**Interim Physics Report**

**Fusion Ignition Research Experiment (FIRE)**

**An Option for a Major Next Step in Magnetic Fusion Research**

Dale Meade (July 2, 1999)

**Challenges for Major Next Steps in Magnetic Fusion**

TFTR and JET have successfully carried out the initial burning plasma
experiments with D-T fuel where alpha particle effects were
perturbations on the plasma dynamics. These experiments were carried out
as part of the Performance Extension Phase. A crucial step needed to
provide the science foundation for the Energy Development Phase of
magnetic fusion is the exploration and understanding of reactor plasma
dynamics where alpha particles dominate the plasma dynamics. The tokamak
is the only magnetic configuration capable of accessing the
alpha-dominated burning plasma regime within the next decade to provide
generic information on magnetically confined burning plasmas.

![](media/image1.pct){width="5.288194444444445in" height="3.6875in"}

Fig. 1. The Lawson diagram showing the present status of laboratory
experiments in magnetic and inertial fusion.

The Fusion Policy Advisory Committee (FPAC) Report (1990) recommended a
strategy that included CIT, a steady-state confinement experiment and a
materials test facility that were followed by ITER. The 1995 PCAST
review of Magnetic Fusion recommended a similar modular strategy with
programs and facilities specialized to address the ignition,
steady-state and technology issues. The Grunder FESAC Panel (January
1998) recommended that "A burning plasma facility should be built at the
earliest possible time" and that the U.S. should examine "lower cost
reduced scope options in the interest of achieving a fusion energy
producing plasma experiment on the fastest possible schedule." The
Modular Pathway was preferred by the fusion community attendees at a
Forum for Major Next Step Fusion Experiments held in Madison, Wisconsin
(April 1998). At that time, the Next Step Experiments were described in
terms of hardware; a short-pulse copper-coil burning-plasma tokamak, a
long-pulse superconducting-coil advanced tokamak and supporting
technology facilities. Recently, more emphasis has been given to
determining the functional requirements for the plasma that would be
needed to address the following fusion plasma issues:

> Burning Plasma Physics - The achievement and understanding of
> self-heated plasmas with high gain that have characteristics similar
> to those expected in a fusion energy source, and
>
> Advanced Toroidal Physics - The achievement and understanding of
> sustained self-heated plasmas with characteristics (steady-state or
> high duty factor pulsed systems) similar to those expected in a
> competitive fusion system

The plasma performance and duration to address these issues are shown
schematically in Fig.2 in terms of the natural time scale for the
important plasma processes.

![](media/image2.pct){width="5.884722222222222in"
height="3.9895833333333335in"}

This report documents the results of a study to evaluate the capability
of compact high field tokamaks to address the alpha-dominated burning
plasma physics, long-pulse advanced-toroidal physics and fusion
technology as part of a Modular Pathway to Magnetic Fusion Energy (MFE).
The conclusion is that a compact high field tokamak utilizing LN cooled
copper-alloy coils has the capability to address a major portion of both
the Burning Plasma Experiment Step, the Advanced Toroidal Experiment
Step and also has significant capability to integrate burning plasma
physics with advanced toroidal physics. The device studied resembles
CIT. The plasma configuration was drawn from TPX and is a ≈ 1/3 scale
model of ARIES-RS. The size was constrained with the goal to achieve the
most important physics goals at a construction cost of \<\$1B. Much
remains to be done in both physics and engineering, and the suggestions
from the fusion community are both needed and welcome.

**The Next Frontier in MFE Research - Exploration, optimization and
understanding of alpha-dominated burning plasmas.**

The attainment and control of a high Q plasma dominated by alpha heating
is the single most important requirement for any approach to fusion
power. Fusion gains Q \~ 20 are needed for an economical magnetic fusion
reactor that is sustained at near steady-state conditions; at this Q
value the alpha particles dominate the plasma dynamics, providing 80% of
the plasma heating. The goal for the Next Step in Magnetic Fusion is to
access sustained alpha dominated plasmas with alpha heating fractions
more than an order of magnitude higher than present experiments.

Present day tokamaks improve plasma performance by controlling density,
pressure, plasma current and transport profiles. The advanced tokamak,
advanced stellarator and the spherical torus also plan to have the
bootstrap current, generated by gradients in the pressure profile,
produce a large fraction of the current needed to define the stabilizing
magnetic field. Since the alpha heating profile is directly linked to
the pressure profile, this process becomes very non-linear in
alpha-dominated plasmas required for a fusion reactor based on an
advanced tokamak. This coupling of advanced tokamak confinement and MHD
stability physics with alpha-dominated plasmas is a key generic issue
for the development of attractive toroidal magnetic reactors.

**Fusion Ignition Research Experiment (FIRE) - A Next Step Option for
MFE.**

The mission of FIRE is to attain, explore, understand and optimize
alpha-dominated plasmas that will provide the knowledge for attractive
MFE systems. The guiding design philosophy is that FIRE must have the
capability and flexibility of studying and resolving the physics issues
relevant to the design of a subsequent advanced integrated fusion
facility. A major consideration is to accomplish this physics mission at
the lowest possible cost, with a target cost \<\$1B. This report
summarizes the first nine months of a study to evaluate the physics and
engineering capabilities of a compact high field tokamak utilizing
cryogenically-cooled copper-alloy coils to enable this mission.

**FIRE Physics Objectives**

The physics objectives of FIRE developed to satisfy the mission are to:

1\. Determine and understand the conditions required to achieve
alpha-dominated plasmas:

• Energy confinement scaling with dominant-alpha heating

• β-limits with dominant-alpha heating

• Density limit scaling with dominant-alpha heating

2\. Explore the dynamics of alpha-dominated plasmas using active control
techniques.

3\. Sustain alpha-dominated plasmas with high-power-density exhaust of
plasma particles and energy and alpha ash exhaust in regimes suitable
for future toroidal reactors.

4\. Explore and understand alpha-dominated plasmas in advanced operating
modes and configurations that have the potential to lead to attractive
fusion applications.

5.  Understand the effects of fast alpha particles on plasma behavior in
    relevant regimes.

**Alpha Heating Fraction, the metric for alpha-dominated burning
plasmas**

The alpha heating strength can be expressed in terms of fα = Pα/ (Pα +
Pext) where Pα is the alpha heating and Pext is the externally applied
heating. The fraction of alpha heating, fα, is plotted in Fig. 3 in
terms of the ratio nτE/ nτE(Q = ∞). The vision of an MFE fusion reactor
is ARIES-RS, an advanced tokamak with a Q = 25 which has fα = 0.83.
Small reductions in confinement produce only small changes in the alpha
heating fraction, while Q can change by a large amount especially in the
high Q regime.

![](media/image3.pct){width="6.497916666666667in"
height="4.575694444444444in"}

Fig. 3. Fraction of Alpha heating versus nτE/ nτE(Q = ∞) illustrating
the alpha-dominated regime.

D-T experiments on TFTR and JET have measured small temperature
increases in agreement with the expected alpha particle heating. The
sustained D-T discharges on TFTR and JET had Q ≈ 0.2 for \~ 10 energy
confinement times with the alpha particles providing about 4% of the
overall plasma heating. The investigation of an alpha dominated plasma
can begin at fα = 0.5, and plasmas with fα = 0.66 to 0.83 would match
the dynamics expected in the MFE reactor regime.

**Choice of FIRE Plasma Performance Requirements**.

The goal of FIRE is to investigate alpha-dominated burning plasmas
relevant to future advanced tokamaks using a minimum size device with a
cost \<\$1B. FIRE is a physics experiment to extend the frontiers of
fusion plasma physics into previously unexplored parameter space using
advanced capabilities and flexibility for later upgrades; it is not a
demonstration of the scientific and technological feasibility of
magnetic fusion. The strategy for the FIRE program is to have a first
stage of burning plasma experiments aimed at accessing the
alpha-dominated regime with a minimum fα of ≥ 0.5 using projections from
the middle of the present tokamak performance database. This would
provide a test bed where alpha heating effects are easily observable,
and the plasma dynamics could still be controlled externally. This
capability is the natural starting point for an experimental campaign to
study alpha-dominated plasmas and would be sufficient to accomplish a
significant fraction of the stated objectives. The goal for the second
stage of burning plasma experiments is to achieve strongly
alpha-dominated plasmas with fα = 0.66 to 0.83. This level of
performance is projected from the best results of the present tokamak
performance database, or by a modest 20% improvement in confinement from
employing advanced tokamak physics that is expected to be developed by
the ongoing base tokamak program over the next 5 years.

The pulse length, or the burn time, is a very important consideration
for any burning plasma experiment. The physics time scales of interest
(with typical values for FIRE plasmas) are:

• ταs, the time needed for the alpha particle to transfer its energy to
the plasma (\~ 0.1 s)

• τE, the plasma energy confinement time (\~ 0.6s)

• τHe, the confinement time of alpha ash, slowed down alpha particles
(\~ 5 τE \~ 3s)

• τcr, the time for the plasma current profile to redistribute after a
perturbation (\~13 s)

The characteristic time scales for plasma phenomena in FIRE plasmas are
significantly shorter than the corresponding time scales on ITER-RC due
to the smaller size, higher density and somewhat lower plasma
temperature as shown in Table I.

Table I. Characteristic time scales for plasma phenomena in FIRE and
ITER-RC.

  -------------- -------------- -------------- -------------- --------------
  For Q ≈ 10     τE (s)         τHe (s)        τcr (s)        τburn (s)

  FIRE           0.6            3              \~13           10 - 20

  ITER-RC        2.5            7.5            \~200          400
  -------------- -------------- -------------- -------------- --------------

A FIRE plasma with a burn time of 10 s ( \~ 15 τE) would allow the
pressure profile to come into equilibrium with alpha heating and allow
the alpha ash to accumulate for \~ 3 τHe . This pulse length would be
sufficient to address Physics Objectives 1, 2, 3, and 5. A significant
part of Physics Objective 4 could be accomplished using a current
profile that is only partially redistributed. In fact, it would be
advantageous to establish a variety of plasma current profiles using
current ramping as in present advanced tokamak experiments. A pulse
length of \~30 s would be sufficient to allow the bootstrap driven
current in an advanced tokamak mode to come into equilibrium. These
pulse length requirements match the capabilities of liquid nitrogen (LN)
cooled copper coils, which can be designed to allow a burn time of 10 to
20s at full toroidal field. If advanced tokamak physics improves
confinement relative to ITER design guidelines by 25% and β by 50%, then
the toroidal field and plasma current can be reduced by 25% while
maintaining high plasma performance (e.g., Q \~ 10). This small
reduction in the field of the FIRE copper magnet cooled to LN
temperatures would allow the burn time to be increased to 30 to 40s).

**FIRE Device Parameters for Initial Evaluation**

The FIRE plasma configuration is an extension of the advanced tokamak
programs on DIII-D and Alcator C-Mod, and is a ≈ 1/3 scale model of
ARIES-RS, the present vision for an advanced tokamak fusion reactor. The
FIRE plasma has a size and shape very similar to the previously proposed
advanced tokamak (TPX), with the added capability of high performance
D-T operation. The capability of FIRE to carry out long-pulse
non-burning plasmas experiments will be described in a later section.
FIRE will have the flexibility to incorporate new innovations as the
ongoing advanced tokamak program develops them. The parameters
summarized in Fig.4 were chosen as likely to achieve the FIRE mission at
the lowest cost based on results of prior design studies for burning
plasmas experiments (CIT, BPX and BPX-AT), as well as recent information
from the ITER-EDA and ITER-RC design activities. A more extensive list
of parameters and features is given in Appendix 1.

![](media/image4.pct){width="4.690277777777778in"
height="3.2416666666666667in"}

Fig. 4. Cross-section view and design goals of the FIRE.

**Capability for Alpha-Dominated Burning Plasma Experiments on FIRE**

The plasma performance of FIRE is estimated using the guidelines similar
to those used to project the performance of ITER. The primary
considerations are the maximum density limit, plasma energy confinement,
the maximum pressure (β) limit, the power threshold for accessing the
high confinement mode (Elmy H-mode) and limitations imposed by
impurities due either to alpha ash accumulation or impurities from the
first wall and divertor. The guidelines for estimating plasma
performance are described in more detail in Appendix 2. FIRE assumes an
operating density relative to the Greenwald density closer to those in
the ITER confinement data base rather than the higher values assumed in
the ITER performance projections. FIRE assumes a slightly more peaked
density profile (identical to that used in the CIT and BPX projections)
than ITER due to the potential for tritium pellet injection into a much
smaller high-density modest temperature plasma. FIRE also takes credit
for lower impurity fractions characteristic of high-density tokamak
plasmas. In particular, FIRE assumes no significant high-Z impurities in
the plasma core from the divertor.

It is important to note that while these guidelines are quite useful for
estimating the nτET performance of **existing** tokamaks to within 30%,
the guidelines are mainly empirical with a modest amount of theoretical
understanding and can not accurately predict the performance of a Next
Step Burning Plasma experiment much less a technology demonstration. An
affordable flexible experiment with a performance capability about
midway between today\'s tokamaks and a fusion reactor is needed to
benchmark physics understanding and to serve as a stepping stone to a
reactor.

The strategy of FIRE is to minimize the extrapolation in τE, the most
uncertain quantity. The fusion gain is maximized by maximizing nτE at a
plasma temperature of \~10 keV. Analysis of the power balance in the
plasma, first done by J.D. Lawson, shows that nτE values of \~ 4 x
10^20^ m^-3^ s are required to achieve Q values \~ 10 for a D-T plasma
with modest impurity contamination and typical profiles. The compact
high field tokamaks (IGNITOR and FIRE) reduce the requirement on τE by
operating at densities almost an order of magnitude higher than larger
lower field devices such as ITER. Operating in the high-density regime
ne(0) = 6.75 x 1020 m-3 is a straightforward matter since Alcator C-Mod
has already operated up to \~ 10^21^m^-3^. For Q = 10, FIRE requires an
energy confinement time, τE , of only \~ 0.6 s, which has been achieved
in existing tokamaks such as JET, rather than the \~ 2.5 s required in
the reduced size ITER or the 6 s required for ignition in ITER. The
dimensionless confinement time, BτE, is useful to quantify the
extrapolation required in plasma energy confinement from present
experiments to potential Next Step Options for burning plasmas.

Table II. Extrapolation of dimensionless energy confinement time for
potential Next Step Options.

  --------- -------- ---------------- ------------------ -------------------
            JET      FIRE (Q = 10)    ITER-RC (Q = 10)   ITER-EDA (Q = ∞)

  BτE (T-s) 3        6                14                 34
  --------- -------- ---------------- ------------------ -------------------

The extrapolation to Q \~ 10 conditions in FIRE is a factor of two
beyond JET, while ITER-RC requires a factor of four extrapolation, and
ITER-EDA required an extrapolation of \~11 to achieve the objective of
ignition. Therefore, the uncertainty in projecting confinement would be
less in FIRE than for ITER-RC.

The plasma parameters for a nominal FIRE operating point calculated
using a zero dimensional model and the physics guidelines (Appendix 2)
are given in detail in Appendix 3. The alpha heating fractions for FIRE
and ITER-RC are illustrated in Fig.5 under the assumptions of modestly
peaked density profiles (triangles) and flat density profiles (crosses).
The initial design point selected for FIRE satisfies all of the standard
tokamak design guidelines (Appendix 2) needed to access the alpha
dominated range with Pα / Pheat ≥ 0.5 (Q ≥ 5). This represents more than
an order of magnitude advance beyond the capability of TFTR/JET to study
alpha driven physics, and would provide a checkpoint more than half way
to the alpha heating fraction Pα/Pheat ≥ 0.8 required in a fusion
reactor.

![](media/image1.pct){width="6.285416666666666in"
height="4.029166666666667in"}

Fig. 5. Performance of FIRE and ITER-RC versus H-mode multiplier. HH =
1.0 is the center of present tokamak H-mode data base (ITER DB3). The
triangles are for slightly peaked density profiles, α~n~ = 0.5. The MFE
reactor points are for ARIES-RS at densities ranging from 1.0 to 1.8
times the Greenwald density.

The operating range for an H-mode confinement multiplier of H98 = 1.05
given in Fig. 6 shows an extensive alpha dominated range (Pα / Pheat ≥
0.5) with Pα / Pheat ≥ 0.66 being accessed but only in a small range.

![](media/image5.pct){width="6.5in" height="3.782638888888889in"}

Fig. 6. Alpha dominated (Q \> 5) operating range for FIRE with an
ITER-98 H-mode factor = 1.05, β~N~ ≤ 2.5, P~heat~ ≥ 1.0 P~th~ to access
H-mode and P~heat~ ≥ 0.5 P~th~ to remain in the H-mode while satisfying
a density limit of n/n~GW~ ≤ 0.75.

The technical basis for a compact high-field tokamak like FIRE has
improved markedly since the CIT (R = 2.14 m) and BPX (R = 2.59 m)
studies of 1989-91. Tokamak experiments (1989 -1999) have led to the
development of a new scaling relation (e.g., ITER-98H) which predicts
1.3 times higher confinement than the 1989 CIT design assumption.
Alcator C-Mod, which can be considered as a prototype of FIRE, has come
into operation and demonstrated:

> • Confinement of 1.4 times the 1989 CIT design assumptions, \~ 15%
> higher than the ITER - 98H scaling.
>
> • High power density ICRF heating of high density shaped plasmas with
> a divertor.
>
> • Detached divertor operation at high power density.

In addition, D-T experiments on TFTR and JET have shown that tritium can
be handled safely in a laboratory fusion experiment. The D-T plasmas
behaved roughly as expected with slight improvements in confinement for
the very weak alpha heating conditions available. The behavior of the
energetic alpha particles was in agreement with theoretical
expectations.

The performance projections (Fig. 7) indicate that FIRE is also capable
of exploring strongly alpha-dominated regimes with Pα/Pheat ≥ 0.66 (Q =
10 to 30) if the relatively higher performance (H98 = 1.2) of the
smaller compact high field tokamak, Alcator C-Mod, or the top end of the
JET confinement results are obtained at burning plasma conditions in
FIRE. During the next ten years the ongoing world wide advanced tokamak
program is expected to provide additional improvements of at least \~25%
in confinement and 50% in β. This capability would allow FIRE to explore
"ignited" plasma conditions with Pα/Pheat ≥ 0.8 (Q up to 30) at reduced
fields and longer pulses comparable to the plasma current redistribution
time.

![](media/image6.pct){width="6.5in" height="3.720138888888889in"}

Fig. 7. High Gain (Q \>10) FIRE operating range with an ITER-98 H-mode
factor = 1.2, β~N~ ≤ 2.5, P~heat~ ≥ 1.0 P~th~ to access H-mode and
P~heat~ ≥ 0.5 P~th~ to remain in the H-mode while satisfying a density
limit of n/n~GW~ ≤ 0.75.

Success with FIRE would establish the basis for an advanced toroidal MFE
reactor, namely a smaller and cheaper system obtained by introducing
modest advanced features to make something that is good even more
attractive as a fusion reactor concept.

**Capability for Long Pulse Advanced Tokamak Experiments in FIRE**

The development, exploration and detailed understanding of
high-confinement, high-fusion power-density and steady-state
(high-duty-cycle) plasmas are needed to provide the basis for
economically attractive applications of fusion power. FIRE has
significant capability to address the "steady-state" advanced toroidal
configuration initiative of the Modular Strategy using non-burning
deuterium plasmas.

The issues to be addressed include:

1.  Optimization of confinement and β using current and pressure profile
    control

2.  Integration of high performance plasma with long pulse particle and
    power exhaust

3.  Development of controls and techniques to avoid or mitigate the
    effects of disruptions

The objective of FIRE would be to extend these studies beyond the plasma
performance and duration accessible in present tokamaks to values closer
to those for a fusion plasma. The physics of a fusion plasma is
characterized by the dimensionless parameters, ρ\* (normalized
gyro-radius), ν\* (normalized collision frequency) and β. Existing
tokamaks are able to replicate the ν\* and β for a fusion reactor plasma
but not simultaneously ρ\*, which requires a plasma with a larger Ba.
The important time scale for exploring advanced toroidal physics is
τ~cr,~ the time for the plasma current profile to redistribute after a
disturbance. The current redistribution time is often called the plasma
current skin time. A plasma duration of \~ 3 τ~cr~ would be sufficient
to allow the current profile to

relax to within 5% of the steady-state current profile, and would be
sufficiently long to address long pulse advanced tokamak physics.
![](media/image7.pct){width="5.888888888888889in"
height="4.207638888888889in"}

The flattop of the magnetic field in FIRE increases rapidly as the
magnetic field is reduced as shown in Fig. 8. Recall that FIRE has about
the same size and shape as the previously proposed long-pulse advanced
tokamak experiment, TPX. In FIRE, the skin time is typically \~13 s at Q
\~ 10 parameters due to the small minor radius, so pulses longer than
\~30s would allow the plasma current profile to approach its equilibrium
value. FIRE operated at 8T would be able to extend the range of
long-pulse advanced tokamak physics studies a factor of two in ρ\*
beyond the capability of any existing shaped divertor tokamak or any
under construction for "long-pulse" durations ≥ 3 τ~cr~. The addition of
lower hybrid in a later upgrade phase would be suitable for current
drive to sustain the long pulse plasmas in FIRE. This capability would
be within a factor of two of ρ\* in ARIES-RS.

Table III. Comparison of Long Pulse Advanced Tokamak Experiments. (β/ν\*
= constant)

  ----------- -------- ------ --------- ------- ------- ---------- -----------
              DIII-D   JET    KSTAR     TPX     FIRE    ITER-RC    ARIES-RS

  B(T)        2.2      2      3.5       4       8       5.5        8

  Nρ† = 1/ρ\* 0.25     0.27   0.25      0.3     0.45    1.05       1.0

  Duration/   1.7      4      7-100     \>100   4       2.2        ∞
  τ~cr~                                                            
  ----------- -------- ------ --------- ------- ------- ---------- -----------

† normalized to 1/ρ\* for ARIES-RS

**Integration of Long-Pulse Advanced Tokamak Physics and Alpha Dominated
Burning Plasmas, the critical issue for Advanced Toroidal
Configurations, can be explored on FIRE.**

It is anticipated that over the next ten years the ongoing world-wide
advanced tokamak program (total funding \~\$1.2B) will develop the
advanced tokamak tools and techniques to improve confinement (e.g., H98)
by 20% and β~N~ by at least 50%. Most importantly, an experiment, such
as FIRE, could then exploit these techniques to extend the range of
burning plasma experiments and begin to address the critical issue of
integrating of advanced tokamak physics with alpha dominated plasmas.

**Fusion Technology and Materials Development and Experience from FIRE**

FIRE would also make significant contributions to the Fusion Technology
and Materials Initiative of the Modular Strategy. FIRE will produce
reactor-like fusion power density and neutron wall loading, but neither
FIRE nor ITER-RC would produce the neutron fluence needed to provide
data on neutron damage to structural materials (Table IV). Nonetheless,
some information could be obtained from tests involving blanket modules.

Table IV. Fusion neutron parameters for Next Step Options compared to a
reactor.

  ------------------------------- -------------- ------------ ------------
                                  FIRE           ITER-RC      ARIES-RS

  Fusion Power Density (MWm^-3^)  10             0.5          6

  Neutron Wall Loading (MWm^-2^)  3              0.6          4

  Neutron Fluence (MW y m^-2^),   \< 0.01        \~1          120
  lifetime                                                    
  ------------------------------- -------------- ------------ ------------

The lifetime neutron production in FIRE will be limited to 5 TJ (3 x
10^-3^ MW-y m^-2^) to reduce damage to insulators in the toroidal field
coil. Consideration will also be given to the use of low activation
materials in various components to gain experience for the follow-on
Advanced Fusion Integration Facility in the Modular Strategy. Remote
handling will be implemented to maintain and replace components inside
the vacuum vessel. Shielding inside the double walled vacuum vessel
would allow hands-on work in the region outside the TF magnets.

High speed tritium pellet injectors with vertical injection paths inside
the magnetic axis and lower speed pellets guided toward the inboard
mid-plane will be implemented to improve plasma performance and to
reduce tritium inventory. FIRE will also develop and test materials for
plasma facing components (W and Be) that are compatible with the fusion
reactor requirement of low tritium retention. A fast low inventory
tritium re-circulating system will be developed to minimize on-site
tritium inventory.

The combination of high magnetic field and advanced tokamak plasma
regimes could lead to very attractive fusion reactor concept with high
fusion power density. This emphasizes the importance of developing the
full capability of high temperature superconductors with high critical
fields and high strength structural materials for fusion applications.

**Status of FIRE Interim Engineering Study**

The NSO design team has significant experience from prior CIT, BPX and
ITER design studies and has been able to quickly assess the capabilities
of the interim FIRE design point. The toroidal magnet is comprised of 16
coils with 15 turns each of wedged plates of BeCu conductor. Detailed
3-D stress analyses have been carried out including the effects of
electromagnetic loads, and diffusion of the current and temperature in
the presence of nuclear heating. The TF copper stress is ≤76% of the
allowable stress and the temperature rise is within design allowables.
The TF insulation shear stress is within allowables except near the ends
of the central leg where dewedging occurs. This is a common problem in
wedged tokamaks and a solution is expected to be developed within the
current envelope. A design configuration with a toroidal field flattop ≈
18 second at 10 T has been developed.

The poloidal coils consist of a segmented OH solenoid with BeCu
conductor and shaping coils using OFHC conductor. The poloidal coil
stresses and temperatures are within the design allowables. The OH
solenoid coil stress is only \~ 40% of the design allowable for BeCu.
Higher conductivity copper alloys, with lower strength and cost, are
being evaluated for the OH solenoid. The vacuum vessel is a double wall
stainless steel vessel similar to that in the TPX and ITER designs. The
inter space of the double wall will be filled with shielding material to
reduce the neutron activation outside the toroidal coil envelope,
thereby allowing hands-on maintenance in the region outside the TF
coils.

The plasma facing components (PFC) must withstand very high power
densities while minimizing tritium retention. For the interim design,
carbon is excluded as a PFC material due to the high retention of
tritium experienced in TFTR and JET. While schemes might be developed to
mitigate the carbon tritium retention issue in FIRE, the mission of FIRE
is to develop a material that would also be compatible with the
requirements of a high duty cycle reactor. The leading candidate for the
first wall material is Be with inertially cooled tiles in the main
chamber and Be coating on copper backing plates for stabilizing plates.
The power density on the outer divertor plate without peaking factors or
the effects of elms is estimated to be \~ 25 MWm^-2^ without a detached
or radiative divertor. Tungsten rods mounted on Cu backing plates
capable of withstanding 25 MWm^-2^ have been developed for ITER and are
the leading candidate for the FIRE divertor plates. The outer divertor
plates will be actively cooled during the pulse while the inner plates
are cooled between pulses. Plasma modeling is underway to ensure
detached or radiative divertor operation so that there will be some
margin to accommodate the effects on peaking, elms and disruptions. The
engineering analyses of FIRE are evolving rapidly. The most up to date
information can be found at
[http://fire.pppl.gov](http://fire.pppl.gov/).

**Cost and Schedule Considerations for a Next Step Burning Plasma
Experiment.**

Cost estimates have not been done for FIRE except to note comparisons
with comparably sized devices, CIT (R = 2.14m, 11T) and BPX-AT (R =
2.0m, 10T), which had costs estimated to be in the range of \$700M for
experiments located at the TFTR site with about \$200M of site credits.
Some of the physical parameters that drive the costs are toroidal field
magnet energy, plasma volume and plasma surface area are compared for an
existing large tokamak and for potential Next Step Options in the
following Table V. Cost estimates, consistent with a preconceptual
design, will be developed by the end of FY 1999. Industrial
participation will be sought in this effort to minimize reliance on
historical databases that may not accurately reflect modern
manufacturing costs.

Table V. Comparison of some important cost drivers for burning plasma
devices.

+------------------+-----+--------+-----+-----+------+------+--------+
|                  | JET | I      | F   | CIT | BPX  | I    | ITER-  |
|                  |     | GNITOR | IRE |     |      | TER- |        |
|                  |     |        |     |     |      |      | EDA    |
|                  |     |        |     |     |      | RC   |        |
+------------------+-----+--------+-----+-----+------+------+--------+
| TF Coil          | H~  | 30 °K  | LN  | LN  | LN   | Nb~  | N      |
| Technology       | 2~0 |        | B   | B   |      | 3~Sn | b~3~Sn |
|                  | Cu  | Cu     | eCu | eCu | BeCu |      |        |
+------------------+-----+--------+-----+-----+------+------+--------+
| TF Magnet Energy | 1.6 | 4      | 4   | 8   | 8    | 41   | 100    |
| (GJ)             |     |        |     |     |      |      |        |
+------------------+-----+--------+-----+-----+------+------+--------+
| Plasma Volume    | 85  | 11     | 18  | 35  | 65   | 740  | 2000   |
| (m^3^)           |     |        |     |     |      |      |        |
+------------------+-----+--------+-----+-----+------+------+--------+
| Plasma Surface   | 147 | 36     | 60  | 85  | 130  | 640  | 1200   |
| (m^2^)           |     |        |     |     |      |      |        |
+------------------+-----+--------+-----+-----+------+------+--------+
| Cost Range (\$B) | \~  | \~0.5  | \<1 | \~  | \    | \~5  | \~10   |
|                  | 0.5 |        |     | 0.6 | ~1.5 |      |        |
|                  |     |        |     | F   |      |      |        |
|                  | F   |        |     | Y89 | FY92 |      |        |
|                  | Y84 |        |     |     |      |      |        |
+------------------+-----+--------+-----+-----+------+------+--------+

The schedule to attain an MFE burning plasma test bed is an important
consideration in the strategy and planning of the MFE program. Plans for
the ITER-EDA required 15 years after the initiation of construction
before D-T operation would be initiated; a 12 year construction period
followed by three years of hydrogen and deuterium operation before D-T
experiments. A similar time period would be required for ITER-RC. If
ITER-RC construction were initiated in 2002, then D-T operation would
begin \~ 2017. A small compact high field tokamak like IGNITOR or FIRE
is expected to have a construction period of \~ 7 years followed by 3
years of hydrogen and deuterium operation based on previous schedule
estimates for CIT and BPX. If construction were initiated on IGNITOR or
FIRE in 2002, then D-T operation could begin in \~ 2012, which is \~ 5
years later than the initiation of similar inertial fusion burning
plasma experiments using NIF.

**Siting Considerations for a Next Step Burning Plasma Option**

The siting requirements for a FIRE/IGNITOR class device is much reduced
relative to that required for an ITER class device, thereby
significantly reducing the time and cost. The lowest cost Next Step
Option of the FIRE/IGNITOR class would be to utilize an existing site
such as TFTR or JET. However, these sites might limit later upgrades,
and might preclude a natural progression of follow-on devices.
Consideration should be given to the possibility of a unified site for
the development of fusion which would support not only the development
of tokamak burning plasma experiments, but tests of burning plasmas in
other magnetic configurations as well as providing an infrastructure for
plasma based fusion technology facilities.

**The Modular Pathway for Magnetic Fusion Energy**

The Next Steps in the Modular Strategy beyond the existing programs and
facilities could include a compact high field tokamak(s) of the
IGNITOR/FIRE class, a superconducting high performance advanced tokamak
of the JT-60SU class and specialized fusion technology facilities to
address the first wall problem for MFE. The fusion technology facilities
might include intense neutron sources for materials damage studies and
facilities to investigate the viability of renewable first wall concepts
such as liquid surfaces facing the plasma. These facilities would feed
into an Assessment of Magnetic Fusion milestone in \~ 2015, which if
successful would lead to the construction of an Advanced Fusion
Integration Facility with characteristics similar to those of an
attractive MFE reactor. In this scenario a commercial fusion power plant
could be operating midway through the 21^st^ century. A more detailed
description of the Modular Strategy can be found on the World Wide Web
at \<http://nso.ucsd.edu/chapter3.pdf\>.

**High Leverage Activities and Issues**

Continued 3-D modeling of power and particle handling with experimental
benchmarks will be essential in defining the divertor design.
Experimental and theoretical studies to clarify the scaling of the L to
H-mode power threshold and the H to L mode power threshold for plasmas
significantly below the Greenwald density would be important in
optimizing the design for both standard regimes as well as advanced
reverse shear regimes. Much more work is needed to develop
self-consistent advanced tokamak modes of operation and to evaluate
stability to alpha driven modes. More comprehensive 1-D simulations of
standard and advanced regimes including burn control techniques are
needed.

**Summary**

Exploration, understanding and optimization of alpha-dominated plasmas
is a critical issue for all approaches to fusion. The tokamak is the
most cost-effective vehicle to investigate alpha-dominated plasma
physics, and its coupling to advanced toroidal physics for MFE. The
performance of a burning plasma depends sensitively on the details of
confinement, β-limits, density limits and edge plasma conditions. This
uncertainty can only be reduced by studying actual alpha-dominated
plasmas in the laboratory in conjunction with other advanced toroidal
experiments and improved numerical simulations. The compact high field
tokamak offers the possibility of addressing the important
alpha-dominated plasma issues, many of the long pulse advanced tokamak
issues and beginning the integration of alpha-dominated plasmas with
advanced toroidal physics in a \~\$1B class facility.

**Appendix 1. Basic Parameters and Features of FIRE**

  ----------------------------------- -----------------------------------
  R, major radius                     2.0 m

  a, minor radius                     0.525 m

  κ95, plasma elongation at 95% flux  \~1.8
  surface                             

  δ95, plasma triangularity at 95%    \~0.4
  flux surface                        

  q95, plasma safety factor at 95%    \>3
  flux surface                        

  Bt, toroidal magnetic field         10 T with 16 coils, \< 0.5% ripple
                                      @ OuterMP

  Toroidal magnet energy              3.7 GJ

  Ip, plasma current                  \~6.5 MA

  Magnetic field flat top, burn time  ≥10 s ( =21 s at 10 T, Pfusion \~
                                      200 MW)

  Pulse Repetition time               2 hr

  ICRF heating power, maximum         30 MW

  Neutral beam heating                None

  Lower Hybrid Current Drive          None in baseline, upgrade for AT

  Plasma Fueling                      Pellet injection (≥2.5km/s vertical
                                      launch inside mag axis, possible
                                      guided slower speed pellets)

  First wall materials                Be tiles, no carbon

  First wall cooling                  Inertial between pulses

  Divertor configuration              Double null, fixed X point,
                                      detached mode

  Divertor plate                      W rods on Cu backing plate (ITER
                                      R&D)

  Divertor plate cooling              Inner plate-inertial, outer plate
                                      active - water

  Fusion Power                        \~200 MW

  Fusion Power Density (plasma)       \~10 MW m-3

  Neutron wall loading                \~ 3 MW m-2

  Lifetime Fusion Production          5 TJ (BPX had 6.5 TJ)

  Total pulses at full field/power    3,000 (same as BPX), 30,000 at 2/3
                                      Bt and Ip

  Tritium site inventory              TBD (\< 30 g)
  ----------------------------------- -----------------------------------

Possibility of upgrading to 12T and 7.7 MA with a 12 s flat top has been
identified and is discussed in the Interim Engineering Report
(http://fire.pppl.gov).

**Appendix 2. Guidelines for Estimating Tokamak Plasma Performance**

Maximum density limit -

Fusion performance is optimized by operating at the maximum density.
Operational experience in tokamaks has shown that the maximum operating
density, often called the Greenwald limit. Plasma energy confinement is
observed to deteriorate as this limit is approached. While ITER design
guidelines allow the plasma density to approach or exceed the Greenwald
limit, FIRE has taken a more conservative approach and has restricted
the density to

n \< 0.75 nGW = 0.75 Ip/πa2

where n is the line average density in 10^20^ m^-3^, Ip is the plasma
current in MA and a is the minor radius in m. This limit is easily
satisfied for compact high field tokamaks which can have an operating
density \> 5 times higher than ITER. A nominal central plasma density
for FIRE is \~ 6.5 x 10~20~m^-3^ and IGNITOR central densities are
typically \~ 10^21^ m^-3^.

Plasma Energy Confinement -

The plasma in FIRE is presumed to operate in the "standard" high
confinement mode (H-mode) with small edge localized instabilities
(ELMs). This elmy H-Mode has been thoroughly studied on many tokamaks
and is most readily achieved when a divertor separatrix is present as is
the case for FIRE. The plasma energy confinement for this mode of
operation for existing tokamaks is described quite well (rms ≈ 15%) by

τE(ITER 98H) = 0.094 Ip0.97 R1.7 a0.23 n200.41 B0.08 Ai 0.2 κ 0.67 Pheat
-0.63

where Ip is the plasma current in MA, R is the major radius in m, a is
the minor radius in m, n20 is the plasma electron density in 10~20~
m~-3~, B is the toroidal magnetic field in T, Ai is the atomic mass, κ
is the plasma elongation and Pheat is the net power heating the plasma
in MW. Since the FIRE would operate at high density, the required
confinement time is only \~ 0.6 s which is within the range of
confinement times achieved by existing tokamaks of slightly larger size.

Pressure (β)Limit -

The fusion power density is given by \<σvi\>nDnT E~DT~ which can be
rewritten as

P~fusion~ /V~olume~ \~ \<σvi\>nDnT E~DT~ \~ β^2^B^4^

where β ≤ βN (I~p~/aB) and βN ≤ 2.5 for conventional tokamaks and ≤ 4
for advanced tokamaks with tailored pressure and current profiles. This
relation follows from the theory of MHD stability to macroscopic modes
and has been verified by many experiments. Since I~p~ \~ aB for fixed
q~95~ and plasma shape, this relation illustrates the importance of
operating at high magnetic field since the power density goes as B^4^.

H-mode Power Threshold -

The high confinement mode can be accessed when the plasma heating power
transported across the plasma surface exceeds a threshold value given by

Pheat ≥ Pth = (0.9/Ai) n20^0.75^ B R^2^.

Once the plasma has transistioned to the H-mode, it remains there as
long as P exceeds about 1/2 Pth. This hysteresis is reduced as the
operating density approaches the density limit. This relationship was
derived initially from an analysis of experimental data. Recently,
theoretical models have been developed that describe some of the general
features of this transition. This requirement is a significant
operational constraint for FIRE.

Particle and Helium Ash Confinement -

A broad range of detailed experimental studies has shown that the
particle confinement within the plasma is given by τp ≈ τE for the elmy
H-mode. After the ionized plasma particles escape, they can be
neutralized on the plasma facing components and a fraction, R, recycled
back to the plasma where they are re-ionized. Similarly, the slowed down
alpha ash has been shown to behave like other plasma particles. The
recycling coefficient is determined by the effectiveness if the
divertor, in FIRE, we assume that 80% of the particles are recycled back
to the plasma. Therefore, the effective confinement time for the helium
ash, τHe \* ≈ τE /(1-R) = 5 τE

**Appendix 3. Nominal FIRE Plasma Parameters**

  -------------------------------------------- --------------------------
  R, plasma major radius, m                    2

  A, plasma minor radius, m                    0.525

  R/a , aspect ratio                           3.8

  κ_95, plasma elongation at 95% flux          1.77

  δ_95, plasma triangularity at 95% flux       0.4

  q_95                                         3.02

  B_t, toroidal magnetic field, T              10

  I_p, plasma current, MA                      6.44

  l_i(3), internal plasma inductance           0.8

  Fraction of bootstrap current                0.25

  Ion Mass, 50/50 D/T                          2.5

  \<ne\>, 10\^20 /m\^3, volume average         4.5

  α_n, density profile peaking = 1 + α_n       0.5

  **\<n\>l/Greenwald Density Limit, ≤ 0.75**   **0.70**

  \<T\>n, density averaged temperature, keV    8.2

  T(0), central temperature, keV               13.1

  α_T, temperature profile peaking = 1 + α_T   1

  Impurities, Be:high Z, %                     3 : 0

  Alpha ash accumulation, n_α/n_e, %           2.6

  Zeff                                         1.41

  ν\*, collisionality at q = 1.5               0.043

  P_ext , MW                                   22

  P_fusion, MW                                 223

  P_heat , MW                                  56.5

  tau_p\*(He)/tau_E                            5.00

  tau_E, energy confinement time s             0.57

  **ITER98H-multiplier, ≤1**                   **1.04**

  ITER89P - Multiplier                         2.41

  nd(0)T(0)τΕ , 1020 s keV                     41.69

  Q_DT                                         10.16

  IA                                           24.5

  Plasma current redistribution time, s        13.9

  **Pheat/P(L-\>H), ≥ 1**                      **1.149**

  W_p, plasma thermal energy, MJ               32.18

  β_total, thermal plasma + alphas, %          3.11

  **β_N, ≤ 2.5**                               **2.54**

  Core Plasma Pressure, atmospheres            \~ 20
  -------------------------------------------- --------------------------