Advanced Tokamak Modes in ITER and FIRE

The need for advanced tokamak (AT) modes capable of steady-state and
high-power-density operation in a magnetic fusion power plant was
recognized in the late 1970s, and resulted in the formation of the US
Advanced Tokamak Program in 1979. A series of Advanced Reactor
Innovation Evaluation Studies (ARIES) in the 1990s quantified the impact
of steady-state and high-power-density on the economic attractiveness of
a tokamak based power plant. The discovery of the reversed shear
(negative central shear) AT mode by Kessel (US) and Ozeki (JA) in 1992
opened the way for self-consistent steady-state high-power-density
advanced tokamak operating modes. Systematic investigation of reversed
shear (negative central shear) AT modes began in 1994 on TFTR and
DIII-D, and a significant effort is now being carried out by the
international tokamak research program.

An attractive steady-state power plant will require a large fraction
(70 - 90%) of the plasma current to be driven by the bootstrap effect
with the remainder driven by neutral beams or RF waves. Good progress
was made in the 1980s and 90s in developing the physics basis and
technology needed to drive plasma current using neutral beams, and
various RF frequency waves - fast wave ion cyclotron, lower hybrid and
electron cyclotron. In addition, self-driven (bootstrap) currents up to
80% have been produced in a number of tokamaks for short pulses. The
ARIES power plant studies had fusion power densities of \~5 MWm^-3^ ,
which will require B^2^ sufficient to produce volume average plasma
pressures, p, of 10 atmospheres (atm). Progress toward achieving the
plasma pressure needed for a MFE power plant is shown in Fig. 1.

![](media/image1.png){width="5.697222222222222in" height="3.1in"}

Plasma pressures of just over 1 atm have been achieved in several
tokamaks including; Alcator C, C-Mod, TFTR, JT-60U, JET and DIII-D. The
maximum pressure of 1.6 atm was achieved by Alcator C in 1983 using only
ohmic heating. Strongly shaped plasmas like PBX-M and NSTX have achieved
plasma pressures of 0.3 atm and 0.25 atm respectively. The maximum
plasma pressure is limited by a combination of plasma physics, coil
geometry and engineering as shown by the three terms in Eq. 1 below.

p = ~to~ B~to~^2^ = ~to~ (B~to~/B~coil~)^2^ B~coil~^2^ (1)

> = ~fusion~ B~coil~^2^ (2)

For superconducting coils, B~coil~ is subject to the limits of
electromagnetic stress and the superconducting transition field while
normal conducting coils are subject to the limits of electromagnetic and
thermal stress as well as coil power dissipation. The operational limits
for various coil systems are also shown in Fig 1. Therefore, the
appropriate "Figure of Merit" for utilization of magnetic field in a
magnetic fusion reactor is ~fusion~ = p/ B~coil~^2^ where B~coil~ is the
maximum field achievable at the toroidal field coil as shown in Eq. 2
above. As shown in Fig 1, the achieved ~fusion~ ranges from 0.2% to 1%.
A variety of power plant design studies and burning plasma experiments
proposals are also shown on Fig. 1 with power plant designs in the range
of ~fusion~= 1 -- 2% that would result in plasma pressures of 10 atm and
DT power densities of 5 MW^-3^.

The ongoing tokamak program and a next step burning plasma experiment
have the goals to understand the physics and to determine the
requirements for attaining, controlling and sustaining high-
steady-state advanced tokamak regimes for time scales long compared to
internal plasma time scales. Activities are underway on ITER and FIRE to
develop the experimental scenarios and determine the hardware
requirements to address advanced tokamak regimes in strongly burning
plasmas. The present ITER regimes are focused on the physics and plasma
technology of moderate power density plasmas sustained for very long
pulse (\~10 ~CR~) while the FIRE regimes are focused on high power
densities sustained for moderate pulse lengths (3 -- 5 ~CR~). The
physics and plasma technology issues of ITER and FIRE are very similar,
and technical solutions for one will likely be applicable to the other.
The major common issues are: (1) refinement of predictive capability and
optimization of confinement modes, (2) improved understanding of edge
plasma behavior leading to reduction of edge plasma power loss during
ELMs and disruptions, (3) extension of advanced tokamak scenarios toward
higher and bootstrap current fraction, (4) analysis of instabilities
driven by energetic particles in fusion plasmas, (5) development plasma
facing components to handle high power densities while maintaining a low
tritium inventory, (6) development of practical plasma control
techniques (profile control and feedback systems) and (7) development of
diagnostics suitable for burning plasma physics and plasma control.

Both ITER and FIRE are being designed to address these issues by
exploring and understanding burning plasma physics in the conventional
H-mode regime, and in the advanced tokamak (~N~ \~ 3 - 4, f~bs~ \~50 -
80%) regime envisioned for an attractive steady-state high-power-density
fusion power plant. The goal of ongoing work is to develop AT modes that
would fully exploit the capability of ITER and FIRE. ITER has employed
conservative scenarios, as appropriate for their nuclear technology
mission, while FIRE has employed more aggressive assumptions aimed at
exploring the scenarios envisioned in the ARIES power plant studies. The
general physics parameters of the AT modes in ITER and FIRE are shown in
Table 1 in comparison to existing experiments and the ARIES-RS/AT power
plant plasma parameters. Both ITER and FIRE use the same physics basis
and would explore 100% non-inductive plasma drive scenarios that would
be capable of steady state operation. The major difference in the plasma
configuration is that FIRE has a double null (DN) pumped divertor while
ITER has a single null (SN) pumped divertor. The DN configuration allows
stronger plasma shaping (triangularity) which leads to higher
confinement and smaller ELMs in present experiments. The SN divertor has
fewer modules, which is expected to lead to reduced costs. ITER employs
negative ion neutral beams for on axis plasma current drive and lower
hybrid current drive for off axis current drive while FIRE would use ion
cyclotron fast waves on axis and lower hybrid off axis. A key physics
requirement is to develop high modes that are stable to resistive wall
modes (RWM) under power plant-like conditions with very low plasma
rotation as would be present in an ARIES-plasma. The planned experiments
on Alcator C-Mod will provide important data on the development of AT
modes using RF waves in a plasma with very low rotation.

![](media/image2.png){width="3.8993055555555554in"
height="2.8041666666666667in"}

ITER and FIRE advanced tokamak operating modes will require significant
advances in plasma technology. Plasma technologies are particularly
important due to the close coupling between burning plasma performance,
plasma profile control and the plasma wall interaction. Parameters
relevant to plasma chamber technologies are summarized in Table
2![](media/image3.png){width="3.8993055555555554in"
height="3.3520833333333333in"}

AT modes place a severe challenge on the divertor and first wall in a
tokamak producing high fusion power while operating at relatively low
plasma densities. A careful optimization of the plasma loss power
radiated to the first wall relative to the power lost along the plasma
scrape-off and then partially radiated in the divertor chamber is needed
to achieve high power performance. The plasma exhaust power densities in
the divertor of ITER and FIRE are comparable, and both divertor chambers
are designed for steady-state cooling. The ITER divertor targets are
carbon with the ability to withstand high pulsed heat loads from ELMs or
disruptions but with the disadvantage of high tritium retention due to
co-deposition. FIRE's divertor targets are a tungsten rod brush design
based on R&D in the US. Samples have been tested to 20 MWm^-2^ for over
10,000 cycles. Tungsten has the advantage of low tritium inventory but
special operational precautions will be needed to reduce the incidence
of Type I elms and effects of disruptions. Additional R&D both on
existing tokamaks and on stand-alone technology facilities is needed to
develop materials and configurations that will satisfy the requirements
for ITER or FIRE.

The high radiated power density and significant nuclear heating of the
first wall in a burning plasma experiment provide a challenge and will
provide a benefit to the design of a DEMO. During normal operation the
radiated power absorbed by the first wall in ITER and FIRE will be in
the range of 0.5 to 1.5 MWm^-2^ depending on the optimization of the
sharing of the power loss between the first wall and the divertor. The
ITER first wall is designed for steady-state cooling while the FIRE
first wall is cooled between pulses.

Launchers for RF heating and current drive face a hostile environment of
plasma thermal loads, neutron irradiation and disruptions. Much more R&D
is needed in this area. The Joint US-EU project on the design and
fabrication of a JET high power ICRF launcher that is similar to the
ITER launcher design is a good example of R&D supporting the
construction of ITER.

Another major area of research is the development of resistive wall mode
(RWM) feedback stabilization systems needed for achieve the full
potential of AT modes in ITER/FIRE burning plasma experiment and
economically attractive performance in an ARIES-like power plant. FIRE
proposes to integrate the RWM coils into the first wall portion of the
port plug assemblies in 8 of the 16 large mid-plane ports. This
configuration allows close coupling of the feedback coils to the plasma
and calculations benchmarked on existing experiments indicate that this
will allow FIRE to achieve ~N~ \~ 4 which is required to produce fusion
power densities of 5 MWm^-3^. The ITER design has RWM coils located
outside the toroidal field coils, which only allows weak coupling of the
feedback coils to the plasma. As part of the International Tokamak
Physics Activity (ITPA), US participants have shown that moving the RWM
coils inside the vacuum vessel in ITER would allow a significant
increase in . A major question is whether RWM coils can be developed
that are compatible with the neutron irradiation environment of a
burning plasma experiment, and eventually an ARIES-like power plant. The
design of a generic port plug assembly that integrates a first wall
capable of high wall loading, RWM coils and diagnostics would be of
significant benefit to the AT capability of ITER and FIRE.

In summary, both ITER or FIRE would benefit from AT mode operation and
each would make important contributions to physics understanding of AT
modes and to the technologies needed to create and control a burning
plasma in an advanced tokamak. A US Burning Plasma Program should be
initiated in the near term to address a number of high leverage physics
R&D items for the conventional mode and the advanced mode for FIRE and
ITER. There needs to be an increased emphasis on physics R&D for
advanced modes, high power density plasma facing components including RF
launchers, burning plasma diagnostics.

Work supported by DOE Contract \# DE-AC02-76CHO3073.