questions be added? Should these be
changed?-DWS]
Core
Group:
A
core group is being formed to address these questions before and during the
Snowmass meeting. We plan to have a draft of the report on this topic completed
before the meeting. Members of the core group are [
these
are tentative names; need more, especially for fueling, NB, LH and needs
]:
Person Institution Speciality
D.
Swain
ORNL ICH
R.
Temkin
MIT ECH
R.
Callis
GA ECH/ICH
H.
Neilson
PPPL Future
machine needs
R.
Majeski?
PPPL ICH
R.
Pinsker
GA ICH/(ECH?)
P.
Bonoli
MIT LH
R. Wilson
PPPL ICH
L.
Baylor
ORNL Pellet
fueling
L.
Grisham
LBL Neutral
beams
M.
Schaffer
GA Helicity
injection
D.
Hwang
UC
Davis
Compact
toroid injection
T.
Jernigan
ORNL Disruption
Mitigation
Report
outline:
A
tentative
outline of the report is shown below. Each of the technology sections should
contain subsections addressing each of the questions described above. Names by
respective topics are
extremely
tentative and preliminary!
Outline
Introduction
and Executive Summary
Needs
in the H/CD/F areas to improve plasma performance
Present-day
Future Neilson
(PPPL)
Heating
and Current Drive
Neutral
Beam Technology
Grisham
Electron
Cyclotron Technology
Temkin
Lower
Hybrid Technology
Bonoli
Ion
Cyclotron Technology
Swain
Helicity
injection
Schaffer
Fueling
Pellet
injection
Baylor
Compact
torus injection
Hwang
Gas
and impurity injection
G.
Jackson?
Disruption
mitigation
Jernigan