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\begin{document}

\title{\bf 1. EXECUTIVE SUMMARY}
\author{\hbox to \hsize
{ \hbox to 2.truein{ Ronald L.~Miller \hfil } \hfil
  \hbox to 2.truein{\hfil Farrokh Najmabadi  \hfil } \hfil
  \hbox to 2.truein{\hfil Xueren Wang} } \\ }
\date{ }
\maketitle

\begin{titlepage}
{%\baselineskip=1.8ex
\tableofcontents
}
\end{titlepage}

\setcounter{chapter}{0}
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\chapter{EXECUTIVE SUMMARY}

\section{INTRODUCTION}
\label{s-1-1}

The Stellarator Power Plant Study (SPPS) examines the key issues posed by the
\st approach to future magnetic fusion central-station electric power plants.
While the tokamak remains the primary approach to magnetic confinement of
fusion plasmas in the various worldwide fusion R\&D programs, the \st approach
portends a viable, complementary alternative.  From among the broad range of
configurational options encompassed by the \st class, a modular-coil,
four-field-period was selected as a baseline configuration for the present
study, representing a consensus recommendation of the U.S. \st physics
community of an attractive baseline concept, coupled with modern
magnetic-fusion engineering design practice, incorporating safety and
environmental considerations, consistent with the recent
ARIES~\cite{ARIES4} tokamak conceptual designs.

The SPPS was launched and technical work was largely completed prior to the
recent fundamental restructuring of the U.S. fusion program~\cite{Dav96} in the
face of budgetary reductions.  Concept innovation and attention to alternative
(non-tokamak) magnetic fusion concepts, with a view toward improved concepts
for energy applications, are emphasized as themes of the strategy of the new
program.  Also, given the several strong \st programs outside the U.S., an
opportunity to strengthen international collaboration has been afforded by this
work.  The results of the SPPS contribute to the assessment of the \st
approach as a promising alternative concept~\cite{Naj96}.

Intrinsically steady-state, the \st requires no net plasma current and avoids
problems associated with (tokamak) current-drive requirements, control with
high bootstrap current fraction, major disruptions, and positional control
systems and instabilities.  While sharing a list of generic concerns with the
tokamak, specific reservations regarding complexity and relatively low system
power density have traditionally been attributed to the \st.  These issues are
addressed by the SPPS to elucidate those high-leverage factors influencing the
attractiveness of this class of power plant.

Generally, conceptual power-plant design studies extrapolate from the present
knowledge base to characterize the future economic and environmental features
of competing approaches.  Such studies are increasingly guided by an awareness
of the needs~\cite{Kas94} of the potential customer, be it a traditional
utility or a non-utility generator (NUG) that might emerge in a (future)
deregulated electricity market.  The customer stresses competitive economics,
public acceptance, and regulatory simplicity as desireable generic features of
any suitable generating option.  The conceptual power-plant design study
provides a basis for the comparative assessment of feasible alternatives and
contributes to the quantitative decision-making process characteristic of
capital investment in genertating capacity or, even, of competing R\&D
programs.

The SPPS addresses a widely perceived need to consider nontokamak alternative
power-plant options in the post-ITER~\cite{Aym96} period.  The SPPS assumed a
deuterium-tritium (DT) burning plasma and the need to incorporate a
lithium-bearing, tritium-producing blanket system.  The steady-state coil
system was based on superconducting technology.  Passive-safety considerations
are reflected in low-activation materials choices and limits in (afterheat)
power density.  Maintenance issues are addresed in the context of assuring high
plant availability and consequently favorable economics.

\section{THE STELLARATOR CONCEPT}
\label{s-1-2}

The \st is a toroidal magnetic confinement approach producing closed magnetic
confining surfaces by means of external conductors exclusively.  The \st is
one of the earliest approaches to magnetic fusion~\cite{Spi51,Spi54,Spi58}.  A
number of review articles discussing the \st concept have been
published~\cite{Miy78,Sho81,Car88,Lyo90a,Rau94}.

The term ``stellarator'' (inclusive of torsatron and heliotron configurations)
refers to a class of helical ({\it i.e.}, toroidal plus poloidal field
components) magnetic fusion devices that produces confining magnetic fields by
currents in external coils only.  Relative to the tokamak, the \st promises
disruption-free, steady-state operation with reduced recirculating power due to
the absence of current-drive requirements.  Such advantages may be offset by
more complicated coil, blanket/shield, and divertor ({\it i.e.}, impurity
control) configurations.

The fundamental \st configuration and the anticipated higher plasma aspect
ratio, \hbox{$A^{\star} \equiv R_{T}/r_{p} \simeq 8.5$}, relative to the
tokamak also result in different overall performance as monitored by the usual
figures of merit ({\it e.g.,} power density, stored magnetic energy, peak coil
magnetic field strength, neutron wall load, and projected cost of electricity,
COE).

The future payoff of the \st approach is a possible post-ITER candidate for an
attractive demonstration (Demo) power plant and subsequent commercial power
plant, fueled with deuterium-tritium (DT), with attractive features and
performance relative to the tokamak.  Generally, the development of suitable
structural materials and nuclear technologies can be anticipated to proceed
under the (tokamak) ITER program, without unnecessary duplication for \st
applications.

\section{SCOPE OF THE STUDY}
\label{s-1-3}

The Stellarator Power Plant Study (SPPS) can be said to follow from a letter
proposal~\cite{Pri90} made in 1990, proposing a three-stage study:
(1)~Optimization of the stellarator (physics) concept; (2)~Integrated systems
optimization; and (3)~Realistic engineering development.  A charter for a
multi-year study was developed to ``identify and assess the feasibility of
critical issues and their consequences for development of the stellarator
concept as a steady-state fusion reactor.  The approach will be similar to that
used in the ARIES and TITAN RFP reactor studies: an integrated physics,
engineering, reactor component, and cost optimization.  The same assumptions
and algorithms can be used to compare different stellarators and tokamaks.  The
activities during the first year will focus on physics optimization and
selection of one or more stellarator coil configurations for more detailed
engineering design evaluation.''

The U.S. SPPS began in 1993 with the investigation of candidate configurations.
Interim results were reported~\cite{Lyo93a,Lyo94a,Lyo94b}.  Following
investigations that included consideration of a continuous, helical coil
configuration~\cite{Lyo94a}, a consensus emerged around the selection of the
modular helias-like heliac (MHH) as the reference approach.  Some preliminary
results are summarized elsewhere~\cite{Mil94}, but the present report
represents the most comprehensive documentation of the work.

The Engineering Phase, conducted in CY1994-5, adapted technologies from ARIES
tokamak power-plant conceptual designs, where applicable, and considered the
key configurational aspects of the stellarator.  The latter phases of the SPPS
effort concentrated on the development of reference power plant parameters,
conceptual engineering of key subsystems, and integration of the Fusion Power
Core (FPC), encompassing the magnetic-coil set, nuclear systems ({\it i.e.},
shields and tritium producing blanket), and plasma-facing components ({\it
i.e.}, first wall and divertor surfaces).

Limited in scope and depth relative to the ARIES tokamak-based studies, the
SPPS focussed on the consensus identification of a reference configuration,
{\it i.e.}, the modular, helias-like heliac (MHH), and applied engineering
assumptions, constraints, and extrapolations consistent with the ARIES tokamak
studies to address key issues for the \st approach.  Projected Cost of
Electricity [COE (mill/kWeh)] estimates are reported as a figure of merit.  The
SPPS does not purport to be a fully integrated engineering design.

Participants in the study included a broad representation of the \st
physics community, as well as a subset of the multi-institutional ARIES Team,
augmented by specialized industrial support.  Many other scientists attended the
SPPS project meetings and contributed to the development of the concept.  They
are identified in the acknowldegements.

The SPPS technical work was coordinated by a series of workshops and project
meetings involving the participants, as follows:

\noindent$\bullet$ Feb. 3, 1993 \ USDOE, Germantown, MD \hfil\break
$\bullet$ Feb. 16-17, 1993 \quad Oak Ridge, TN \hfil\break
$\bullet$ May 5-7, 1993    \quad Los Angeles, CA \hfil\break
$\bullet$ Jul. 24, 1993    \quad Madison, WI \hfil\break
$\bullet$ Nov. 2, 1993     \quad St. Louis, MO \hfil\break
$\bullet$ Jan. 5-7, 1994   \quad Oak Ridge, TN \hfil\break
$\bullet$ Mar. 13, 1994    \quad Dallas, TX \hfil\break
$\bullet$ Mar. 17, 1994    \quad La Jolla, CA \hfil\break
$\bullet$ Jul. 12, 1994    \quad Chicago, IL \hfil\break
$\bullet$ Sep. 12-13, 1994 \quad Lisle, IL \hfil\break
$\bullet$ Nov. 6, 1994     \quad Minneapolis, MN \hfil\break

\section{REFERENCE CONFIGURATION}
\label{s-1-4}

The \st magnetic-coil configuration can be produced by continuous, helical
coils.  The most recent example of this approach is that of the Large Helical
Device (LHD)~\cite{Fuj96,Mot96}, presently under construction in Japan.  It is
possible to project this basic configuration to commercial power plant
scale~\cite{Sag95}.  Alternatively, modular-coil configurations, represented by
that of the Wendelstein 7-X~\cite{Sap96}, which is also under construction in
Germany.  A corresponding power plant design has been developed~\cite{Bei95}.
The coil configuration adopted for the present SPPS is also of the modular
variety and a variant of the Wendelstein/Helias type.  The baseline MHH
configuration has four field periods, produced by 32 modular, non-planar coils
of four distinct types, as illustrated in \hbox{Fig.~\ref{f-1-4-1}}.

\begin{figure}[p]
\centerline{\psfig{file=spps-1-4-1.ps,width=6.25in,bbllx=88pt,bblly=179pt,%
            bburx=523pt,bbury=614pt,clip=}}
\vspace{2.5cm}
\caption[Fig-1-4-1]       %       950910_SP_blktcoil.ps
{MHH modular-coil configuration, showing the last-closed flux surface of the
plasma and the blanket/shield envelope interposed between the plasma and
the coil set.  The scale of the configuration is not specified here.}
\label{f-1-4-1}
\end{figure}

The SPPS projects a 1,000-MWe(net), DT-fuelled central power station,
with key parameters summarized in \hbox{Table~\ref{t-1-4-1}}.  The target net
electrical power output (design electrical rating) was selected to be the
same as that of the ARIES and other recent conceptual designs to permit
direct-cost (and other) comparisons.  An alternative approach would be
to postulate a common balance of plant ({\it i.e.,} same gross electrical
power), letting the net power output of the \st increase as a
consequense of the lower recirculating power fraction, and gaining a
cost advantage as a result of so-called `economies of scale'.  There is
engineering margin in the coil design and high-heat-flux surfaces to
accomodate this option.

\begin{table}[p]
\caption{}
\centering{\bf Major Parameters of the SPPS Power Plant}
\vspace{1.0cm}
\renewcommand{\arraystretch} {1.15}
\begin{tabular} {l c }
\hline \\
Number of field periods                           & 4 \\
Number of modular, non-planar coils               &32 \\
Plasma major toroidal radius, $R_{T}$ (m)         & 13.95 \\
Plasma half-width, $a_{p}$ (m)                    &  1.16 \\
Plasma aspect ratio ({\it cf.}, tokamak), $A \equiv R_{T}/a_{p}$ & 12.08 \\
Plasma aspect ratio, $A^{\star} = R_{T}/r_{p}$    & 8.54 \\
Circularized (average) plasma radius, $r_{p}$ (m) &  1.63 \\
Plasma volume, $V_{p}$ (m$^3$)                      &734.7  \\
Plasma beta                                   &  0.05 \\
Lackner-Gottardi confinement multiplier, H    &  2.3  \\
On-axis magnetic-field strength, $B_{o}$ (T)      &  4.94 \\
Peak magnetic-field strength at coil, $B_{c}$ (T) & 14.5  \\
Stored magnetic energy, $W_{B}$ (GJ)              & 80 (est.) \\
Fusion power, $P_{F}$ (GWth)                      &  1.73 \\
Thermal conversion efficiency, $\eta_{TH}$    &  0.46 \\
Thermal power, $P_{TH}$ (GWth)                    &  2.29 \\
Gross electrical power, $P_{ET}$ (GWe)            &  1.05 \\
Net electrical power, $P_{E}$ (GWe)               &  1.0  \\
Recirculating power fraction, $\epsilon$      &  0.052\\
Plant capacity factor, $p_{f}$                    &  0.76 \\
Total direct cost  (B\$)$^{(a)}$                 &2,249.   \\
Total capital cost (B\$)$^{(b)}$                 &4,340.   \\
Cost of electricity, COE (mill/kWeh)$^{(c)}$          & 74.6  \\
\\ \hline
\multicolumn{2}{l}{$(a)$ Assumes 10th-of-a-kind learning-curve and
     certain safety-related cost credits.}\\
\multicolumn{2}{l}{$(b)$ Assumes 6-yr construction lead time.}\\
\multicolumn{2}{l}{$(c)$ Levelized COE, reported in (constant)
1992 U.S. dollars (1 mill = 0.001 dollar).}
\end{tabular}
\label{t-1-4-1}
\end{table}

Recirculating power is low for the \st because of no current drive is required,
as for tokamaks.  System power density is typically lower because of the
requirement for nominal stand-offs at higher plasma aspect ratio.  The
engineering configuration is complicated by the helical plasma geometry, but
fundamental choices of materials and coolants are shared with the tokamak
approach.  A critical design constraint is the desire to minimize the standoff
between the inboard plasma edge and the coil winding pack consistent with
adequate tritium breeding in the lithium-bearing blanket and sufficient
neutronic shielding of the coils.  The reference blanket configuration
invokes a vanadium-alloy structure (with CaO-coated coolant channes to reduce
MHD pumping power) with low-activation properties that is cooled by flowing
liquid lithium, which serves as the tritium-breeding material and the primary
coolant.  Operation of the blanket at high temperature allows an advanced
thermal cycle with gross efficiency, $\eta_{TH}=0.46$, as for the ARIES
tokamaks.  A nominal plant factor, $p_{f}=0.76$, reflects an availability
equal to that taken for the ARIES tokamaks.  A more detailed comparative
assessment would have to be mounted to make differential estimates of forced
and scheduled outages, if the plant factor were to be used as a discriminator
between stellarators and tokamaks.

The incorporation of low-activation materials allows the SPPS design to be
passively safe under typical accident scenarios.  Cost credits, reflecting the
Level of Safety Assurance (LSA)~\cite{Hol89}, contribute to lower direct costs
and lower COE projections, relative to a device relying on active safety
measures [{\it e.g.}, Emergency Core Cooling System (ECCS) as for present
fission plants].

The SPPS Fusion Power Core is illustrated conceptually in
\hbox{Fig.~\ref{f-1-4-2}}.  Major systems are included, but such items as
diagnostics, fuelling systems, coil busbars, {\it etc.}, that tend to obscure
the device itself, are omitted in the interest of clarity.  With artistic
license, the plasma (last closed flux surface) is shown despite the vacuum
having been broken during a maintenance withdrawal of a corner sector
containing several modular coils and the underlying systems.

\begin{sidewaysfigure}[p]
\centerline{\psfig{file=spps-1-4-2.ps,height=6.truein,bbllx=130pt,bblly=110pt%
                   ,bburx=535pt,bbury=680pt,angle=-90,clip=}}
\caption[f-1-4-2]
{Isometric view of the SPPS Fusion Power Core.}
\label{f-1-4-2}
\end{sidewaysfigure}

\section{TECHNICAL ASSESSMENT}
\label{s-1-5}

The Stellarator Power Plant Study (SPPS) examines the key issues posed by the
\st approach to magnetic fusion central station electric power plants.  The
remainder of this section examines the key findings of the study.

\subsection{Plasma Physics}
\label{s-1-5-1}

Physics issues for the MHH include equilibrium/stability at the reference beta
value of 5\%, the influence of magnetic islands on divertor performance,
neoclassical transport in the presence of electric fields, suppression of the
bootstrap current, and startup/control scenarios.  MHH physics performance is
projected to be adequate and robust.

The MHH configuration was chosen because its helias-like features led to good
physics properties and its heliac feature, the large helical excursion of the
magnetic axis that provides most of the rotational transform, allows the coils
to be farther from the plasma, permitting a 2-m distance between the plasma and
the coils for blankets, shields, structure, clearance, {\it etc.} for a 14-m
average major radius.  This new configuration has the best physics properties
(beta limit, neoclassical transport, orbit losses, reduction of bootstrap
current, {\it etc.}).  However, its smaller plasma-coil separation leads to a
larger major radius (larger blanket and shield area) and consequently higher
cost.

\subsection{Engineering}
\label{s-1-5-2}

Engineering emphasis in the SPPS is on the coil design, neutronics
optimization, integration of the blanket/shield/divertor systems into the MHH
configuration, and assessment of a suitable maintenance approach.  A suitable
divertor/blanket/shield configuration can be accommodated in MHH-based fusion
power plants and maintained.  Preliminary, but serious, assessment of \NB\ CICC
graded, modular, nonplanar coils at 14.5~T is encouraging.  The MHH as examined
in the SPPS was found to be cost competitive with tokamak approaches.  No
unit-cost penalty was assessed for the complications introduced by the
out-of-plane winding of the MHH coils.  Such coils would be fabricated at a
factory under `clean' conditions and shipped to the plant site for
installation, as for the ARIES tokamaks.  For power-plant-scale installations,
this approach to modularity is generally viewed as advantageous to the {\it in
situ} winding of large continuous helical coils.  Similarly, although
life-of-plant coils are typically (and wishfully) assumed, the rare replacement
of a failed modular coil is more acceptable than the repair/replacement of a
continuous coil.  Further reductions in the overall size of the fusion power
core, with corresponding increases in the system power density, may be required
to improve competitiveness.

\section{SUMMARY AND CONCLUSIONS}
\label{s-1-6}

Detailed conclusions may be found in the body of this report.
In concise form, the following conclusions of the SPPS have been drawn:

\begin{itemize}

\item
The Modular Helias-like Heliac (MHH) configuration, selected for emphasis in
the SPPS, is one of a number of credible stellarator embodiments.

\item
The MHH physics performance is projected to be adequate and robust.

\item
A suitable divertor/blanket/shield configuration, based on
structural/breeder/coolant choices in common with tokamak technologies, can be
accommodated in MHH-based fusion power plants and maintained.

\item
Preliminary, but serious, assessment of Nb$_{3}$Sn CICC graded, modular,
nonplanar coils at 14.5~T is encouraging.

\item
The reference SPPS MHH design point, with low system power density and low
recirculating power, is found to be cost competitive with tokamak MFE designs.

\item
Interesting variations of the MHH configuration were not explored in detail,
but could provide improved performance.

\end{itemize}

\section{ORGANIZATION OF THE REPORT}
\label{s-1-7}

This SPPS design report is devided into six major sections.
Section 2 summarizes the physics basis of the Modular Helias-like Heliac (MHH).
Section 3 summarizes the systems analysis activity supporting the selection
of the reference design point.  Section 4 provides a detailed discussion of
the MHH coil configuration, internal design, and technical analysis.
Section 5 describes the design of the SPPS fusion power core, emphasizing
the issues of materials selection, neutronics, as well as configurational
layout and maintenance.  Section 6 examines several issues relating to
safety and environmental aspects of the SPPS having broad application
to the stellarator approach.  Some detailed information appears in an
Appendix.

\clearpage
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\end{thebibliography}
\end{document}

\bye
