IEC Electric Powerplants

Transcription

Preliminary Study of Inertial!
Electrostatic!Fusion "IEF#
for Electric Power Plants
EPRI TR-103394s — Report Summary, February 1994
This study of IEF systems focused on the use of
deuterium!deuterium "DD# as the principal fusion fuel
source, operating with the immediate burn of the recov!
ered helium, produced in one branch of the reaction
process. Such IEF systems o$er far greater promise at
lower cost and less development time than the current
magnetic tokamak approach. The IEF study discusses
physics issues, plant concepts, engineering constraints,
hazards, stability, R&D characteristics, and development
paths and costs.
Background
IEF systems use the kinetic energy of ions projected
radially inward by the imposition of an accelerating po!
tential gradient outside the central core region of a
spherical system. By this means, ions will converge to!
ward the center point of such systems, increasing in
density and energy. At the system center, they will either
undergo fusion reactions, producing fusion products
that leave the system at high energy, or be scattered
back up into the conning electric eld. This accelerat!
ing electric eld can be provided either by a set of
spherical grids biased so as to accelerate the positive
ions inward "called an IXL system# or by injection of
energetic electrons into a magnetically conned quasi!
spherical geometry "an EXL system#.
Objective
To identify fusion fuel combinations and system speci!
cations that o$er the most promise for near!term prac!
tical application to utility central station power systems.
Approach
The project team analyzed plasma particle and potential
distributions using the existing EIXL code for IEF
sources. Next they used the PBAL code to perform
power balance calculations fora large number of para!
metric cases for fuel combinations of interest. In addi!
IEF in Electric Power Plants%
tion, the team analyzed neutron output inherent in a
deuterium!bearing system as well as tritium inventory,
stored plasma energy, and stored magnetic eld energy.
Finally, they examined the critical proof!of!principle
issues for IEF systems. Specically excluded from this
project was the commonly studied fuel combination of
deuterium!tritium, due to its prohibitively high neutron
output and associated damage potential. This report
received peer review, which noted that the IEF concept
could meet utility requirements for a desirable reactor
better than any other fusion concept proposed to date.
Results
One branch of the DD fusion cycle results in the pro!
duction of 3He, making it feasible to capture and re!
inject this helium to produce D + 3He fusion reactions.
This process is commonly called the DD!1/2!catalyzed
cycle. Two alternative fuel cycles exist as potential up!
grades to the basic DD!1/2!catalyzed cycle. These in!
clude burn of additional 3He produced from the decay
of tritium from one branch of the DD reaction and the
reaction of p "hydrogen# with 11B to provide clean 4He
production with high potential for direct conversion.
Use of the PBAL code analysis to assess overall IEF sys!
tem performance demonstrated that large power gains
"the ratio of gross electrical power output to required
input power# are achievable in systems of modest size
within standard stress parameters of conventional mate!
rials. Analysis of neutron output showed that neutron
damage e$ects need be no more severe than those in
current ssion reactors. Finally, the hazards analysis
showed that tritium inventory, stored plasma energy, and
stored magnetic eld energy were hundreds of times less
than in conventional magnetic tokamak reactor systems.
EPRI Perspective
If IEF systems prove viable, they could provide useful
fusion power plants decades sooner than conventional
fusion concepts. In addition, IEF plants will o$er
1
cleaner, less hazardous performance with upgrade poten!
tial to totally clean systems. Freedom from empirical
scaling laws enables IEF systems to be studied conclu!
sively in simple experiments at scaled models, thus has!
tening the development to deployment process. Further
study and analysis remains to be performed to deter!
mine specic engineering aspects of the DD!1/2!
catalyzed cycle plant systems. The next step involves
conducting detailed experiments of the critical physics
to provide proof of feasibility of the IEF approach.
Project RP8012-16
Project Manager: Robert L. Hirsch, Washington Of!
ce
Managed for: O"ce of Exploratory and Applied Re!
search
Contractor: Energy/Matter Conversion Corporation
#EMC2$
Note
Report reformatted into two columns, i!ustrations updated and
implemented in color, and scientic notation used throughout i"
November 2008.
2
Preliminary Study of
Inertial!Electrostatic!Fusion "IEF#
for Electric Utility Power Plants
EMC2, Final Report, February 1994, EPRI Research Project 8012-16
Abstract
Study of inertial!electrostatic!fusion "IEF# systems re!
vealed that the use of deuterium!deuterium "DD# as the
principal fusion fuel source o$ers the most promise for
near!term practical application to utility central station
power systems. The DD!1/2 catalyzed cycle can be im!
plemented to generate electrical power through lower
cost conventional thermal conversion faster than any
other system. Use of the PBAL code to assess overall
IEF system performance demonstrated that large power
gains are achievable in systems of modest size within
standard stress parameters of conventional materials.
Analysis of neutron output showed that neutron damage
e$ects need be no more severe than those in current
ssion reactors. Finally, the hazards analysis showed that
tritium inventory, stored plasma energy, and stored mag!
netic eld energy were hundreds of times less than in
conventional magnetic tokomak reactor systems.
3
C. Recommended Actions!
Table of Contents
9. References!
Glossary!
5
Executive Summary!
6
2. Statement of Work!
A. System Studies!
B. Physics Studies!
C. Engineering Studies!
D. Plant/System Development Plan!
E. Deliverables!
7
7
7
7
8
8
3. Introduction!
A. Origins and History!
B. Basic Approaches and Characteristics!
C. Summary of Current Status!
59
60
10
10
15
17
4. Plasma Physics Technical Features!
19
A. Particle and Potential Distributions!
19
B. Virtual Anodes, Core Convergence, and Multiple
Wells!
21
C. Plasma/Potential Distribution Stability!
25
D. System Constraints from Physics Features!
25
5. IEF Power Systems!
A. IEF Plasma Connement Performance!
B. Plasma Power Generation and Losses!
C. Fusion Fuels; Characteristics and Choices!
D. General Power Plant Concepts!
E. Power Balance in IEF Systems!
27
27
32
33
35
38
6. Engineering Issues!
A. Conversion of Fusion Products and Energy!
B. Impurities and “Ash” Production!
C. System Safety!
D. Neutron Production and Materials Damage!
E. System Control and Stability!
43
43
44
45
47
49
7. Development and Deployment!
53
A. Critical Physics Issues.!
53
B. R&D Paths and Programs; Proof"of"Feasibility!54
C. Long"Term Development to Commercial Power;
Strategy and Schedules!
56
8. Conclusions and Recommendations!
A. The Promise of IEF Systems.!
B. Needed Further Study and Analysis!
IEF in Electric Power Plants!
58
58
59
4
spherical geometries. This is a “1.5”
dimensional code that is a boundary"
value Poisson"solver for the simplest
model of IEF systems. It is built so as
to handle both/either EXL or IXL
types of systems. It analyzes DD ion
systems $scales to other fuels% and ac"
cepts parametric values of system ra"
dius, B eld, currents, drive voltages.
and specied virtual anode height $used
as a control mechanism/variable%.
Glossary
IEF !
Inertial"Electrostatic"Fusion: a general
descriptor term used to refer to any
version of spherically"converging ow
devices driven by external electric
elds.
IEC !
Inertial"Electrostatic"Connement: a
general descriptor used in the past to
refer to the systems more properly
called IEF systems. The use of the
word “connement” is unfortunately
quite misleading. as the ions and elec"
trons # while conned in the external
envelope of the machine # are not
conned in the reacting central region
$sometimes referred to as the “core”% of
the device by the applied electric po"
tentials. They all pas through this re"
gion but occupy it for only a small frac"
tion of their lifetime in the “conned”
total system.
SCIF !
Spherical"Converging"Ion"Flow system.
An acronym used in the DARPA pro"
gram initially to dene the experiment
undertaken therein, and later applied
to IEF generally.
Polywell!
A trademarked name used to refer to
the polyhedral magnetic"connement
system used to conne electrons in the
electron injection version of the rEF
system. This is the device concept in"
vented and patented by Robert W. Bus"
sard that formed the basis for the re"
cent DARPA program.
EXL!
Electron ACCELeration version of the
IEF SCIF approach; used alternatively
with Polywell to refer to electron"
driven potential well ion"trapping IEF
systems.
IXL !
Ion ACCELeration version of the IEF
SCIF approach. This refers to those
systems without magnetic elds that
use direct $gun or grid% acceleration of
ions to forms central dense ion clouds
for IEF. This is the approach used by
Robert L. Hirsch and Philo Farnsworth
in their 1960’s studies of DD and DT
fusion in small scale IEF systems.
EIXL $code%!
Acronym referring to the Vlasov"
Maxwell code now used to analyze par"
ticle and potential distributions in
EXL $code%!
An earlier version of the EIXL code,
that could handle only EXL systems.
GJ $code%!
A computer code that analyzes electron
trapping and recirculation in EXL sys"
tems, including both mirror reection
$MR% and diamagnetic “wi&e ball”
$WB% electron connement e'ects.
This code is an essential piece of the
the EIXL code and of the power bal"
ance analysis code PBAL $below%.
PBAL $code%!
A computer code that determines scal"
ing of EXL $and IXL by analogy% sys"
tems with parameters of size, drive
voltage, drive current. virtual anode
height $desired%, external B eld $in
EXL systems%, fuel mixtures $DT to
p6Li, p11B, et al%, particle transverse
momenta, system gross gain $inverse of
recirculating power fraction%, power
conversion mechanisms and e(cien"
cies, bremsstrahlung temperatures, and
gross and net power output.
5
Executive Summary
bearing system# showed that neutron damage e$ects
need be no more severe than those in current ssion
reactors.
Study was made of the characteristics of and prospects
for inertial!electrostatic!connement "IEC# means use!
ful for inertial!electrostatic!fusion "IEF# systems, and of
these IEF systems themselves. This work covered both
types of IEF systems; those that operate with direct
grid!biased acceleration of ions towards a convergent
central core1,2 "called IXL systems#, and those that oper!
ate by quasi!spherical magnetically!conned electron!
driven potential well radial acceleration of ions3 "called
EXL systems#. Study was made of physics issues, plant
concepts, engineering constraints, hazards, stability,
R&D characteristics, and development paths and costs
to useful power plants. It was found that IEF systems
o$er far greater promise at less cost and much shorter
time scales than for the magnetic tokamak approach of
the mainline DoE fusion program.
Other hazards studied included T inventory "from the
tritium branch of the DD reaction#, stored plasma en!
ergy and stored B eld energy "EXL#. These were found
to be hundreds of times less than in “mainline” Mag!
netic and Maxwellian "M&M# tokamak fusion reactor
systems of other studies. The generation of fusion “ash”
and of wall!sputtered impurities was analyzed and
shown to be insignicant, as was 14N and 14C generation
in p11B systems. All of these were found to be limited to
stable densities less than 10+ of fuel ion particle densi!
ties in the IEF system. A brief summary of IEF device
concepts and power systems applications is given in a
technical note in Appendix B, herewith.
The work was directed towards those fusion fuel combi!
nations that o$ered most promise for near!term practi!
cal application to utilities central station power systems,
and to materials choices and power conversion tech!
nologies that mirrored or were less stringent than cur!
rent practice in conventional nuclear "ssion# power
plants. This focussed study on use of DD as the princi!
pal fusion fuel source, operating with immediate “burn”
of the 3He produced in one branch of the DD reaction
process. This is called the “DD!1/2!cat” cycle. The com!
monly studied fuel combination of DT was specically
excluded from this study, due to its prohibitively high
neutron output, and associated damage potential. Two
alternative fuel cycles were studied as potential upgrades
to the basic DD!1/2!cat cycle. These included burn of
additional 3He produced from decay of tritium from the
other branch of the DD reaction "a DD!full!cat reac!
tion#, and reaction of p with 11B to give clean 4He pro!
duction, with high potential for direct conversion.
Plasma particle and potential distributions were ana!
lyzed by an existing Vlasov Maxwell code "the EIXL
code# for such IEF sources, and a power balance com!
puter code was devised to allow assessment of overall
system performance. This code "PBAL# includes losses
to magnet power, both normal and superconducting "for
EXL systems#, grid heating "IXL systems#, electron and
ion particle losses, bremsstrahlung, and system conver!
sion options and e%ciencies. Power balance calculations
were made for a very large number of parametric cases,
for the fuel combinations of interest. These showed that
system power gains "ratio of gross electrical power out!
put to required input power# can be made acceptably
large "> 8x# in systems of modest size "<3.5 m radius#,
within temperature, heat transfer and stress limits of
high strength stainless steels, at normal steam cycle con!
ditions. Analysis of neutron output "inherent in any D!
IEF in Electric Power Plants&
Examination was made of the critical issues confronting
proof!of!principle for the IEF systems and of the critical
physics issues governing such proof. From this it became
clear that a fundamental di$erence exists in the nature
of development between these relatively small IEF sys!
tems and conventional large!scale M&M tokamak fusion
systems. This is that IEF systems are all governed by
classical physics issues, while the M&M systems are not
but are determined by empirical “scaling laws.”
This has the consequence that IEF systems can be stud!
ied conclusively in simple experiments at their "small#
full!scale size, while M&M systems must be tested
through a succession of ever!larger experiments to test
continuing function of the scaling laws as very large sizes
required for net power are sought. As a result, it is pos!
sible to conduct development to deployment status very
much more cheaply and quickly than for the M&M
route. If IEF systems are found to work they can then
give useful fusion power plants decades sooner and at
very much less cost than for M&M concepts. And, they
o$er cleaner and less hazardous plants, with upgrade
potential to totally clean systems that are forever inac!
cessible to M&M concepts.
Further study and analysis remains to be done to deter!
mine startup and shutdown control sequences and lim!
its, and to further detail the engineering aspects of the
DD!cat cycle plant systems. Blanket heat loads, materi!
als activation, and optimization of system neutronics
"from the DD neutron branch# should be further exam!
ined, together with more detailed design of the IEF ves!
sel and wall cooling structures and systems. Finally, ef!
fort should be given to design of careful and detailed
experiments that can test the critical physics issues of
IEF systems, and plans for such experiments be laid out.
The next step is their actual conduct, which can give
unequivocal proof of feasibility of the IEF approach.
6
2. Statement of
Work
The initial Statement of Work for this EPRI study pro!
gram is given in Appendix A, herewith. This outlines the
areas in which work was to be conducted, broken down
into four main sections:
"1# Systems Studies
directly produced "DD!1/2!cat#, to 14.63 MeV for addi!
tional burn of the 3He resulting from decay of the trit!
ium produced. Capture of the neutrons from the neu!
tron branch of the associated DD reactions could also
add energy to the plant system, through their use to
cause ssion in a 10B!loaded blanket; a non!hazardous
application not practically available to DT fusion sys!
tems "which must use their neutrons for tritium!
breeding#. DD!cycles thus grew in interest and impor!
tance as the study progressed, and were concluded to be
the most promising of all available fuel combinations for
earliest practical use in fusion power systems.
"2# Physics Studies
"3# Engineering Studies
B. Physics Studies
"4# Plant/System Development Plan
This area covered particle and plasma physics studies
made by both analytical and computer!based means,
which used existing computer codes with upgrade
modications to determine potential and density distri!
butions in IEF geometries, as a function of drive and
operating conditions. The analytical work was planned
to include studies of ion core convergence, electron
losses, high!Z ion generation and retention, ion/grid
collisions, wall sputtering, bremsstrahlung losses, fusion
product “ash” buildup, multiple ionization e$ects, con!
trol physics, particle ow stability, et al.
A. System Studies
This covered analysis of potential IEF power systems
with emphasis on technology tradeo$s and R&D re!
quirements for economic viability. Power systems were
to be chosen that t EPRI criteria for use and utility in
central station power generation, and modeling codes
developed for both direct electric and thermal conver!
sion power system technologies, for several candidate
fusion fuels. Special e$ort was to be given to aneutronic
fuel systems. Optimal power systems were to be selected
and cost estimates made for full plant systems in com!
mercial power generation use. And e$ort was planned
on study of alternate uses of IEF systems for non!power
applications to desalination, nuclear fuel production,
hybrid "ssion!fusion# systems, ssion!product waste
transmutation, low!cost ethanol production, et al.
Early in the program, preliminary results and review led
to the redirection of e$ort to limit applications studies
solely to central station electric power systems, and to
focus the work on DD fusion fuels, rather than the con!
ventional DT and/or other alternate combinations. This
simplied the work, and provided a clear route for de!
tailed analyses. Subsequently only secondary attention
was given to other fuels, and this was limited to D3He
and p11B combinations. It was early noted that the “con!
ventional” 50:50 D3He combination represented only
one mixture ratio of an innitely!variable of D/3He fuel
mix, and that this would arise in any DD cycle that cata!
lytically burned its own produced 3He, whether from
direct 3He production or from decay of the tritium pro!
duced. Thus the DD cycle fuel system became the “DD!
cat” "for “3He!catalyzed”# systems with variable 3He in!
put.
The result of these observations was that the average
energy of one net DD reaction could be made to vary
from the low value of 3.65 MeV arising from the basic
DD fusions alone, to 10.24 MeV for burn of the 3He
Subsequent program direction reduced emphasis on
high!Z fuels, and increased e$ort on ion/grid collisional
losses and mechanisms, wall sputtering and impurity
buildup, and gross system stability. This focus led to a
deeper understanding of results of the early work by
Hirsch2 on IEF/IXL systems and to limits on this and
on the EXL approach. These, in turn, allowed the de!
velopment of physics “Decision Trees” for both types of
system, which showed the critical issues that must be
addressed to determine true concept feasibility and es!
tablish proof!of!principle experimental criteria.
Analytical work on the issues of concern all showed that
these were tractable and, in general, far less problematic
than the critical physics issues for M&M machines. In
addition, new computer studies were nally able to show
conditions that logically obtained in early IXL experi!
ments, with a good match to the multiple well neutron
outputs observed in these tests. Further analyses showed
that the anomalous increased ion density in the central
wells of these experiments could be described by either
classical collisional physical phenomena or electrostatic
wave!group!trapping e$ects.
C. Engineering Studies
This area was concerned with the integration of the
relevant physics phenomenological models and numeri!
cal computer code results into an existing basic systems
performance modeling code "the PBAL code#, capable
7
of use in studying their e!ects on overall plant system
performance. It was planned to use early performance
modeling to highlight critical engineering issues, and to
develop models for these to be used as constraint algo"
rithms in the code, itself. By this means, the code could
be made to provide tradeo! studies of the relationship
of engineering technologies to system output.
The engineering issues of concern included thermal and
particle loads on walls and structures, stored energy in B
elds and plasma motion, neutron ux, hazards and ma"
terials damage e!ects #for neutron"producing reactions$,
system fueling and control margins, and general hazards
arising from X"rays, high"voltage equipment arcing and
breakdown, et al.
The work accomplished under this task area expanded
the PBAL power balance code to cover all of the impor"
tant parameters of the IEF systems studied. This code
was then used to determine performance ranges over a
very wide range of parameters, for many di!ering fuel
combinations. Emphasis was given to DD and D3He
combinations, but many cases were also run for p11B,
DT, and such exotic fuel cycles as 3He3He and 6Li6Li"#p,
3He$catalyzed. Fusion reaction cross"sections were mod"
eled for multiply"ionized #Z > 1$ fuels in the center"of"
potential #CP$ frame in these machines, and it was
found that the high CP energies reached by fully"ionized
Z > 1 fuels always gave greater performance improve"
ment than for singly"ionized fuels. This possibility is
available only in IEF systems; it can not be employed in
normal M&M fusion machine concepts.
In addition to engineering studies on the issues listed
above, other work included analysis of stored tritium
radiation hazards, in"core secondary reactions of fusion
products with the dense core ions #found to be negligi"
ble$, on conventional thermal stress and heat transfer
limits on system design and sizing, and on detailed con"
siderations of fuels selection, and of fueling in IEF ma"
chines. Complete power plant diagrams were developed,
including all elements of the gas supply and exhaust
handling systems. A baseline DD system design was laid
out at the conceptual level, and the thermal power cycle
for this baseline system was determined. Blankets were
designed for neutron capture in 10B ssion, with associ"
ated increased power generation. Applying one DD IEF
system #a “Fusion Power Core” " FPC$ to each blanket
cell system allowed a complete power plant to be gener"
ated from a multiplicity of FPC/blanket cells. This type
of arrangement, which gives easy replacement and main"
tenance of the FPC units, is also not possible for con"
ventional large"scale M&M tokamak plant concepts.
In general, engineering study work covered a larger span
of more detailed concern for DD systems than originally
contemplated, but less work was done on the purely
electrical equipment side of complete plant systems. In
every case, the IEF systems showed favorable perform"
ance, operations, or maintenance features comparable to
or better than those of nuclear ssion plants, and always
better than those of the large tokamak systems.
D. Plant/System Development Plan
This covered work planned to devise and detail research
and development #R&D$ plans for the attainment of
IEF power systems. It was to be based on the denition
of critical issues #both engineering and physics$ deter"
mined from the work of the other areas, above. Initial
plans also called for the development of deployment
schedules for commercialization, and cost estimates of
both R&D and deployment phases for three levels of
optimism/pessimism in costing and scheduling. In the
early phases of this e!ort this area was redirected to
focus on the next R&D steps that might logically be
required to prove out all of the critical physics required
to give high condence in the IEF concepts of interest,
with less #or no$ e!ort on long"range deployment or cost
estimation of such commercialization.
The critical physics issues were developed, as described
above, and straightforward R&D plans laid out for their
testing. Interestingly, it was found that these issues are
all “classical” physics questions about the natur! of ma"
chine operation % i.e. will a particular classical physics
e!ect appear as theory and numerical calculations pre"
dict % no" about whether or not an empirical relation"
ship #e.g. transport scaling$ will still hold at larger system
scale % as is the case for the conventional M&M ap"
proach. Further, it was found that two of the critical
issues #and there are only three$ can be tested at very
small scale #i.e. systems of a few tens of cm radius$, with
non"fusion"reactive gases, while the full"scale proof tests
at fusion conditions need be only 2~3 m in #spherical$
radius. There is no scaling e!ect in these devices.
This allows the cost of R&D to be estimated rather well,
since there is no open"ended size or power level that
must be pursued to nd “the answer” to any piece of
physics. Because of this, the next step physics tests can
be well"predicted, sized and costed, and this was done in
this task area. In addition, although it was not required,
plans and schedules for commercial deployment and
upgrading were developed, for implementation of IEF
power systems into the global economy over the long
time scales consistent with historical transitions to
wholly new energy sources #e.g. wood/coal, coal/oil, oil/
nuclear ssion$.
E. Deliverables
This covered the deliverable items required for this
study program. It included two"page monthly letter re"
ports of progress, a Final Report at the end of the pro"
gram #which is this document$, and two major program
8
reviews at specied intervals. These two reviews took
place at Redwood City, CA in July 1992, and in a major
review and discussion held in Washington, D.C. in mid!
October 1992. The summary and review information
provided to EPRI for this latter meeting was also a draft
outline of the information to be covered in this Final
Report. The report of the Review Committee for this
meeting is appended herewith in Appendix A.
In general, all of the study work required under the
Statement of Work, as modied and redirected, was ac!
complished about on the schedule originally contem!
plated. Remaining work is that removed by direction
and that determined by the study e"ort itself as desir!
able for further investigation and analysis.
IEF in Electric Power Plants#
9
3. Introduction
A. Origins and History
The rst work relevant to the idea of IEF electric eld
control of charged particles was that done in the 1920’s
by Irving Langmuir and Katherine Blodgett, who wrote
a series of papers on recirculating electron and ion ows
in planar, cylindrical and spherical geometries. These
showed that a series of virtual electrodes could be
formed by the recirculating currents !in the absence of
particle angular momentum" and gave conditions for
their formation.4 Figure 1 shows the beginning of their
paper on spherical systems.
Physics Review, volume 23, pp. 49-59, 1924
Figure 1 — Early Langmuir-Blodgett paper on virtual electrode formation in spherically recirculating charged particle
ows
The second documented ideas of interest here were
those of W. W. Salisbury, who patented a cylindrically#
convergent ow scheme for the generation of fusion
neutrons, in a patent led in 1948.5 This was followed by
a paper by Elmore, Tuck and Watson at Los Alamos, in
1959, concerning spherical systems.6 This work analyzed
the spherically convergent/ divergent ow of electrons
injected radially to form a deep negative potential well,
with ions “dropped” into this well owing radially to a
central convergence core. The electrons were acceler#
ated into the spherical chamber by means of a spherical
screen grid biased positively relative to the !assumed"
electron#emitting shell that formed the boundary of the
system. Ions had to be put into the system inside this
screen grid.
They showed that the power balance for fusion was very
unfavorable for the radially Maxwellian electrons that
they assumed, and that the plasma ow system would be
unstable against pinch type instabilities at high core
densities. In this system an immutable di$culty lay in
the nite solidity of the screen grid, which inevitably
swept out electrons as they recirculated through it, thus
losing power on each pass. Later work by others showed
that the Maxwellian assumption would be incorrect, but
the grid collisional loss problem remained.
The next step forward was in two seminal patents led
in the late 1950’s and early 1960’s by Philo T. Farnsworth,
who proposed to create fusion reactions in ion#
converged cores by injection of ions, directly through
ion accelerating grids, rather than by use of electron#
injected potential elds for ion acceleration.1,7 Figure 2
shows the rst page of Farnsworth’s rst patent on this
idea. The virtue of this approach lay in the fact of the
ions’ greater mass giving much more “sti%ness” to the
particle “beams” that constituted the radially convergent
ows. The defect of this system remained that of the
nite solidity of the grids, that would here intercept and
remove ions from the recirculating ow system.
U.S. Patent 3,258,402
Figure 2 — First major patent by Philo T. Farnsworth on
inertial-electrostatic-connement
It was hypothesized that injected ions would quickly
form a virtual anode !or series of such anodes" inter#
leaved with virtual cathodes due to electron ows made
convergent by the potential elds set up by the ion mo#
tion and particle distributions. For this to occur, it was
necessary to provide a source of electrons to follow the
ions, to provide the necessary quasi#neutrality in the
gross system. Experiments on this concept were carried
out in the early#middle 1960’s, and reported by Hirsch,
in a now#classical paper summarizing results obtained
from work with DD and DT fusion fuel mixtures.3 Fig#
ure 3 shows the rst page of this paper. This work
showed fusion reaction output as high as 1010 fusions/
second in a system operated at 100 keV and 60 mA,
with DT. The actual system tested utilized 6 opposing
ion guns on opposite sides of a spherical vacuum cham#
ber !i.e. cubic geometry" in which electrons were pro#
vided by thermal emission from an internal screen grid
located at the potential surface of the ion gun accelera#
tion grids. This had the e%ect of simulating a spherical
surface emission system by the time the ions had
reached the convergent core region. As in the Farn#
sworth patent, the ions had to recirculate through the
10
screen grid !or their gun accelerator grids" on each pass
through the system.
Journal of Applied Physics, October 1967
Figure 3 — Robert L. Hirsch!s classic paper on experiments
on fusion generation in lEC devices
Hirsch’s data showed fusion rates far in excess of those
to be expected from simple colliding#beam theory,
which would predict output lower by a factor of 1E4 to
1E5 than that observed. The explanation suggested was
that ions could be trapped inside of interior potential
wells formed by the successive virtual electrode struc#
tures shown by Langmuir and Blodgett, and by Farn#
sworth, with a density enhancement su$cient to yield
the observed fusion rate !fusion rate varies as the square
of the reactant density". However, no account was taken
of the limitation on this well structure by transverse
!angular" momentum in the particles, which later work
showed would prevent such buildup. Thus, these intrigu#
ing experiments o%ered output far above any simple
model or theory, and no adequate theory seemed avail#
able to allow prediction or explanation of their results.
Without this core ion densication, the di$culty of par#
ticle losses to the screen grids remained as an insuper#
able obstacle to net power production.
that Hirsch’s data were in error by a factor of 20x. This
improbable set of conditions renders this explanation
quite suspect.
Analytical work conducted under this program, with the
joint support of the DoE, O$ce of Basic Sciences
!through the Los Alamos National Laboratory", was fo#
cussed on developing a better understanding of these
experiments than had heretofore been obtained. This
resulted in an explanation of Hirsch’s results, on the ba#
sis of simple physical models involving synergistic e%ects
in both the electron and ion recirculation through the
IXL gridded system. A relatively complete theory was
developed,9,10 that showed the multiple#peaked features
of the neutron emission data, and could account for the
anomalously large neutron !fusion" output as a logical
consequence of either internal classical collisional trap#
ping within hollow central cores formed by dual virtual
electrodes, or by core density enhancement by wave#
group#trapping of ions within intense electrostatic wave
elds around the core region, set up by ion#acoustic
electrostatic standing waves generated in the ion/
electron counter#ow in this part of the system.
An alternate means of producing ion#accelerating poten#
tial wells was proposed by Elmore, Tuck and Watson6 in
1959, in their paper on electron#driven formation of
negative potential wells. This work assumed radially#
Maxwellian energy distributions of electrons, and util#
ized concentric spherical grids to accelerate these to#
wards the system center to form the desired well. Thus,
as in the case of Farnsworth/Hirsch, the driving particles
!here the electrons" had to pass through a grid screen on
each recirculation through the system. For the condi#
tions assumed, it was found that net power could not be
achieved from fusions among ions trapped in such a sys#
tem, and the power losses due to grid collisions were
excessive. Figure 4 shows the front section of their pa#
per.
One attempt to provide a theory consistent with the
observations was made by Baxter and Stuart, who devel#
oped a model8 invoking charge exchange neutralization
of fast ions traversing the system as they emerged from
their ion acceleration guns, being transformed into fast
neutrals which propagated to the core region without
Coulombic repulsion, and there were collision ally ion#
ized to give anomalous high core densities with their
resulting high fusion rates. The di$culty with this ex#
planation was that the model involved two competing
physics phenomena and, in order to work, it had to op#
erate at the mini#max point and simultaneously declare
11
Physics of Fluids, May-June 1959
Figure 4 — Electron injection-driven negative potential well
trapping of ions, in an inversion of the Hirsch/Farnsworth
system, by Elmore, Tuck and Watson.6
This fatal aw, collision of recirculating particles with
their accelerating grid structures, was an inherent char!
acteristic of both the Farnsworth/Hirsch "ion accelera!
tion# and Elmore, Tuck, Watson "electron acceleration#
schemes. In the former, ions collided with the grids at
their peak energy, while energetic electrons comprised
the lost particles in the latter case.
U.S. Patent 4,826,646
Figure 5a — Patent on means of eliminating grids by use of
quasi-spherical magnetic elds to conne energetic electrons injected to make negative potential well for trapping
fusion ions
U.S. Patent 5,160,695
There are only two ways to reach net power in IEC sys!
tems in the face of this fundamental di$culty:
1.
Devise a means to accelerate and contain the
particles that does not embody structure and
associated collisional losses, or;
2.
Find a means to enhance density only in the
central cor! of these devices, by mechanisms
that operate solely within the boundary of the
inner grid.
Both conceptual solutions render grid/structure colli!
sions impossible. Without one or the other of these so!
lutions, there can be no hope of achieving net power in
either system.
Solutions on the lines of both approaches were invented
by Robert W. Bussard, at EMC2. The rst approach so!
lution was invented in 1983/84, and patents issued in
1989,11 and that for the second approach solution was
invented in 1989, with patents issued in 1992.12 Figures
5a,b show the front pages of these patents.
Figure 5b — Patent on means of greatly increasing core ion
density by wave-group-trapping in standing ion-acoustic
electrostatic wave elds in core region, thus removing ions
from grids
The rst method employed to remove grid collisions
was simply to remove the grids. But then, how are the
ions to be accelerated? The answer lay in the application
of magnetic connement concepts to the connement
of energetic electrons, injected through the conning B
eld, to provide a net negative quasi!spherical region of
potential, into which fusion fuel ions could be inserted
at very low energy, as indicated in Figure 6. The key to
success of this method was the determination of that
magnetic eld conguration that can provide both good
connement and greatly reduced particle "electron#
losses from those experienced in conventional “mag!
netic mirror” or “magnetic bottle” type of congura!
tions. The essence of the patented approach is the de!
nition of this eld structure. It was found that polyhedral
magnetic elds whose generating current carrying coils
t a simple geometric prescription, could always provide
such minimum loss connement.
The prescription is: Any magnetic eld coil arrangement
that satises the criterion that the current conductors
12
lie along the edges of a polyhedron in which each vertex
is surrounded by an even number of faces, can be
formed with the conductors arranged so as to always
produce adjacent magnetic poles of alternating sign or
parity. In such an arrangement there are no “line” cusps;
all of the magnetic losses arise only from polar cusps. A
more detailed description of this system and its opera!
tion has been given by Bussard13 and, later, by Krall.14 It
is of some interest to note that none of the conventional
mirror and solenoidal trap concepts meet this prescrip!
tion, as all have an odd number of faces around each ver!
tex. Figure 7 shows this arrangement schematically and
Figure 8 gives an artist's picture of an EXL machine
based on this prescription "the device shown is topo!
logically equivalent to a truncated cube#.
EXL System General Principle of Operation
3
4
Reaction
Products
R
Electric
Potential
Well
R
0
2
Ion
Density
1
1
Deep negative electric potential well
2
Traps positive ion fuels
3
In spherical radial oscillation
4
Until they make fusion reactions
Voltage
-200 keV (typical)
Figure 6 — General principle of operation of EXL system,
showing spherical negative potential well to trap fusion ions
injected at low energy at well boundary into radial recirculation
The second method invented to eliminate or reduce grid
collisions was that of wave!group trapping of ions
!ithi" the dense core region of the device. This was
found to be possible with conventional plasma physics
phenomenology, by the generation of nonlinearly!stable
ion!acoustic electrostatic waves in and around the core
region.
EXL System Approach
Ion Source
1
Electron Gun
3
Injected Electrons
Magnetic
Field Lines
Magnetic
Field Coils
2
1
Trapping well formed by energetic electron injection
2
Cusps of polyhedral magnetic field
3
Ions fall into well and remain until reacted
Figure 7 — EXL system approach; showing electron injection through cusps of quasi-spherical polyhedral magnetic
eld conguration to conne electrons to form negative potential well
If these can be made large enough in amplitude relative
to the ion energy in this region, then the ions will be!
come trapped within the e.s. wave elds, and will scatter
o$ these elds with a scattering mean free path the or!
der of the e.s. wave length. It is easy to show that this
wave length is always very much smaller than the ordi!
nary Coulomb collision mean free path for ion/ion colli!
sions at comparable density and, in fact, can be made
much smaller than the core radius if the basic core sur!
face density is made su%ciently large.
‘There is thus a critical core region density that must be
reached if the desired e.s. waves are to be set up. The
existence of such ion!acoustic waves can provide a
means of forcing the core region ions to remain in and
near the core because of their extremely small scattering
length. Ion passage through the core region then be!
comes one of collisional di$usion, rather than simple
“straight!through” small!angle Coulomb collision. Ions
will then become trapped in the e.s. wave structure and
their density will build up to levels much higher than
without such wave!group!trapping, as suggested in Fig!
ure 9, taken from the patent.
13
EXL System Artist Sketch
U.S. Patent 5,160,695
Vacuum Envelope
Polyhedral
Magnet Coil
System
Electron
e-Guns
(6 places)
Magnet and
e-gun Power
Vacuum
Pumping
Port
Figure 9 — Particle density buildup in central ion
convergence-limited core, under operation with ICC wavegroup-trapping effect, showing collision-diffusion trapping in
core region
U.S. Patent 5,160,695
Supply Lines
Plant unit cell
mounting and
disconnect plug
(note folded lines
- for shielding
Figure 8 — Artist!s sketch; EXL system shown has a truncated cube eld conguration, and is mounted on a unit
base post containing all cooling, electrical and fuel supply
lines
This has the e!ect of collision ally compressing the core
density; it is called the Inertial"Collisional"Compression
#ICC$ e!ect. It is a well"known phenomena from the
eld of collective particle accelerators, and other areas
of plasma physics; it is wave"group trapping and, as ob"
served by Luce,15 is a well"proven phenomenon. Its invo"
cation and application to these IEC fusion concepts is
new, however.
Hirsch’s data seem to suggest that such a mechanism
might have been operable in his IXL"type experiments,
for they all fell just within the boundary at which the
ICC e!ects should rst become operable. Figure 10
shows the relation of Hirsch’s experimental data to the
e!ective ICC operational regime, from the ICC patent.
Figure 10 — Parameter phase plot showing region of operation of ICC effect for various ion energies and recirculation ratios, and relationship to region of Hirsch!s experiments
However, it is more likely that Hirsch’s data can be cor"
rectly explained by classical trapping in a hollow central
potential well, itself resulting from the convergence of
recirculating electrons to a convergence radius smaller
than that of the spherically converging ions in his sys"
tems. Classical collisional trapping can yield increasing
core densities of both ions and electrons, while still
maintaining a net negative “hollow” central potential, if
certain conditions on particle energies and densities are
met.
Analysis of this shows that these conditions were satis"
ed in Hirsch’s experiments. However, it was found that
operation with 10"100x higher recirculating electron
currents would have destroyed the hollow central well
and made it impossible to reach high core densities by
such classical collisional means. Since Hirsch’s experi"
14
ments were at densities too low by a factor of at least
100,000x for signicant fusion power production, it
seems clear that these classical mechanisms can not give
net power in such devices. This leaves the ICC e!ect as
the only hope presently seen to provide a means for ex"
cess core densication in IXL/IEC devices. However,
the e#cacy of this e!ect remains to be proven $or oth"
erwise% by experiment.
B. Basic Approaches and Characteristics
The two basic approaches to inertial"electrostatic"fusion
$IEF% systems are those in which ions are injected
through a grid"supplied potential, as used by Hirsch, and
that in which electrons are injected through a grid"
supplied potential, to make a negative well for ion trap"
ping, as considered by Elmore, Tuck and Watson. As
noted above, the rst of these is called the IXL system,
for Ion"ACCELeration input, and the second is analo"
gously called the EXL system, for Electron ACCELera"
tion drive. Figure 11 shows these two approaches in out"
line diagram.
These machines all have certain inherent features that
are very di!erent from the conventional concepts for
fusion. These latter are all characterized by Maxwellian
plasmas in local thermodynamic equilibrium $LTE%, held
$or attempted so% by strong magnetic elds; they are
M&M/LTE systems. In IEF/IEC devices the spatial par"
ticle density distribution across the machine is constant
in time. However, it is formed of constantly changing
individual particles. Thus fusion ions are not really
“conned” in the core of the device, but only within the
machine outer boundary. They just pass through the core,
at a rate su#cient to maintain a constant core density
distribution. This reacting core region is inherently non"
neutral and very far out of equilibrium in respect to the
energies of local core region ions and electrons. Figure 12
summarizes these kinetic features of IEF concepts.
IXL vs. EXL Acceleration
IXL - Ion Acceleration
EXL - Electron Acceleration
Magnet
Coils
Grids
Ion
Gun
Injection
at High Ei
Cusp B
Fields
Ion
Gun
Electron
Gun
Injection
at Low Ei
Injection
at High E0
Q
0
Ion
Accelerating
Well
E0
E0
Ion/Grid Collision Losses
Cusp Electron Losses
Figure 11 — Two basic approaches to inertial-electrostaticfusion (IEF); EXL and IXL systems
IEF - Inherent Characteristics - I
1. Local region particle density is constant in steady
state machine
•
But formed of continuously changing population
members
2. Thus ions are not “conned” in core
•
They just pass through it
3. Reacting “core” region is:
•
Non"neutral and highly non"equilibrium
Figure 12 — Inherent thermodynamic and kinetic features
of IEF systems
Collisional upscattering of both ions and electrons can
a!ect only those particles that are within the core, while
the fusion rate depends only on the net particle ion den"
sity $independent of which ion is being considered%.
Thus, since both ions and electrons “live” in the core
only for a very small fraction of their lifetime in the ma"
chine, the ratio of fusion rate to upscattering rate is
heavily biased in favor of fusion, by the ratio of total
particle lifetime to dwell time in the core. This can be
the order of 1000x, thus greatly inhibiting Maxwelliani"
zation of either particle.
Other inherent characteristics are favorable to system
power production. These include the fact that the core
is always su#ciently un"dense that fusion products never
$i.e. < 10"9% deposit their energy within the core plasma,
and do not react with core ions. Their energy is carried
out of the potential well region of the device, to impinge
on external structures or otherwise be handled. One
such means, inherent to IEF devices, is to direct"convert
the kinetic energy of the charged fusion product ions, by
causing them to leave the machine through a concentric
radial grid system, biased electrically so as to slow down
their motion, and thus convert it into direct electrical
power. And, of course, since fusion product energy is no!
deposited in the core ions $or electrons%, and since the
core reactions are not thermonuclear, anyway, these ma"
15
chines never can ignite, as in M&M systems. They will
always be power ampliers. These features are summa!
rized in Figure 13.
IEF - Inherent Characteristics - II
4. Fusion products do not deposit their energy in
the “plasma”
•
•
But into structures exterior to the “core”
They do not react with core ions
5. Thus IEF machines never ignite as in M & M
machines
•
•
They are electric power ampliers
Pelectric in gives Pfusio! "thermal and/or electric# out
device that exchanges kinetic energy and momentum
between ions and electrons as they move radially in and
out, makes the ion and electron spatial energy distribu!
tion very di$erent from and much more favorable for
IEC than for M&M systems. Figure 16 shows ion colli!
sional scattering rate and fusion rate spatial distributions
for both types of concept. Here, again, the kinetic ex!
change feature of the IEC gives it a strong advantage
over M&M systems.
Comparative Spatial Distribution
M&M
IEC
18
nmx ! 10 /cm3
14
nmx ! 10 /cm3
12
n ! 10 /cm3
Density
12
6. Direct conversion inherently possible in region
outside of reacting “core”
Figure 13 — Inherent IEF power system, fusion product and
power conversion features
Other characteristics are shown in Figures 14, 15, and 16,
that show comparisons of IEC/IEF systems with M&M/
LTE systems concepts. Figure 14 shows the energy dis!
tributions expected in each system for the fusion ions.
Note that the IEC system operates with mono!energetic
ions, chosen to t the design energy for the desired fu!
sion reaction cross section.
M&M vs. IEC Comparison
IEC
M&M
Em , E0
n ! 10 /cm3
Ions
Eo ! 50 keV
Em ! 10 keV
Energy
-4
E ! 10 Eo
E ! Em /4
Figure 15 — Comparative spatial distribution of particle
densities and energy; note strong peaking and non-LTE
distribution in IEC system, and weak variation in M&M system
Ion/Ion Scattering and Fusion Density
IEC
M&M
qf ~ 1012/cm3 sec
E0
qf ~ 1020-1021/cm3 sec
Fusion
Rate
Density
Em/4
[n]
Total
Scattering
Rate
[n]
Em [E]
E0
[E] E0
Figure 14 — Comparison of IEC with M&M systems, showing typical ion paths and energy distribution functions; note
IEC ions all at same energy, but fusions only in M&M tail
In contrast, M&M systems su$er from the very large
loss!generating collisionality of low energy ions, with
fusion power production limited to the small fraction of
the ion distribution found at high energy in the Maxwel!
lian “tail.” Figure 15 shows particle density and energy
distribution across the machine for each device type.
Here note the sharp peak density at the center of the
IEC system, with the comparative slow variation across
the M&M machine. This, plus the fact that the IEC is a
Electrons
Eo ! 50 keV
High Energy
Collisional Losses
Qe ~ 104 Qc
Qc ~ 30 Qm
Qm
Low Energy
Isotropization
Figure 16 — Ion/ion scattering and fusion density distributions; note core peaking and low energy boundary isotropization in IEC, while M&M has high edge scattering collision
losses
In particular, note that the very low ion energy at the
IEC boundary leads to a very high isotropizing collision
rate, that successfully randomizes any anisotropy intro!
duced in radial ion ow in passage through the mantle
or core of the IEC system. This ensures that a small
momentum!limited converged core will remai! small;
transverse momentum will not build up in successive ion
passes through the machine.
16
C. Summary of Current Status
Background source work in this eld is very generally
summarized in Figures 17 and 18, that cite early research
work in the U.S. and briey list major non!U.S. past ef!
forts. Current "1992# work is similarly listed in Figure 19,
that shows the relatively small!scale programs underway
with support from the DOE "Basic Energy Sciences#,
the EPRI, NASA/SDIO, and the US Navy.
•
•
•
•
•
Agency
Program
Time
Frame
DoE
Basic energy sciences,
3 year joint program
"LANL/ EMC2/
University Illinois
1992!1994
IEC ! BACKGROUND OF U.S. WORK
EPRI
1 year program
1992
Sporadic work since 1924 "Langmuir, et. al.#
Key experiments, mid ! 1960’s "Hirsch, Farn!
sworth#
New concept, research program, 1986 ! 1992 "Bus!
sard, et. al.#
Total U.S. e$ort to date ca. %15M ! %20M
Current U.S. e$ort ca. %1.0M / year
"5 sources, 5 contractors#
SDIO/NASA
1 year program
1992!1993
US Navy
6 month SBIR study
1992
Figure 17 — Background of current U.S. work on IEC/IEF
concepts
Each of these is aimed principally at further theory,
computer code development and analytical work; little
support exists at this time for signicant experimental
work. The sole exception to this is a modest subcontract
activity underway at the University of Illinois under
DOE support "through the Los Alamos National Labo!
ratory#. This is studying IXL/IEC devices experimen!
tally at a scale of approximately 15 cm radius, and ion
drive voltages up to 25!40 kV, using concentric wire
grids in Hirsch/Farnsworth!type experiments.
IEC — Non U.S. Work
France
“Pleiades” ! Fontenay
1960’s=1970’s
Quebec
Spindle cusp ! Varenne
mid 1970’s
USSR
“Jupiter” ! Kharkov
1960’s!1980’s "?#
Figure 19 — Current programs and sources of support of
U.S. IEC/IEF work (1992)
This research "and analytical and theoretical research
done in 1989!1991 under DARP A sponsorship# has suc!
ceeded in dispelling three principal objections raised to
IEF/IEC systems by earlier workers. These were:
"1# Inability to produce small tightly converged ion
cores;
"2# Inherent instability of the potential well due to
localized “pinch” e$ects "the “prevalent” Weibel
instability#, and;
"3# Inability to maintain mono!energetic ions over
their system lifetime.
Each of these has been proven inapplicable or of little
signicance in the IEC concepts of interest here. Ion
radial convergence has been shown to be easy to achieve
in either EXL or IXL systems, the “prevalent” instabil!
ity was found not to exist for the actual electron and ion
distribution functions found within these systems, and
collisional thermalization and upscatter was shown to be
insignicant vis a vis fusion times over particle lifetimes
"except for ICC e$ect dense cores#.
Figure 18 — Earlier non-U.S. work on IEC/IEF concepts
To date, neutron production rates from DD in such sys!
tems has not exceeded about 1E6 neutrons/second, well
below those obtained by Hirsch in the mid!1960’s.
Other work done by EMC2 under the LANL/DOE and
the support "the NASA/SDIO and US Navy has im!
proved parametric computer codes for phenomemologi!
cal scaling studies, and basic theory of system stability
and electrostatic wave formation.
17
Status of Technical Issues I
Technical
Issue
Core region
stability
What Has Been Done
Vlasov theory for specied edge
densities
Ion conver!
gence stability
Edge core mantle collisional mod!
eling
Maxwellianiza!
tion
Collisional energy transfer analyses
Particle and
potential dis!
tributions
First order grid code modeling
Potential well
stability
Analytical modeling
Figure 20a — Status of Technical Issues: Potential and particle distributions
The current status of technical issues is as summarized
in the topical listings given in Figures 20"a!f# "Status of
Technical Issues ! I!VI#. These list each major technical
topic area in which signicant work has been done over
the past 6 years. The nature of the work accomplished is
suggested in the brief descriptor given for each topical
area.
Status of Technical Issues II
Technical
Issue
What Has Been Done
Electrostatic
wave generation
Vlasov Theory and multi!
streaming analyses
Particle / Wave
trapping
Collisional and energy modeling
Diamagnetic elec!
tron ow
PIC simulations and phenome!
nological models
Cusp electron
loses
First order models and numeri!
cal simulations
Ion/electron grid
interactions
collisional models, secondary
emission analyses
Figure 20b — Status of technical issues: Electrostatic
waves, trapping, and particle losses
Although most of the basic issues were developed under
prior and companion work, work under this program has
been of great value in the assembly of all of this, and for
its extension to consideration of specic issues of inter!
est and concern to utilities power plant applications of
IEF/IEC systems. The listings in Figures 20"a!f# are self!
explanatory, and are not repeated here. Further details
are given later in this report.
Status of Technical Issues III
Technical
Issue
What Has Been Done
Plasma system
performance
Parametric studies using phe!
nomenological models and Vlasov!
Maxwell computer codes
Bremsstrahlung
radiation
Collisional energy transfer analy!
ses
First wall stabil!
ity
Heating/cooling loads, thermal
stress calculations
Materials
damage/
activation
Fuel choice, neutron spectrum,
ux/uence analyses
Figure 20c — Plasma performance, power losses, structure
damage
IEF in Electric Power Plants$
18
4. Plasma Physics
Technical Features
A. Particle and Potential Distributions
Ion and electron distributions and potential distribution
in either type of IEC system can be calculated by use of
a well!proven Vlasov!Maxwell code that solves Poisson’s
equation for a one!dimensional system, with conserva!
tion of transverse momentum for each particle species,
given boundary conditions for particle density, energy
and transverse energy spread. It was early found that
this code "and all other equivalent approaches to nu!
merical computation of the subject problem# could not
be used ab initio to determine distributions above some
rather small average density limit. This was because
these machines are all very nearly charge!neutral, and
the higher the total charge density, the smaller became
the fractional deviation from charge neutrality required
to maintain a potential well of any given depth. Above
the limiting density it was simply not practically!
possible to “guess” the boundary conditions at which the
equations could be satised.
Status of Technical Issues IV
could be calculated without code failure, and corre!
sponding recirculating electron currents up to 1E12 Amps
could be handled, as well. Once the code run has
reached the desired or specied core density "or other
stopping parameter# its output is given as the radial dis!
tribution of density of ions and electrons, and of the
electrostatic potential that successfully accommodates
these distributions.
Status of Technical Issues V
Technical
Issue
What Has Been Done
Power system
performance
Steam cycle and direct conversion
system design studies
Magnet heating
Design studies, ohmic losses, neu!
tron heating superconductor coils
Sputtering
Ion/grid and wall collisions, high Z
migration and removal
Ash buildup
Collisional trapping and up!
scattering losses
Figure 20e — Power system performance and exhaust production
Status of Technical Issues VI
Technical
Issue
What Has Been Done
Technical
Issue
What Has Been Done
Hazard poten!
tials
Stored energy in B elds and
plasmas, radioisotope inventory,
neutron power
Direct conver!
sion
Conceptual studies, geometry
Power engineer!
ing constraints
Thermal/hydraulics/stress analy!
ses; limits of conventional materi!
als
Direct conver!
sion electrical
systems
Power/mass scaling of inverters/
converter, rectiers, high voltage
transformers
Systems con!
trols
Zero order analyses, plasma and
power engineering parameter
space
Power system
packaging
System layout and optimizations,
102 ! 104 MWe
Power system
dynamics
Phenomenological models, startup
and operating state stability
IEF power ap!
plications
System concepts/applications
studies; electricity, process steam,
neutrons
Figure 20d — Status of technical issues: Hazards, engineering limits, dynamics and control
Figure 20f — Electrical conversion, production and power
applications
To avoid this fundamental problem, this code has been
developed as a “time!dependent” analysis tool, by link!
ing successive static solutions from one density/time
regime to the next, “inching up” the density at each such
“time!cycle.” This “adiabatic” procedure was developed
to such a degree that densities in the core up to 1024/cm3
The computation is set up to treat all ions as though
they were deuterons "D#, but other ions "e.g. T, H, 3He,
etc# can be distributed from this basic data, by simple
scaling adjusted for their charge and mass. With this
output, the code then computes the local fusion rate
that would obtain for the given distributions, and can
19
display fusion rate density spatial distributions and total
fusion rates for any fuel of interest. Figures 21, 22 and 23
summarize this code, called the EIXL code. Further
information on its nature and construction has been
provided in papers given at recent meetings of the
Plasma Physics Division of the American Physical
Society.16 ,17 One of these is included here in Appendix B.
EIXL Code for Various Fuels
R
Ggain !
Pfus
Pinj
!
E fus + bij ni2 (r ), ( E )vi (r )4* r 2 dr
0
Peinj ) Piinj
Code computes Ggai!, reactions, and density for DD
fuel
Ion, Electron and Potential Distributions
Determined by the EIXL computer code:
DT, D3He, p11B modeled at 50:50 equal density mix!
tures
•
DT = Total ND + NT = Total ND "EIXL#
Figure 21 — EIXL code general features; a Poisson-solver
for charge and potential distribution
Two examples of distributions obtained by the EIXL
code are shown in Figures 24 and 25. Figure 24 shows an
EXL/IEC system of radius 100 cm driven by electrons at
100 keV, ions injected at 100 eV, with transverse ener!
gies of 1 keV and 0.1 eV, respectively, and recirculating
currents of 105 and 104 Amps, respectively. Note that the
central potential has risen from its minimum well “bot!
tom” by about 30 keV, to !80 keV, and that the central
core ion and electron densities are both about 0.8E25/m3
"8 x 1021/cm3#. The DD fusion rate in this device would
be 3.3 x 1018/sec; DT would be about two orders of mag!
nitude higher.
Time Evolution of Steady-State Solutions
Electronns:
dned # 3ninj ned & 1
!%
"
(
dt
%$ Fn Ft G j (' ttrans
# 3ninj
# 1
dn
1 & DD fus & 1
Ions: ed ! %
" ned % ) ( "
(
dt
%$ G j Fut (' Fn FDD (' ttrans
%$ Fn Ft
r / vinj
tup
N
( 4 / 3)* r 3
Ft !
Fu !
where Fn !
FDD !
ned
ttrans
ttrans
ttrans
Figure 22 — Synthetic “time-dependent” adiabatic operation
of EIXL code to high densities
EXL DD System
0
26
-20
24
-40
22
-60
20
-80
18
-100
16
-120 -4
10
10-3
10-2
Radius (m)
10-1
1
Log of Particle Density (1/m3)
dned
dN
derived from global
! Input " Losses
dt
dt
Figure 23 — EIXL code fusion power and gain output
schemata for various fuels
14
Figure 24 — EXL DD system; ion, electron and potential
distributions at conditions given
DD EXL System at Low Energy
0
27
-5
25
-10
23
-15
21
-20
19
-25
17
-30
10-4
10-3
Radius (m)
10-2
10-1
15
Series of intermediate steady!state solutions linked
by “time!dependent” change in edge density, "d!ed/
d"#
p11B = Total ND + N11B = "1/3# Total ND "EIXL#
Potential (keV)
•
D3He = Total ND + N3H# = "2/3# Total ND "EIXL#
Potential (keV)
•
Solves Vlasov!Poisson equation in spherical ge!
ometry
Conserves transverse momentum of ions and
electrons
Collision!less 1!D boundary value problem
Figure 25 — DD EXL system at very low energy, high ion
current; at conditions given in text
The system shown in Figure 25 is very di$erent from
this. It is one in which the electron injection energy is
only 27.0 ±3.0 eV, with a ±10.0 eV spread, driving a well
of radius 30 em, with an injection current of about 600
Amps, increased by the recirculation provided by trap!
ping in a B eld of 1500 Gauss. The ion input is at 7500
Amps and 1.0 ±0.1 eV radial energy, with a transverse
spread "i.e. a “temperature”# assumed for computational
20
purposes to be only ±2.7 x 10!3 eV. This system shows a
central core density of order 8 x 1025/m3 "8 x 1022/cm3#.
This system could not produce fusion, but is illustrative
of those that might be of interest for atomic "rather
than nuclear# interactions for chemical processing or
light generation by central core collisional ionization
and subsequent recombination of atomic species.
B. Virtual Anodes, Core Convergence,
and Multiple Wells
The previous gures showed the rise of central virtual
anodes. This is a natural consequence of the fact that
the ions always outrun the electrons in their path
through the central potential well to the core region,
simply because the ions carry vastly greater momentum
than do the electrons, in either version of the IEC ma!
chine. In IXL the ions are injected at high energy, and
drag the electrons after them towards the core, while in
EXL the ions reach high energy by falling down the
electron!driven well, then dragging the now!less!
energetic electrons along to the core. It is clear that the
degree to which the central anode will rise will depend
on the relative levels of the electron and ion currents
recirculating through the machine.
Extensive studies of this have been made using the
EIXL code; one set of which are shown in Figure 26.
This shows the change in height of the central. virtual
anode over a range of ion currents in a system with xed
electron drive parameters at 22.5±2.25 keV injection en!
ergy with a transverse energy spread fraction of about
0.3. The core density in this 100 cm system was the or!
der of 2!4E16/m3, and the ion current varied from 0.66
to 4.2 Amps, with the result that the central anode po!
tential went from near!well!bottom to well “blowout.”
Thus, for this range of conditions, a variation of a factor
of about 6!7x in ion current "or in the ratio of electron
to ion current# would lead to destruction of the poten!
tial well. Other studies show that variation in ion/
electron current ratio of 2!4 is acceptable for nearly all
conditions of interest for fusion power production. This
result has useful consequences for system control, since
it means that control systems need not provide exact
control of injection currents to avoid well blowout;
rather a very wide control band is quite acceptable for
system well stability.
21
12
Potential (keV)
0
10-2
10-1
Radius (m)
1
Ia = 1.68 Amp
10
20
-5
18
-10
16
-15
14
-20
12
-25 -3
10
10
0
10-2
10-1
Radius (m)
1
Ia = 1.97 Amp
20
-5
18
-10
16
-15
14
-20
12
-25 -3
10
10
10-2
10-1
Radius (m)
1
-10
16
-15
14
-20
12
-25 -3
10
10-2
10-1
Radius (m)
1
Ia = 2.98 Amp
0
10
20
-5
18
-10
16
-15
14
-20
12
-25 -3
10
10-2
10-1
Radius (m)
1
Ia = 4.27 Amp
0
10
20
-5
18
-10
16
-15
14
-20
12
-25 -3
10
10-2
10-1
Radius (m)
1
10
Log of Particle Density
(1/m3)
-20
18
(1/m3)
14
20
-5
(1/m3)
-15
Potential (keV)
16
Potential (keV)
-10
Ia = 2.27 Amp
0
Potential (keV)
18
(1/m3)
-5
-25 -3
10
Potential (keV)
20
(1/m3)
Potential (keV)
0
(1/m3)
Variations of Central Virtual Anode Height
Ia = 0.663 Amp
Figure 26 — Variation of central virtual anode height with ion current for xed EXL electron injection current and energy drive
conditions; note anode rise with ion current increase
The provision of initial core convergence is also a non!
problem. In IXL systems the limiting issue for conver!
gence of the ions recirculating back!and!forth through
the inner "accelerating# grid is the deection introduced
by local grid distortion of the E!eld, and by collisions
with the grid wires "or cooled conductors# themselves.
In addition, assembly errors or fabrication tolerance
errors can contribute to ion deection from strict radial
paths, if the grids are o$!center, for example, or are lop!
sided or ellipsoidal rather than spherical. However, a
considerable latitude is allowed in these deection
sources. If a core convergence radius ratio of <rc> = rc/R =
0.0033 is desired, for example, a 2.0 m sphere could tol!
erate errors up to ±0.5 cm around the mean. As indicated
in Figure 27, this level of error had already been achieved
by a graduate student, over a decade ago, using a rela!
tively crude wire grid system. It is expected that IXL
convergence ratios up to 1E!3 can be obtained for sys!
tems in the range of R > 2.0 m.
IEF in Electric Power Plants'
Core Convergence - IXL
•
Limited in IXL by grid precision and placement
<rc> = %x/R; %x = Error, R = grid radius
for R = 2.0 m, <rc> = 0.0033, %x = 0.5 cm
Note: B. Edwards !1979, University of I"inois# achieved
$x = 0.5 cm on 0.25 m syste%
Figure 27 — Core ion convergence in IXL systems
For EXL systems, the situation is both more and less
complex. Less complex because there are no grids or
wire!distorted elds to drive the ions from their ap!
pointed paths, but more complex because the surface
polyhedral elds are not spherical and place V x B de!
ection forces on ions as they rise from the well center
towards its edge. In an EXL system the ions must be
introduced slightly inside the system B eld boundary, in
order to avoid just these V x B forces in their initial path
formation. If injected so as to make a small core, limited
only by their transverse energy, they can & as in IXL &
reach core convergence ratios of 10!3, or smaller, depend!
ing on exact system design conditions. However, un!
22
Core Convergence - EXL
•
Limited in EXL by mechanical precision and
placement of B eld coils
Tolerances similar to IXL grids
Added limit set by ion transverse deection in
cusps of surface eld
•
•
rc !
dE3
E0
dE3 - dE3 0
; /
Average over all ions
E0
. E0 21
4
AVG
Ecusp ( Adiabatic)
G j E0
10 "2
rc 4
; if G j ! 10 4 ;
Gj
; Ecusp 4 10 "2 E0 , thu
us
rc 4 10 "3
Figure 28 — Core ion convergence in EXL systems
In either system a situation can arise in which the elec!
trons may converge to a radius that is smaller than the
ion convergence radius. In this situation, the central
portion of the potential well will become net negatively
charged, and the well will become “hollow.” In such a
hollow central potential it is possible to trap both ions
and electrons by classical collisional mechanisms involv!
ing interactions with background gas. Since background
gas is inevitably present in most experiments, and is ac!
tively desired in others, this sort of behavior can be used
as a powerful means of increasing central core densities
above those from simple near!1/r2 geometric conver!
gence. Indeed, analysis shows that the anomalously large
core densities found in Hirsch’s experiments "deduced
from anomalously large neutron output# could well have
been a result of such mechanisms. To test the ability of
properly constrained electrons and ions to produce hol!
low wells, a series of EIXL computations was made, at
conditions found in the experiments in question. Figure
29 shows such a hollow well at the size, current, and par!
ticle energy conditions that characterized Hirsch’s ex!
periments.
0
20
-20
18
Electrons
-40
16
-60
14
12
-80
Ions
-100
10-4
10-3
Radius (m)
10-2
Hollow Central Potential in IXL System
Potential (keV)
avoidable V x B scattering of these ions will inevitably
occur as they move out and are captured in adiabatic
motion by the B eld cusps, before they have reached
their full extent of possible radial travel against the
electron!driven E!eld. Study of this issue suggests that
the adiabatic capture energy may decrease as captured
ions mirror!reected from positions deeper into the
magnetic cusps. As reected, they will reenter the sys!
tem "leaving the adiabatic condition# with random angu!
lar distributions, thus assuming the adiabatic capture
energy as their transverse energy dE!. The magnitude of
this energy will itself, be a function of the degree to
which electron diamagnetic currents have pushed out
the surface B eld, thus increasing the adiabatic capture
radius for ions and reducing their capture energy. This is
found to give an approximate relationship between ion
adiabatic capture energy and ion well depth that de!
pends on the electron recirculation ratio Gj. This is be!
cause Gj is a measure of the diamagnetic e$ect. The
formula is dE!. = Ec/Gj so that the convergence radius
ratio becomes <rc> % 0.1/"Gj#0.5. Since Gj >> 104 is re!
quired for net fusion power in any of these systems, it
appears that the convergence ratio can always be made
<rc> < 10!3. Figure 28 summarizes this situation for EXL
systems.
10
10-1
Figure 29 — Hollow central potential in IXL system, resulting from electron convergence to a specie radius less than
that of the ion convergence momentum-limited core
Results of these showed the anticipated hollow wells,
but gave neutron production "fusion# rates that were
typically 100!1000x too low to agree with the data. But
then, EIXL only calculates fusion rate on the basis of
the distributions obtained by satisfying Poisson’s equa!
tion in a collision less system, with xed boundary con!
ditions. However, ions created in the central well can
oscillate therein until upscattered over the edge, while
electrons created in such a central well can remain
trapped even after upscattering. Thus, net negative
charge ca! be preserved as the total density builds up to
values limited only by the balance between kinetic pres!
sure "nE# and local electric eld strength "!2/8&#. Using
the hollow wells obtained from the EIXL code, and ion
trapping due to ion background charge exchange, with
electron trapping due to electron background ionization
"which also adds ions to the hollow core#, it was found
that increased core densities up to a factor of 100x or so
could be obtained by these classical collisional mecha!
nisms, without destruction of the central well. This is
enough to explain the very large neutron output.
23
1018
1014
1011
1017
DD
108
DD w/ICC
1016
105
1015
102
1014
10-1
10-4
10-3
Radius (m)
10-2
Ion Density (1/m3)
DD Neutron Rate Density (1/sm3)
ICC-Effect Enhancement
1013
10-1
trons will be scattered so as to become locally isotropic
in velocity distribution. This absolutely prevents the
formation of small electron core convergence, and thus
of hollow central potential well structures. Analysis of
this situation #and results of EIXL code runs that show
this instability$ show that this will occur at total recircu!
lating electron currents above about 100!1000 Amps or
so, equivalent to total recirculating ion currents of 5!50
Amps; approximately 10!100x times higher than those
used in the experiments. Since core densities must be
increased by ca. 1E5x to reach useful fusion power levels,
it is clear that hollow wells will not provide a means to
reach fusion.
e- — Two Stream Instability - I
Figure 30 — ICC-effect enhancement and fusion power
density distribution in IXL system
Alternatively, as discussed in an earlier section of this
report, the possibility exists that core density was en!
hanced by the action of electrostatic waves in particle
trapping by producing a collision!di"usion dominated
central core region, whether!or!not it was hollow. Mod!
els for onset of the ICC e"ect are built into the EIXL
code; analysis of this e"ect shows that its onset condi!
tion requires that ! > 0.4!0.5. With these models/
algorithms it is possible to determine the existence, dis!
tribution and e"ect of ICC wave!group!trapping on
particle density and fusion density distribution, under
the assumption that the ICC e"ect will not materially
change the unenhanced potential distribution. With this
feature of the code in operation, computation of ICC
e"ect enhancement #Gicc$ and resulting total fusion rate
can be made from the basic unenhanced data. This was
done for the example shown in Figure 29, with the re!
sults given in Figure 30. This shows the fusion power
density distribution with the ICC running, and gives the
Gicc enhancement factor and fusion output for DD and
DT fuels. Note from the gure that an output gain of
order 100x has been achieved. Adjustment of drive con!
ditions and more detailed calculation of output could
easily raise this to over 2000x. Thus, either mechanism
could explain the high output in Hirsch’s work.
While these mechanisms may be correct, they may not
be able to operate over every plausible range of system
drive conditions. In fact, it has been found % both by
code calculations and by elementary theory % that the
hollow well model can not be sustained as the recirculat!
ing electron current is increased much over the values
used in the early experiments. This is because counter!
streaming electrons in a plasma of ions will go “two
stream” unstable above some density and energy, as indi!
cated in Figures 31 and 32, taken from the text by
Chen.18 If the electrons “go two!stream unstable” they
will form electrostatic waves of amplitude su&cient to
scatter each other with scattering lengths that are com!
parable to the e.s. wave length. These will always be
small compared to the system dimensions, thus the elec!
F(x,y)
The dispersion relation is
&
#m / M
1
)
1 ! 7 2p %
(
2
57 " kv0 62 ('
%$ 7
7
kv
; y8 0
Let x 8
7p
7p
Then 1 !
1
0
0
y
x
The function F(x,y) in the two-stream instability,
when the plasma is stable.
m/M
1
)
8 F 5 x, y 6
x2
5 x " y 62
F(x,y)
1
0
0
y
x
The function F(x,y) in the two-stream instability,
when the plasma is unstable.
Figure 31 — Two-steam instability in counter-ow system;
map of dispersion relation18
e- — Two Stream Instability - II
Result:
2* vb mi
7 p gives
:
me
9w
-m 0
2.4 x10 4 Ee &
Case 1: Ee ;; Ei / e 2 ; ne ;
(
A9w2
. mi 1
( E in eV; E in eV
i
( e
- me 0
12.63E
Case 2: Ee :: Ei / 2 ; ne ; 2 2 i (
A 9w
. mi 1
('
3
9w in cm; Z ! 1; ne in 1/cm ; A ! mi / m p
2
or for 1: I e ;
3.7 x10 "6 - r 0 3/ 2
/. 9 21 Ee is unstable at 7 p
A
w
2
for 2:: I e ;
3.5 x10 "11 - r 0 3/ 2
/. 9 21 Ei
A
w
Figure 32 — Two-steam electron instability in counter-ow
system; instability onset conditions18
However, the ICC e"ect remains, as it will work with
central anodes that are not hollow, as well as otherwise.
Unfortunately, its existence and e"ectiveness yet remain
to be studied and proven by experiments that cover a
range of current beyond that of the two!stream instabil!
ity regime discussed above.
24
A more pervasive instability has been found that can
a!ect the IXL system; this is a basic global instability of
the entire potential well. Fortunately, it applies most
strongly to fusion fuel ions that have high charge num"
ber, Z > 1. To see this instability, consider a simple model
of the potential well inside the inner grid of an IXL sys"
tem. Into this region ions are injected at high energy,
and electrons are dragged in at very low energy. As the
ions focus towards the central core, they produce a vir"
tual central anode. This positive “hill” attracts the slow
electrons and accelerates them as they rush to the core
region. From the core, both species recirculate out to
the inner grid #electrons$ and beyond it to the accelerat"
ing space between the inner and outer grids # ions$,
thence to return to the inner grid region again.
Now, as the central anode becomes higher and higher,
for xed ion and electron currents, the average time an
electron spends in the region within the inner grid will
become less and less, because it is being accelerated to
higher speeds with each pass. Conversely, the ions are
slowed down by a central anode of increasing height,
and spend more time within this region. The result is
that, as the central anode is raised, the inner region be"
comes more positive, thus raising it still more; until the
entire well blows out.
Detailed analysis shows that the criteria for global sta"
bility is that the central anode always be kept so that ! <
1/#Z+"$, where ! is the fractional virtual anode height #!
= #Eanod!/E0$ and " is a parameter combination that is
always << Z. Thus, IXL well stability can be preserved
for all fuels with Z = 1, but not necessarily for fuels with
Z > 1.
D. System Constraints from Physics
Features
As noted above, the ICC e!ect will be initiated if the
central virtual anode can be driven to a fractional height
of ! > 0.4"0.5. Taking this together with the IXL global
well stability limit, above, immediately gives the result
that IXL systems can use the ICC e!ect only with fuel
ions that have Z < 2"2.5. Thus, this would limit the IXL
machine to use with DT, DD, or the DD"cat#3He$ fuel
combinations #i.e. 11B, 6Li, et al are excluded$. This is not
necessarily a fatal weakness, however, since the DD"cat
cycles are probably the most promising fuel cycles for
any reasonably practical fusion power system.
In contrast, the EXL system can % in principle % be
run with almost any combination of fuels, since it is not
subject to the global well instability of the IXL system.
This because increasing anode height hardly a!ects elec"
tron or ion lifetime in well"converged EXL systems.
Their lifetime is determined almost entirely by the time
they spend in the region near the system boundary,
where the electron"driven potential well is changing rap"
idly with radius. This is the region in EXL systems
analogous to that between the inner and outer grids in
the IXL system. But, unlike IXL, the particles in EXL
systems circulate all through the system, to the outer
boundary which is analogous to the IXL outer grid.
Another di!erence favors the EXL system. This is that
EXL can be run will arbitrarily small central virtual an"
odes, while IXL % if ICC is required to provide su&"
cient core density for fusion % can not. This is because
ICC onset demands high central anodes while EXL will
operate with any anode height. Because of this, EXL can
use high Z fuel such as p11B #Z = 5$, which otherwise
could not be used because of excessive bremsstrahlung
losses. IXL machines can not use these fuels, because of
both bremsstrahlung and well stability. EXL is the only
concept that can use completely clean fuels; albeit their
use demands severe drive and eld conditions, and very
small central anodes.
IXL System with Double Grids
0
!Eib
Potential
C. Plasma/Potential
Distribution Stability
0
!Eib > !Ei upscatter
!Eeb > !Ee upscatter
!E0
Ions injected at r = rgi by
low voltage i+ guns (ca. < 50 eV)
!Eeb
rg1a biased -E0 relative to rgi
rg1a rg1b
rg1 r0
Radius
Figure 33 — IXL system with double grids showing bucking
voltages to suppress up-scattering
And, if ICC is used to make IXL workable, it will also
produce Maxwellianized particles because of the very
high density maintained in the particles captured by
collision di!usion in the dense core. This is the density
that produces the high fusion rate, but will also produce
strongly upscattered electrons that give rise to intoler"
able losses into the double"grid space. Prevention of
these losses requires the imposition of bucking voltages
for both ions and electrons through the means of a sec"
ond set of grids, proximate to the rst, as shown in Fig"
ure 33. To be e!ective these bucking voltage must be
very large, typically 8"10x the desired well depth. This,
in turn, drives the ICC enhancement requirement
higher to counter the larger loss per particle experienced
by particles that escape the bucking voltage barrier.
And, this also creates a much larger particle impact
thermal load on the #now dual$ set of grids in the system,
with consequent extremely severe cooling problems.
Finally, the ICC enhancement required is found to ex"
25
ceed that projected as an upper limit on Gicc by a factor
of about 10x. Thus it is not clear that the IXL system
can work at net power, even with the ICC e!ect operat"
ing. Since this is only poorly understood at present, it is
imperative to conduct experiments to test these consid"
erations.
Meanwhile, the EXL system does no! su!er from either
of these di#culties, as it does not need the ICC e!ect
to operate, can operate at any small anode height, and
has no grids to need cooling. Nor does it go well"
unstable at high anodes, and can be operated so as to
preserve mono"energetic particles $i.e. it does not go
Maxwellian%. Because of all this, it can use clean fuels,
and & in fact & any fusion fuel combination yet found.
26
5. IEF Power
Systems
quiring bucking voltages of 500!800 keV; and these are
much more di%cult to sustain.
A. IEF Plasma Connement
Performance
Cesiated with
Face Strip
In the IXL machine the plasma performance is gov!
erned entirely by the balance between losses due to ion
grid collisions and particle kinetic energy losses from
electron upscattering, and the power generation from
the dense, reacting fusion core. Electron collisions with
grids are never large compared to ion losses, and they
can always be suppressed by use of magnetic insulation,
if necessary. And, ion losses due to ion/ion collisional
upscattering by thermalization in the core, for instance,
are always going to be less than those from electron/
electron thermalization "which are much faster than for
ions#, which inherently give large electron particle loss
rates beyond their “birth” grid position, into the ion!
accelerating space. Any electron that enters this space
will be accelerated to the full!!acceleration drive poten!
tial, and will carry this energy to collision with the sys!
tem outer wall. Thus, the key to favorable power balance
in IXL is to minimize the two loss mechanisms outlined
above.
Unfortunately, as discussed previously, the IXL seems
unlikely to work at high fusion power unless the ICC
e$ect works or some other wave!group!trapping "WGT#
means of core densication can be employed, and this
inherently forces the system particles into a thermalized
or Maxwellian state. In such a state, the losses due to
electron upscattering into the ion!accelerating inter!grid
space will become too severe unless large bucking volt!
ages are applied to a secondary grids near the primary
inner grid. Similarly, a secondary grid must be applied to
the outer base grid at the ion injection position, to sup!
press losses of upscattered ions past their injection
point. By these means the upscattering losses of both
particles can be reduced to arbitrarily small values.
However, the voltages required may be as high as ten
times the desired well depth operating voltage
This can be achieved rather readily for DT, where the
well depth needed may be only 15 keV, for example, so
that the bucking voltages need be only about 150 keV;
and this is a tractable voltage. This level of electric
“stando$ ” will suppress particle losses by a factor of the
order of 22,000, limiting these to those found out in the
Maxwellian tail above an energy ten times that of the
mean. For DD or the DD!cat cycles the desired well
depth is more likely to be in the range of 50!80 keV, re!
IXL - Grid Cross-Section Showing
Face Emitter and Bucking Voltage
Insulating
Spacer
Coolant Channels
0.2 cm
1.0 cm
!Eeb
Figure 34 — Dual inner grid required in IXL for support of
bucking voltages; note “shadowing”
For this reason, IXL systems may well be able to operate
satisfactorily at net power only with DT fuels. If the
energy losses by particle upscattering are suppressed to
arbitrarily large values by bucking grids, the only remain!
ing fusion particle losses are those due to ion collisions
with the interior grids over which the ion ow passes. In
the case of a fully!bucked!grid protected system, the
ions will circulate over three grids; the injection position
outer grid, and the two inner grids. Thus ion/grid colli!
sions will be three times greater than for a single grid
pass system, reduced by the degree to which one of the
inner grids can be made to “shadow” the other, as sug!
gested in Figure 34.
Unfortunately, suppression of ion/grid collisions by
magnetic insulation is not practical, because of the in!
troduction of both V x B scattering into the ion system
in the vicinity of the grids by grid eld deections and
by ion/ion collisions near the grids, within the eld re!
gion. This B eld!induced deection will completely
defocus the ion radial motion and prevent formation of
the necessary small well!converged fusion core. Thus,
the limiting condition for upscatter protected IXL sys!
tems, operating with large ICC gain factors will be the
unavoidable geometric intercept fraction posed by the
multiple grid system. And, of course, the fusion product
particles will always “see” the grids as well, thus a frac!
tion of the total fusion power will be lost to the grid
system.
If the grid transparency is fgr, then the average number
of passes that an ion can make before striking a grid is
Ggr = 1/2 Ngrfgr, where Ngr is the number of grids through
which the ions pass, and the factor of two accounts for
the fact that the ion must “see” each grid twice in a
complete path from edge to core and back again. The
mechanical grid transparency can be related to the grid
geometry and system size by simple accounting of the
area subtended by the grid against the total sphere area
at each grid position. The e$ective intercept width of a
grid conductor is about twice its physical width, because
27
of the: distortion of local E elds introduced in the near!
eld region of each grid conductor. Thus, the e"ective
ion intercept transparency of a single grid conductor can
be written as fgr = 2dgrLgr/4#Rgr2. Here Lgr is the total
length of grid conductor, and dgr is its physical width. In
a great circle grid system $almost a minimum grid con!
guration% this conductor length is L = 6#Rgr thus fgr =
3dgr / Lgr. For three bucking grids in such an arrange!
ment, Ngr = 3, and the grid recirculation ratio becomes
Ggr = R/18dgr. There is thus an incentive to use narrow
grids to increase the grid recirculation. For example, a
system 270 cm in radius with a grid width of 0.3 cm
would have a grid recirculation of Ggr = 50. The trans!
parency of this system for fusion product intercept, as!
suming complete multiple grid shadowing, would be
0.0033 for a single pass over a single grid $or over a com!
pletely shadowed multi!grid system%.
The plasma performance with DT in such a system is
straightforward. System power gain is simply the ratio
of plasma fusion power to all losses. If the particle up!
scattering losses are kept to arbitrarily small values, then
the system gain will be determined by the unavoidable
grid intercept losses discussed just above, and the degree
to which central core densication can be achieved by
WGT/ICC phenomena. The overall system recircula!
tion factor for ion lifetime is given by the product of the
grid and ICC factors, Gi = GgrGICC. For net power, an
IXL system typically requires a total ion recirculation
factor above about Gi > 3E4, thus GICC & 600 would be
required with the grid system outlined above.
However, the ICC e"ect is limited by internal core pres!
sure balance $plasma beta% against the conning electric
eld energy density of the conning e.s. wave structures,
and by the surface pressure balance against internal elec!
tric wave eld pressure. Analysis of the maximum value
attainable for GICC under these limiting conditions
shows that it may. be as low as 10 or as high as 600, de!
pending on the actual wave length of the e.s. waves and
the e"ective core density as set by the density distribu!
tion. Thus, the ion recirculation required for net e"ec!
tive power in the example IXL system seems to lie at
the upper boundary of the GICC range. While this does
not auger well for its prospects, the actual physics limits
remain to be determined by experiments designed to do
so.
The EXL system behaves quite di"erently from the IXL
system, having none of its structure collision loss prob!
lems. In the EXL system all of the losses are due to elec!
trons escaping from the system through the magnetic
boundary provided by the polyhedral elds. Unlike the
IXL system, which exhibits losses from both species, it
is not di'cult to provide a potential well environment
in which the ions can be trapped for much longer than
their fusion lifetime without su"ering signicant parti!
cle losses from ion/ion collisional upscattering. And,
IEF in Electric Power Plants)
even if held long enough to create some upscattered
ions, their loss does not constitute a signicant energy
loss; it simply puts a greater load on the vacuum pump!
ing system, and the ions can always be replaced by fur!
ther injection.
Rather, in this device, the system is aimed entirely at
conserving the electrons that are the injection drive for
the ion!conning potential well. To achieve this e"ec!
tively, operation must be in an internal mode in which
the recirculating electrons provide diamagnetic currents
that push out the vacuum B elds towards a more!
nearly!spherical shape; limited only by the requirements
of MHD stability. Initially, when the system has no par!
ticles within it, the rst electrons will simply oscillate
across the core and be reected internally by mirror re!
ection $MR% in the vacuum eld cusps of the polyhe!
dral external eld. As ions and electrons are added, al!
ways controlled in proper proportion so as to maintain
the desired potential well depth, the central density be!
gins to build up, and that surface at which the electron
beta $in the cusp elds% is unity expands. As the position
of this beta=one surface $called the rb radius% grows with
continued particle input, it eventually reaches a radius at
which its surface area exceeds the area of all the cusp
gyromagnetic “holes” at that radius. Beyond this critical
radius the electrons can begin to exhibit diamagnetic
behavior against the conning cusp elds. Further injec!
tion drives rb further out, until it reaches a point at
which the ion and electron density gradient changes
sign. Density drops from the core outward until this
point, at which the density of both species grows with
increasing radial position. Beyond this radius $called r!%
the rb surface is unstable, and jumps to the outer bound!
ary so that rb = R.
At this condition the mirror reection recirculation ra!
tio GMR of electrons reaches its minimum $it is not zero,
because the maximum B eld lies slightly outside the
outer boundary radius, due to the curvature of the entire
system% and electron connement by MR is small. How!
ever, every electron that reects from a diamagnetically!
compressed mirror cusp eld will also enter and recircu!
late through the internal rb region. Electrons will pass
back and forth inside this inner boundary, reecting
from the external eld in the fashion of a steel ball re!
ecting from the interior of a sphere with small holes
drilled in its surface. Since such a sphere resembles a
“wi(e ball” plastic toy, this means of trapping is called
the Wi(e Ball $WB% mode, and the recirculation ratio is
GWB. The number of such recirculations possible before
it again “sees” a cusp “hole” through which to try to es!
cape the system is given just by the ratio of system sur!
face area to the area of the total number of cusp gyro!
magnetic escape holes in the polyhedral eld system.
Since the gyro hole size depends on the surface eld
strength, this internal recirculation ratio can be made
very large by using a su'ciently large magnetic eld.
28
And, because electrons are not very massive, the elds
actually required for high GWB are quite reasonable.
The overall system electron recirculation is simply Gj =
GWBGMB. Typically this must be in the range of Gj = 4 x
105 to 2 x 106 for net fusion power production in such
machines, using the range of fuels from DT to the DD!
cat cycles. This is, as expected, higher than the io! re!
circulation required by a factor that is approximately the
ratio of average electron speed to average ion speed in
comparable potential wells, or roughly as the square root
of the ion/electron mass ratio. For the fuels of interest
here, this is about a factor of 60x, comparable to ion
recirculation ratios of Gi = 7 x 103 to 3 x 104.
Wifeball Transitions
(MR Mode)
0
rx
V
rb
rado
rk
R
rk
R
(No WB Mode)
rx
0
(WB Mode)
(MR Mode)
V
rado
rad
rb
(MR Mode)
rx
0
rado
V
rad
rk
R
rb
(WB Mode)
Figure 35 — Sketch of transition from startup to fullydeveloped diamagnetic (WB) operation with increasing
drive current; system center is at left, r = 0, edge at right,
where r = R
Mirror/Wife Model for Gjo
G jMRo # 4
& 1
(1 " rb )t MR ) rb tWB ( ; tb : rx
%
(1 " < R ) $ N
' ttot
G jMRo # 4
& 1
(1 " rb )t MR ) G jWB rb tWB ( ; rb ; rx
G jo !
%
(1 " < ) $ N
't
G jo !
R
tot
e=
where: < R ! < t ) (< g " < r ) ; < q ! w
N
E0
1
m
#
&;
G jMRo !
m $1 " < q (1 " r
'
r
Ng
r ! rad or rb , whichever is greater
G jWB !
rb
2 re Z
E
; Z ! 8* nc rc2 and W ! 2 0 2
B0 R
Nk L2 S 2
(2 m)2)
#1 " < q (1 " rb
! ( ZW ) $
2
#$ f0 rb &'
m
)&
'; f r !
0
2
(m)2)
1) r
Figure 36 — Outline of GJ code formalism showing both
MR and WB components in computation of electron total
recirculation Gjo; note WB condition where <rb> = 1
This behavior is suggested in the sketch of Figure 35,
which shows the transition from no internal diamag!
netic electron current "at top of gure# towards fully!
developed diamagnetism "at bottom of gure# as a series
IEF in Electric Power Plants(
of positions along a radial line at di$ering degrees of
electron drive current and central density. The radial
marker rad shows the position of the “adiabatic” capture
radius of electrons in the local B eld. Note that it
moves outward as the B eld is pushed out and com!
pressed by the diamagnetic e$ect. The entire range of
electron operation has been modeled in a computer
code designed to determine electron recirculation at any
given set of operating conditions; this is called the GJ
code. It is used as a subroutine in the EIXL code to de!
termine Gj throughout the EIXL computations for any
given problem.
The general formalism of the GJ code is shown in Figure
36, which shows the relationship between the MR and
WB components. Note that, as rb ➝ 1, the mirror com!
ponent drops out in this formulation, and all that re!
mains is the wi%e!ball contribution. Later code modi!
cations "done since the work reported here# have
changed this to include the residual MR factor at full rb
expansion. Note, also, that the entire system behavior is
governed by one principal parameter, W = E0"BoR#2. This
is a measure of the gyro hole loss area to the system sur!
face area, and is inversely proportional to the maximum
value possible for Gj. The system can not operate at a
condition beyond <rb> = rb/R = 1 "called the rb1 condi!
tion#, because this is the beta=one limit at the outer
boundary. Operation at this state gives the highest pos!
sible Gj value, and allows estimation of the core density
that can be attained from the extremely simple formula
given in the gure; GjWB = 2r"Z/N#L2S2, where N is the
e$ective number of lossy cusps, #L is the size of the gyro
loss hole relative to the gyromagnetic radius of electrons
in the surface eld, and S2 is the mean!square value of
the sine of the particle velocity!vector capture!angle at
adiabatic capture of electrons in the residual cusp eld.
The parameter Z = 8&!crc2 is a measure of the core den!
sity for any given core size. Optimal operation is at this
rb1 condition.
There are two di$ering ways to drive an EXL system
from startup to “fully!wi%ed” operation at rb1. First is to
start the device as described above, with no internal
density, a high B eld, and slowly build up the internal
WB as the core density is increased. This eventually
reaches a condition where the WB sphere pushes out
the B eld and reduces the MR e$ect, until MR is gone
and only the WB connement remains. This mode is
shown in Figure 37, with constant "high# Bo eld and
variable "increasing# electron current. The other ap!
proach is to start with a Bo eld su'ciently small that
the system is in the rb1 condition at startup, and then
increase the B eld slowly, while continuing the electron
injection current at its initial "large# rate. This has the
e$ect of starting with very large loss holes through the
cusps, and decreasing these by increasing the B eld, to
raise the core density. Figure 38 shows this mode of op!
eration.
29
Start with Mirror ! End with Wifeball
Start with Wifeball ! Maintain Wifeball
B0 = constant
B0 = variable
Ie = variable
Ie = constant
Pure Mirror
Pure Wiffleball
Pure Wiffleball
Figure 37 — EXL operation from MR to WB modes, with
xed B eld and injection energy, but increasing drive current; system starts in pure MR mode and transits to pure
WB mode
The full spectrum of possibilities is shown by a three!
dimensional plot of operation in Gj, W, Z space. An ex!
ample of this is shown in Figure 39 for a truncated cube
polyhedral eld system with a central virtual anode
height fraction of ! = 0.272, and gyro hole radius loss
factor of !L = 2. The two extreme modes discussed above
are represented by following a line of constant W across
the 3!d surface, for the MR ➝ WB mode, and by riding
up directly along the limiting rb1 “ridge” line for the WB
➝ WB mode. These two modes can also be traced out in
the 2!d projections of Figure 40, that shows the system
behavior for a series of lines at constant W, as core den!
sity is varied, increasing with Z to terminate on the rb1
line. Note that starting along any line of constant W at a
value above that at which Gj = 1 intersects the Z axis,
results in a decrease in Gj as Z is increased, followed by a
rapid increase beyond some Z value, with an end point
on the rb1 line where beta=one "this is shown as the
dashed line rising linearly with Gj and Z across the g!
ure#.
Pure Wiffleball
Figure 38 — EXL operation in WB mode along rb1 line at
xed drive current and injection !energy with increasing B
eld; system operates in pure WB mode from start to end
point
Beyond that point on the W line at which the slope
reaches and exceeds unity "i.e. dGj/dZ > 1# the system is
unstable and will run away to the rb1 line. Because the Gj
value is decreasing with increasing Z on such a line, the
drive current must be increased greatly to pass through
the dipping “gate” of the W path. When the unstable
region is reached, the current then being used is more
than necessary, and the system expands to ll the com!
plete WB sphere, at which time the drive current is very
much less than required to get it through the “gate.”
This decreasing gain region is a result of internal WB
sphere expansion reducing the MR e$ect faster than
WB trapping is increased, leading to a decrease in total
trapping e$ect until the WB sphere becomes su%ciently
large and highly e$ective.
30
3D Plot of EXL System Operation
107
kL = 2, " = 0
.9,
r
"q = 0.995, #
= 0.995,
m = 3, N =
NMR = 8
WB
106
105
Gj
104
106
105
104
103
102
107
103
<rb> = 1
Gj
102
101
101
1020 19
10 18
Z= 10 10171016
8!
n 1015
c
rc
2
(1/
1014
cm 10
)
13
101
10-1
1
10-3
10-4
10-5
10-6
2
R
B
/
2)
E
W= 0kG cm
/
(keV
10-2
2
0
2
Figure 39 — Three-dimensional plot of EXL system operation over the parameter range of Gj, Z and W. Note that transition
from MR behavior on the at, lower-right-hand plane to growing WB operation on the slope, must occur through a “valley” in
the plot surface
Detailed calculations have been made to show the cur!
rent required to follow such a line. This has been found
to be as much as 10!15 times more than needed at the
nal operating state, thus operation as described, along
the MR ➝ WB mode is not useful. Much better is to
operate in the WB ➝ WB mode, which requires the
sam! drive current at all times "to zero order#; this is the
drive current needed at the nal state. Analysis of the
requirements for this all!WB mode show that the drive
current and power are related to the electron injection
energy by simple expressions Idriv! = K"Eo#0.5 and Pdriv! =
K"Eo#1.5. More complex version of these algorithms are
used in the formulation of the computer analysis code
for power balance studies of such systems. This is dis!
cussed later, below.
Electron Recirculation Ratio vs. Z
10
8
W = E0/[B02R2] (keV/kG2 cm2)
107
W=1
W = 10-1
W = 10-2
W = 10-3
W = 10-4
W = 10-5
W = 10-6
106
Gj
105
104
103
102
101
1011
1012
1013
1014
1015
1016
1017
1018
1019
1020
Z = 8!ncrc2 (1/cm)
Figure 40 — Electron recirculation ratio Gj as a function of
core density (Z-parameter) for various W-parameter values,
in an EXL/IEF system; 2-d plot of 3-d operating space
31
B. Plasma Power Generation
and Losses
In order to estimate the potential of IEC/IEF systems
for utilities power plants it is essential to be able to cal!
culate the output performance of typical conceptual
plants using a variety of candidate fusion fuels. To do so
requires careful estimation of those losses discussed as!
sociated with the plasma condition and operation is
these machines. As already discussed, the IXL system
has grid heating inherently unavoidable from collisions
of fusion products with the conning grids. For the ex!
ample given previously this amounted to about 0.33" of
the total fusion power; hardly a noticeable loss fraction.
However, the drive power requirements remain uncer!
tain, because there is no way at present to determine the
e#cacy of the WGT/ICC e$ect required to give IXL
systems a net positive power balance from any level of
fusion power. If the ICC recirculation gain is too low, no
IXL system can work at net power, whereas if it is su#!
ciently high, then net power can be made. Numerical
estimation of “the performance” is clearly impossible in
this state of knowledge. Theory suggests that the ICC
e$ect will likely not be high enough for net power, but
this conclusion is quite suspect in the absence of good
experimental data.
The EXL system can be readily analyzed, in contrast, on
the assumption that the electron diamagnetic e$ect will
%qualitatively& work. Complex PIC computer simula!
tions made since the work reported here support this
assumption. Various loss sources ha been examined, with
the result that the only di#cult one is that of brems!
strahlung from the dense plasma core in such systems.
Synchrotron radiation losses can be controlled by reec!
tive walls, in conventional fashion, and losses from par!
ticle collisional upscattering can be kept small by main!
tenance of ion ow anisotropy and well!converged
cores.19 But bremsstrahlung loss can not be made to go
away; it is the principal unavoidable loss mechanism in
EXL machines.
All IEC devices su$er from the fact that the brems!
strahlung radiation losses must be made up by the parti!
cle injection energy, because the fusion products do not
deposit their energy in the reacting fusion plasma as
they do in conventional M&M/LTE devices.
Bremsstrahlung Radiation
in IEF Systems
Bremmstrahlung power Pbrem (ne / ni )2
!
!
Fusion power
Pfus
, f Ei
in core !
Ee
Ei
>
(ne / ni )2
, f Ei
1" >
•
Non!equilibrium EXL/IEF systems can be made
to operate with:
• !<<1; E" << Ei in core
• Collisional equilibration not signicant over
electron life in machine
• Radiation cooling at high density suppresses
E"
•
Typically E" < 0.01 Ei in EXL %with small !& and
E" > Ei in IXL %with large !&
Figure 41 — Dependence of bremsstrahlung-to-fusion
power ratio on electron and ion energy in EXL system core;
note that small anode height yields greatly reduced bremsstrahlung
However, the EXL system %but not IXL& has a great ad!
vantage over these conventional approaches in that the
mono!energetic nature of the particle energies and the
vast di$erence in ion and electron energy levels across
the machine and especially in the core make it possible
to suppress bremsstrahlung generation to levels very far
below those of equilibrium plasmas. This is summarized
in Figure 41 that shows the basic formula relating
bremsstrahlung to fusion power generation in EXL sys!
tems. It is very important to note that the bremsstrah!
lung occurs at the core electro! energy, while the fusion
rate is set by the core ion energy, and that the two are
not equal. In fact, in an EXL core with small anode
height, the electron energy will be only ! times the ion
energy, thus choice of su#ciently small ! can reduce the
bremsstrahlung!to!fusion power ratio to arbitrarily small
values. This can not be done in the IXL system, as large
central virtual anodes are required to establish and
maintain wave!group!trapping core densication proc!
esses, and these result in Maxwellianized core electrons
as well. In EXL systems, the limit on reduction of core
electron energy is to that energy at which the ion and
electron speeds in the core are equal. This occurs at an
electron energy only %#"/#j& of the ion energy; much
smaller than the ! value needed to reach useful power
output ratios.
32
yearly energy consumption. At ve times the current
global rate, the DD resource would last for about one
billion years! By then the 3He resources of Jupiter !which
exceed the energy resources just cited by at least 105"
should be available! Of these fuels, the three preferred
fuel cycles are those using DD !and its two 3He#
catalyzed cycles", and p11B and self#generated#3He#
catalyzed p6Li, both of which are completely “clean” in
that they produce no neutrons. These are summarized in
Figure 43 which shows their features, and that of 50:50
D3He as compared to DD and p11B for neutron output.
The most desirable available fuel is D, taken straight
from the oceans, where it is found as one part in three
thousand !by mass" relative to hydrogen.
C. Fusion Fuels;
Characteristics and Choices
A wide variety of fuel combinations are available to IEF/
IEC fusion systems, most of which can not be used at all
in conventional M&M/LTE conceptual devices. Figure
42 shows these in tabular form, giving their abundance
and resource base in millions of metric tons !MT", and
equivalent energy content from their standard fusion
reactions, in gigawatt#years !GWY" per ton and total
energy in terawatt#years !TWY". The fuel cycles to
which these apply are also cited, as are the relative levels
of neutron radiation output from these cycles. Note the
huge energy resource available from the DD#half#cat
cycle, 1E11 TWY; about 5E9 larger than current global
Fuels in Electric Fusion IEC Systems
Fusion
Fuel
World Resource
Energy Content
Fuel Cycle
!ppm"
!MT"
GWY/T
Total !TWY"
Reaction
Neutron
1.1 x 105
5.6 x 1010
$
$
$
$
D
38
1.9 x 107
5.5
1.0 x 1011
DD!3He, 1/2 cat"
few MeV
11B !0.81"
4.3
2.1 x 106
2.5
5.3 x 109
p11B
none
6Li !0.07"
3.2 x 10#3
1.7 x 103
5.5
9.4 x 106
p6Li!3He,1/2 cat."
none
Li !natural"
4.6 x 10#2
2.2 x 104
7.2
1.6 x 108
D!T" Li
14.1 MeV
$
1 !estimate"
19.5
2.0 x 104
D3He
few MeV
H!p"
3He
•
GWY/T per ton of limiting isotope, 3He estimate of lunar resource
Figure 42 — Potential fusion fuels for use in IEC/IEF systems, showing availability, energy content, and fuel cycle features;
note abundance of both D and 11B as found in the ocean
33
IEC Fusion Fuel Cycles
Center of Potential Frame
Fusion Cross-sections for p6Li
Fuel
Neutrons
DD !3He 1/2 Cat."
Base
Intermediate
neutrons
p11B
Advanced
No neutrons
p6Li !3He Cat"
Advanced
No neutrons
Gj fusion cross-section (barns)
There are three preferred fuel cycles:
101
DT
1
p+B11
10
-1
TT
10-2
Baseline System
Partially#catalyzed DD fusion cycle, no tritium
burning
D) D?T ) p
D ) D ? He ) n
3
D ) He ? He ) p
3
4
Neutron capture in
in vacuum system
•
•
}
10B
fblanket ! 0.207
fneutrons ! 0.092
E f ! 10.67 MeV / DD
blanket; tritium,
3He
only
Minimizes materials problems, avoids enormous
materials development, money and time
Minimizes radiation hazards and inventory, no
14.1 MeV neutrons, no tritium in blanket, no Li
Figure 44 — Baseline 3He-catalyzed DD fusion fuel cycle,
including fusion with 3He produced in the second DD
branch reaction; tritium from rst branch is removed and
stored for later use
The most useful cycle for D fuel is that which includes
the reaction of the 3He produced in one of the two
branches of the basic DD fusion process, as shown in
Figure 44. Note that the average energy per basic DD
fusion reaction is increased to over 10 MeV by the use
of the !free" 3He fuel ion, and that the fractional neutron
energy production is only about 9$ !cf. DT which has
over 80$ neutron energy fraction". The one neutron
produced here is only at an energy of about 2.5 MeV, and
can readily be thermalized in an external blanket and
captured in 10B, to give still more energy for power pro#
duction. There is no need to conserve this neutron for
the tritium production needed in DT systems, or for
criticality as required by ssion reactors. Thus there is
every incentive to remove it as swiftly as possible by
capture in some useful and non#noxious material.
10-3
DD
1
101
E0 (keV)
102
103
Figure 45 — Center of potential frame fusion crosssections; p6Li connects to 3He6Li, Figure 46
Center of Potential Frame
Fusion Cross-sections for 3He
Gj fusion cross-section (barns)
Figure 43 — Preferred IEC fuel cycles, with least neutron
radiation and reasonable system size
pLi6
101
3
He6Li
1
D3He
10-1
3
10-2
10-3
1
101
He3He
E0 (keV)
102
103
Figure 46 — Center of potential frame fusion reaction crosssection for 3He fuel combinations used eventually to add to
the D-catalysis by T-storage and decay-3He recovery
The other DD branch reaction produces a triton !T"
which is weakly radioactive !6 keV beta decay" over a
12.7 year half#life. This T decays into another 3He, thus
can be important NOT to react the tritium produced
within the reacting IEF core, and this will not happen
for reasons previously given. In this respect, the IEF
system is quite unlike the M&M/LTE systems, in which
it is impossible no! to burn the tritium from such a reac#
tion. That is, it is impossible to remove any such tritium
produced before it is burned. This is because the reac#
tion cross#sections for DT are so much higher than for
anything else in the relevant ion energy regime. This is
shown in Figures 45 and 46 which give the variation of
fusion cross#sections with energy for a variety of fuel
combinations. Note that the energy of interest here is
that in the “Center#of#Potential” !CoP" frame character#
izing the IEF potential well geometry. This is impor#
tantly di%erent from the energy usually considered only
for ions with Z > 1, for these will gain Z times as much
energy when they fall into the potential well as will their
Z = 1 companions. The IEF system thus o%ers au easy
34
means to achieve very high reaction energies, especially
for high Z fuels such as 11B. Note, also, that the D3He
cross!section is nearly as high as that for DT, which
makes attainment of the DD!half!cat cycle quite easy.
The CoP frame is illustrative of the shift in cross!
section energy that the IEF system makes possible but,
of course, all reaction calculations must be made in the
equivalent Center!of!Mass "or momentum# frame to
attain comparative performance information for uni!
form, isotropic scattering and fusion collisions.
IEC Fusion Power Core (FPC)
and Blanket System Conguration
•
•
FPC physically separate and decoupled from
blanket/shield
Separation allows optimal design and function of
each
Blanket/
Shield
D. General Power Plant Concepts
All IEF systems considered here are spherical or quasi!
spherical. This results from their fundamental emphasis
on strong central convergence in counter!streaming ow
of the ions that create fusion. Cylindrical systems can
also use IEF/IEC principles, but their performance can
not be made as attractive as can spherical systems. And,
spherical systems are found to be able to reach useful
power gain conditions at relatively modest total power
output, thus are not limited to systems applications in
the very large single!unit power plants of recent years,
that seem to be of little interest to the electric utilities
industry.
This compact spherical conguration of the IEF fusion
power core "FPC# unit allows adoption of another favor!
able concept for the complete fusion system. Since the
desirable DD!half!cat fuel cycle produces a weak neu!
tron output, it is necessary to have an external blanket
to absorb this neutron and make use of the energy asso!
ciated with it and with its capture. This blanket need
not be connected to or be part of the FPC device, which
can be a stand!alone unit centrally!located within the
blanket system. The advantages of this are numerous.
This allows separate design optimization of each subsys!
tem; allows easy replacement "not maintenance# of the
FPC; adds to safety by separating failure modes of the
FPC from those of the blanket system; allows blanket
design to optimize lifetime and thermal output; etc.
These are suggested in Figures 47 and 48 that show the
arrangement and features just described.
FPC
Figure 47 — Schematic outline of IEC Fusion Power Core
(FPC and blanket conguration)
IEC FPC/Blanket System Conguration
This has three main advantages:
"1# IEC/FPC removal/replacement as a unit
•
•
No in!situ maintenance
Limits long!term radioisotope inventory buildup
to blanket
"2# Subsystem fail!safety
•
•
IEC/FPC failure does not a$ect blanket
Blanket failure does not a$ect FPC
"3# Blanket design optimized for
•
•
Life, economics, fabrication
Safety, handling, maintenance
Figure 48 — Design and operational features of decoupled
FPC/blanket conguration
A typical power plant that uses such IEF/FPC units will
look not markedly di$erent from a conventional nuclear
ssion plant, with the exception of the reactor building.
For the DD~half!cat IEF system, this building will be
somewhat smaller "and less expensive# than that for s!
sion reactor containment, as there are no ssion prod!
ucts and associated hazards or decay heat to worry about
in the IEF system. The balance!of!plant will be essen!
tially the same as for an ordinary ssion reactor plant,
with its steam cycle, turbine generators, cooling towers,
etc. Figures 49 and 50 show an artist’s conception of
such a plant, and of the reactor building with its FPC
units and their individual blankets.
35
Fusion-Driven Electric Plant
Baseline System
•
•
•
•
4
3
2
•
Thermal conversion, steam cycle
FPC power removed through rst wall
Maximum steam/water temperature = 650 oF,
pressure = 2000 psia
Stainless steel rst wall heat ux sets minimum
size
• TZM/Mo/alloy Cu allow smaller sizes
Use stainless steel for all other structures
• FPC shell, envelope, blanket shell, structure
• Blanket power density low "< 4 MWth/m3#
Figure 51 — Properties and features of Baseline System
using the DD-half-cat fusion fuel cycle
1
Figure 49 — Typical IEF/FPC power plant based on DDhalf-cat fuel cycle and thermal conversion using conventional steam turbine-electric power generation
Reactor Building
Figure 50 — Reactor building containing several FPC/
blanket modular unit cells, showing arrangement of FPC
units in individual shielded modules, feeding common
steam system
The plant shown here uses a multiplicity of FPC/blanket
systems to obtain fail!safety, high reliability, ease of re!
placement, and modular power output. By such modu!
larity, plant capacity can be maintained or adjusted by
having one FPC unit always on standby; its cost is small
enough that this is economically feasible.
The baseline FPC system of interest here is chosen to
have the favorable low technology features summarized
in Figure 51. The basic approach allowed by the very
attractive nature of the DD!half!cat fusion fuel cycle is
to utilize this in an IEF/FPC unit that is designed to
operate at conditions no more stringent than those cus!
tomarily used for conventional power plant steam sys!
tems. Thus, stainless steel is used throughout, tempera!
tures and pressures in the cooling water/steam system
are limited to those experienced in conventional PWR
systems, wall thermal stresses are kept low by limited
allowable rst wall heat ux, and the blanket system is
placed at a distance large enough to keep power densi!
ties very low and easy to handle. There are no ssion
products anywhere in the system, and this can be de!
signed so that no signicant decay heating problems can
ever arise, no matter the accident or failure condition.
Under these constraints the IEF system size will be de!
termined by the rst wall thermal stress rather than by
any considerations of fusion plasma behavior. Figure 52
shows the IEF system radius required as a function of
rst wall power ux, and indicates the range allowed for
conventional stainless steels and for advanced materials
such as TZM and special copper alloys. Note that a rst
wall ux of 3!4 MWth/m2 gives a system radius in the
range of 3!4.5 m for power levels of 500!1000 MWth.
The radius required for good plasma performance is only
about 2!2.5 m for the DD!cat cycle, thus this thermal!
design!driven system will provide a very large margin for
successful operation of the plasma system.
36
Geometry Limits Size, Power and Heat Flux
Flux limits for stainless steels
Radius (m)
10
Flux limit for TZM/alloy Cu
blanket system cavity, to swimming pool storage before
shipment to eventual disposal sites. The replacement
unit can be lowered into place on the base plug hole, and
attached under manual operator control.
IEF FPC/Blanket Cell Module
5
Hot water
return line
0
0
5
10
Power Flux (MW/m2)
500 MW
1000 MW
15
20
2000 MW
Figure 52 — Limitation of IEF system radius by thermal
stress in rst wall, as a function of system power and wall
ux; note maximum size for stainless steel, TZM and copper alloy materials
To/from
primary heat
exchanger
External
shell
To/from
primary heat
exchanger
Blanket H2O + 10B
region
Blanket reflector
(Be, C)
e - guns or
ion - guns
(6 places typical)
Cooling
water/steam lines
Inert gas
1m
60 cm
Since the fusion products leave the plasma!conning
region and strike the rst wall, it is essential that the
primary cooling system be placed inside this wall. This is
easy to do by using spherically!spiral tubing inside the
vacuum shell, cooled by pressurized water. Figure 53
shows such an arrangement for top and bottom water
headers, allowing constant coverage and ow velocity in
each cooling tube.
Typical First Wall and Cooling Structure
External Vacuum Shell
0.8 cm
1.0 cm
Water/Steam Tubes
Figure 53 — First wall cooling tube arrangement to provide
constant coolant ow velocity and equal cooling Coverage
per unit area; tube-cooled wall mounted inside vacuum shell
The complete IEF unit can then be mounted inside an
external envelope, as shown in Figure 54, which serves as
a containment shell for a neutral gas ll around the sys!
tem to suppress electrical breakdown and provide addi!
tional safety against leaks in the main IEF unit or the
primary cooling uid piping. The required vacuum is
maintained by pumping through a large vacuum port
located at the bottom polar axis of the system. This can
be brought through a shielded base plug, together with
the electrical drive power and uid supply and removal
lines and fuel feed systems, et al, by means of folded
paths to minimize neutron leakage. The whole assembly
is mounted on a base plug hole in the roof of an under!
ground tunnel, in which man!rated operations can be
conducted while the units are operating. Thus, replace!
ment can be done by operator controlled detachment of
supply systems, followed by remote removal from the
9 cm
31 cm
Base Plug
Deterium
Fueling
Water out, to
Water in, from
}
Primary heat
exchanger system
Guns / grid drive power
Vacuum pumps
Figure 54 — IEF FPC/blanket cell module arrangement,
showing external FPC containment shell and removable
shielded base plug for power, fuel and coolant throughput
The mass of this Baseline FPC System is about 78.3 met!
ric tons, including the mass of the borated concrete and
folded path arrangements in the base plug of the unit.
This plug dominates the total FPC mass, being 55 tons,
over 70" of the total mass. The second largest mass
item is that of the FPC external pressure container shell
and the internal vacuum shell with its associated rst
wall cooling tubes. A summary of the mass breakdown
of the Baseline FPC System is given in Figure 55, which
separates the FPC unit from its base plug. The overall
features of this Baseline System are summarized in Fig!
ure 56, which gives an overview of the major characteris!
tics of the design. Note that every aspect of this system
is aimed at minimizing the level of technology required
for practical fusion power production. For all practical
purposes this system acts like a PWR ssion power sys!
tem, but without its hazards and radioactive ssion
products, and with the fail!safety inherent in the IEF
power unit. It is hard to conceive of a situation in which
any accident could create any signicant hazard to the
general public from failure of an FPC unit.
37
Baseline System
FPC minimum unit mass = 23.3 T
Shells = 14.8 T
Guns = 3.6 T
Pipes = 3.9 T
Misc. = 1.0 T
"EXL# using normal conductors, and those due to the
cryogenic cooling power requirements of superconduct$
ing magnets for sic magnet systems. These latter include
superconducting power removal requirements set by
internal neutron heating in the superconducting mate$
rial.
IEF Power Plant - Power Flow Diagram
FPC base plug = 55.0 T
Concrete = 40.0 T
Blanket plug = 15.0 T
Drive Power Waste Heat
Confinement
Power
Coils
Neutrons
Blanket
FPC assembly mass = 78.3 T
Thermal
Reacting
Grids
Figure 55 — Baseline System removable FPC unit mass
distribution, including base plug; EXL DD-half-cat IEF power
system comparable to PWR ssion reactor
Baseline System
•
•
•
•
•
Thermal conversion, steam cycle
FPC power removed through rst wall
Maximum steam/water temperature ! 650 oF,
pressure ! 2000 psia
Stainless steel rst wall heat ux sets minimum
size
• TZM/Mo/alloy Cu allow smaller sizes
Use stainless steel for all other structures
• FPC shell, envelope, blanket shell, structure
• Blanket power density low "< 4 MWth/m3#
Figure 56 — Overview of IEF/FPC DD-half-cat Baseline
System characteristics and features; note similarity to PWR
system power conversion and operating level technologies
E. Power Balance in IEF Systems
In order to design IEF power systems it is essential to
be able to estimate their power output properties in
terms of their critical design technologies. To do so re$
quires a design analysis tool that relates the performance
of each critical limiting technological feature to the
overall system performance. This has been accomplished
by use of a complex computer program that calculates
power balance in such systems, including all relevant
losses and sources of energy in the system, as a%ected by
the physics and engineering conditions imposed for its
operation. This code, called the PBAL code, was devel$
oped earlier and modied in the current work to study
the DD$half$cat fuel cycle "as well as others#. It includes
detailed determination of the losses due to magnets
Conversion
Fusion
Photons
System
“Plasma”
Guns
Charged
Particles
Acceleration
Power
Direct
Conversion
Gross
Electric
Power
PNET
To Grid
Drive Power Supply
Figure 57 — Power ow diagram for IEF power plant, showing all major interactions and phenomena considered in the
PBAL power systems parametric analysis code
The code also includes the injection power required for
EXL system drive in the WB mode, along the rb1 line,
according to the theoretical model developed for this
plasma physics behavior, and that required to make up
bremsstrahlung losses from the ion/electron system.
Bremsstrahlung is determined by analysis of the radia$
tion output over the density distribution obtained from
the EIXL code, for any given potential well depth. Fu$
sion power generation is calculated by integrating the
fusion power density over the code determined density
distribution throughout the machine. The code allows
input of the ICC e%ect as desired, by an I/O toggle. The
entire collection of losses and fusion power generation is
summed into the PBAL code structure, which then cal$
culates the gross fusion power, bremsstrahlung power,
magnet power, and net fusion power output as functions
of both the electron drive energy "maximum well depth#
and conning B eld for EXL systems. The design pa$
rameters available for variation in input include the sys$
tem size, virtual anode height, fusion fuel choice and
mixture ratio, both thermal and direct conversion e&$
ciencies, ion convergence radius, magnet type and cool$
ing system, and specication of blanket energy per fu$
sion event.
38
IEC/IEF Power Plant Schematic
3
PC
!DC
Plasma
Blanket/
Shield
Cleanup
Recovery
First
Wall
Magnet
Set
Power to
Grid PE
Turbine
Generator
!TH
Heat
Rejection
Main
Coolant
Pump
Condenser
Feedwater
Pump
Coolant
Inlet
FPC
Fueling
Steam
Generator
Coolant
Outlet
D3He Fueling
Vacuum
Pumps
PET
!DC
He
Tritium
Recovery
Electric Plant
Equipment
Direct Energy
Conversion
Ion/Electron
Injection
Current Drive
Deuterium
D3He Fuel
Prep
Miscellaneous Plant
Equipment
Heat
Rejection
Special
Materials
Primary Heat
Transport
Turbine Plant
Equipment
Reactor Equipment
Reactor Plant Equipment
Figure 58 — Overall IEC/IEF power plant schematic outline, showing fuel supplies, recycling system, fusion unit, conversion
equipment and electrical subsystems of complete plant
This code is built to apply to the general IEF fusion
power system ow diagram shown in Figure 57 and cov!
ers the operating features of the complete power plant
shown in Figure 58. The general nature of the PBAL
code used for analysis of these power plant systems is
summarized in Figure 59, while Figures 60a,b show the
formalism used in this code for each of the major
sources of power loss and gain. The output can be dis!
played in a variety of ways. The performance parameters
of most interest are the net electric power from fusion
Pne! and the system gross gain Ggr "given by Ggr = Pgr/Pinj,
where Pgr is the gross electric power produced from the
system#. Typically, these are plotted vs. electron drive
energy for a range of system size and other specied
parameters. Figures 61 and 62 show these parameters for
the DD!half!cat cycle operated at two di$erent B elds;
8 kG and 10 kG. Figure 63 shows these for the D3He
"50:50# fuel mixture at 10 kG, for comparison.
EXL Power Balance Calculation Using PBAL
•
•
•
Design point analysis of fusion power, injection
power, magnet power, and bremsstrahlung losses
Parametric variability in injection energy, B eld,
device radius, anode height, fuel mixture, thermal
and direct conversion e%ciency, ion convergence
radius, magnet type and cooling system, and blan!
ket energy
Reporting options include Ggr "Eo, B#, Ggr "E0, R#,
and Ggr "B, R# and the companion functions for
Pne!, Pfus, Pbre", and Pmag
Figure 59 — Power balance computer code used in analysis of IEF systems, showing design and performance parameters considered, and output data options and reporting
39
PBAL Code Algorithms
P ) >c #$ Pinj ) Pbrem &' ) P
e
fus
Ggr !
e
mag
Pinj ) Pbrem ) Pmag
P
P
f0 rk ke N 5 k L S 6 51 " < R 6 LN 51 / 0.016 3/ 2
Fe
E0 ;
2 me
2 re ks
LN 51 / rc 6
2
3
Pinj !
!
!
e
out
e
in
40.8 51 " < R 6
ln 51 / rc
6
C fus !
10
101
5 x 102
Ggr
4 x 102
3 x 102
1
2.5 x 102
2 x 102
E03/ 2 ; where Fe ! 0.0610
W
0N
< R ! < r ) / < g ) min " < r 2 g
.
1 N
W
e
! C fus
Pfus
Ggr vs. E0, DD-half-cat, B = 8 kG
2
1.5 x 102
10-1
1 x 102
10-2
1
101
Figure 61a — EXL/IEF system gross gain and net electric
power with DD-half-cat, B = 8 kG
5 Bc R 64 ICC;
rc
, 5E6 E K f
480* E03/ 2
e
f
Pnet vs. E0, DD-half-cat, B = 8 kG
bij
2 51 " > 6
1x10 "24
mi ke #1 ) f2 5 Z " 162 &
$
'
E ef ! Etot #$> f 51 " fblkt 6 ) >c fblkt &' ! Ecp> f ) @ Ent ) Eblkt A>c
> f ! >dc 51 " >c 6 ) >c ; Formulas in Pbal
10
Pbrem ! Cbrem
5
rc
Cbrem
6
#1 ) f2 Z 2 " 1 &
' ; Forrmulas in P
Fz ! $
bal
)
"
1
1
f
Z
6&'
25
$#
Pmag ! Cmag B02 R; Cmag !
50 B N
; P e ! >c Pmag (normal)
* fi Fs 5 AR 6 mag
Pnet(W)
2.5 x 102
10
all formulas are for rb ! 1
Figure 60b — PBAL code algorithms for calculation of main
power losses in IEF systems
The Baseline System design performance has been de!
termined by the use of these codes "EIXL and PBAL#,
for an arbitrarily!limited rst wall heat ux of 3.2
MWth/m2 and fusion power generation of 1300 MWfus.
The system was chosen to utilize 10B ssion as the
means for neutron disposal in the blanket system, and
the blanket was designed so that the blanket rst wall
neutron ux would always be less than 0.34 MWnts/m2
"fast neutrons#.
3 x 102
8
2 x 102
1 x 102
107
106
1
101
E0 (keV)
103
102
Figure 61b — EXL/IEF system gross gain and net electric
If the neutron!dose!limited lifetime uence of this wall
is 12 MWyear/m2, then the blanket wall would survive
for over 35 years, and no replacement would be neces!
sary over the power plant economic write!o$ life.
Ggr vs. E0, DD-half-cat, B = 10 kG
e
! >c fi Pcp (superconducting)
Pmag ! Fc F= fi fnt Pnt ; Pmag
Note: Cross-sections are for equal injection radii,
5 x 102
4 x 102
PBAL Code Algorithms
1.69 x10 "32 Fz K b >e
!
48* ke2 E03/ 2
10
109
Figure 60a — PBAL code algorithms for calculation of injection and fusion power and gain
5 Bo R 64 ICC;
103
102
E0 (keV)
10
2
5 x 102
101
4 x 102
Ggr
3 x 102
2.5 x 102
1
2 x 102
1.5 x 102
1 x 102
10-1
10-2
1
101
E0 (keV)
102
103
40
Pnet vs. E0, DD-half-cat, B = 10 kG
1010
5 x 102
4 x 102
109
3 x 102
2.5 x 102
Pnet(W)
2 x 102
108
1.5 x 102
1 x 102
107
106
1
101
103
102
E0 (keV)
Baseline System — Power and Size
The IEF unit rst wall is closer to the neutron source,
and would last only about 20 years at the uence limit
above. If desired, replacement could be made at 17!18
years to assure reliability, with a new unit plugged into
the base plug hole in the modular IEF/FPC array in the
reactor building. Figure 64 shows details of the power
balance for this Baseline System. Note that operation at
1300 MWfus leads to a net electric power output of
slightly over 500 MWe, under the thermal conversion
e"ciency of 40# and other power ow assumptions
used in the design.
Ggr vs. E0, 50:50 D3He, B = 10 kG
5 x 102
4 x 102
3 x 102
Ggr
2.5 x 10
2
2 x 102
1
1.5 x 102
1 x 102
10-1
10-2
1
101
E0 (keV)
Precirc = 122 MWe; fR = 0.195
FPC
Pfus = 1298 MW
Pnst = 154
Pcp = 1141
1263 MW
Blanket
-154
Pbl = 144
Thermal
Conversion
!c = 0.4
Gross Electric Power
Pegr = 624 MWe
298 MW
Ptot = 1561 MWth
Pnet = 502 MWe
FPC radius RFPC = 4.5 m; First wall heat flux "htx = 3.2 MWth/m2
Blanket radius Rblkt = 6.0 m; Blanket wall neutron flux "nt = 0.34 MWnt/m2
Figure 64 — Baseline System power ow and size, showing
detailed power balance in all major subsystems, including
bremsstrahlung conversion and 10B ssion power in blanket
102
101
This gives a net e"ciency of electricity production rela!
tive to fusion power of about 38.7#, even with a total
injection power of 122 MWe in the driving electron
beams. All of this energy eventually appears as heat
available for thermal conversion and this, plus the power
generation from 10B ssion, adds to the total above the
basic fusion reaction power. This system has an IEF ra!
dius of about 4.5 m, set entirely by the small value as!
sumed for rst wall heat ux. The design fusion power
generation could have been achieved with a considerably
smaller unit, but the rst wall heat ux could not have
been handled by conventional stainless steel materials;
advanced copper alloys would be required.
102
103
power with 50:50 D3He, B = 10 kG
Even though larger than needed from plasma physics
reasons, the IEF system is still small relative to other
concepts for comparable fusion power output. This is
indicated in Figure 65 that shows an outline comparison
of the Baseline IEF System with the highly!advanced,
hypothetical Starre Tokamak M&M/LTE fusion sys!
tem and with a conventional central station PWR s!
sion reactor power/steam source. The IEF and PWR are
comparable in size, and both are at least a factor of
about 30x smaller in size $volume% than the large M&M
tokamak machine.
Pnet vs. E0, 50:50 D3He, B = 10 kG
10
10
5 x 102
4 x 102
9
3 x 102
Pnet(W)
2 x 102
10
2.5 x 102
1.5 x 102
108
1 x 102
107
106
1
101
E0 (keV)
102
103
power with 50:50 D3He, B = 10 kG
41
Starfire Tokamak
15
10
5
0
Baseline (DD)
IEC
5
PWR
10
15 meters
Comparative Sizes
of
Nuclear Power Sources
Figure 65 — Comparison of main power source size (m) for
tokamaks, PWR and IEF systems; note that volumes of IEF
and PWR are comparable, and about 30x smaller than tokamak
Finally, it is of some interest to compare the IEF system
to other tokamaks and power sources on the basis of
volumetric power density and specic power per unit
mass. These provide a crude measure of capital cost due
to size !power density", and capital cost due to material
!mass#specic power". This comparison is given in Fig#
ure 66, taken from the work of Krakowski.20 In general,
power plant steam source system costs decrease with
higher values for both of the parameters plotted, thus
lower costs arise for parameters moving up and to the
right in the gure. The DD#half#cat cycle IEF systems
shown are marked D!DD" for normal coil magnets and
thermal conversion, S!DD" for superconducting magnets
and thermal conversion, and DC!DD" for superconduct#
ing magnets and partial direct conversion !ca. 60$ con#
version e%ciency".
Source Volumetric and Mass Power Density
FPC Volume Power Density, PE/VFPC [MW(electric)/m3]
Main Power Source Power Comparison
10
TITAN-II26,27
LWR86,87
CANDU93
Fossil (Coal)
56
MSR (! = 0.08)47 HFCTR
10-2
10
CRFPR25,44
DC(DD)
S(DD)
D(DD)
TPSS55
ARIES-I28,64
NUMAK83
STARFIRE
10-1
p11B
CSR51
94
FBR (MPR)91 OHTE54
AGR89 SG87
GENEROMAK74
FBR (SP)88
MHTGR92
1
PWR86,87
TITAN-I26,27
CRFPR25,44
23
MkIIB84
RFPR85
UWTOR-M48
MARS24
MSR (! = 0.04)47
EBTR37
UWMAK-I90
10
2
PE = 1000 to 1200 MW(net electric)
103
104
105
FPC Mass Power Density, 1000 PE/MFPC [kW(electric)/ton)
Figure 66 — Volumetric and mass power density of various
sources,20 a measure of specic capital cost and cost-ofenergy; note favorable performance of IEF systems
Note that all of these systems o&er performance mark#
edly better than any other power plant systems or con#
cepts shown. Also shown is the region expected for op#
eration of clean p11B systems !no neutrons produced,
blankets are not required, and high e%ciency direct
conversion is possible". This is an order of magnitude
more favorable than even the DD#half#cat systems.
But, of course, the basic IEF fusion source unit is more
di%cult to develop and requires a much higher level of
engineering technology as well as IEC plasma physics to
achieve the desirable operating states that correspond to
its power systems position in the diagram of the gure.
42
6. Engineering
Issues
A. Conversion of Fusion Products and
Energy
As previously discussed, the Baseline System examined
here uses an ordinary pressurized water steam cycle for
heat removal from the IEF unit, and a safe boron s!
sioning blanket system for capture of the small frac!
tional neutron production from the DD!half!cat system.
This conversion proceeds at roughly the same design
and operating conditions as are found in PWR ssion
power plants, thus nothing new or surprising is expected
or found in such steam cycle thermal conversion. How!
ever, it is quite clearly aesthetically unattractive to take
the charged particle fusion products at “temperatures”
of millions of degrees "from their very high speeds# and
use them to produce steam at hundreds of degrees, in
order to undergo a relatively ine$cient subsequent con!
version to electricity.
A more attractive option exists for IEF systems, and this
is to convert the energy of charged fusion product "mo!
tion# directly into electrical energy. This can be done at
high e$ciency by use of electrical decelerating grids
around the charged particle fusion source, biased so as
to slow down the particles and extract all of their energy
into the grid system before they come to nal rest at the
outer wall. If the bias voltage is made the same as the
energy of the fusion product particle divided by its
charge state, then the e$ciency of conversion can ap!
proach 100%. Work already accomplished in the labora!
tory 21 has achieved 70!80% e$ciency with systems in
which mixed plasma energies were found. The IEF sys!
tems are uniquely suited to such conversion because
they inherently must allow their fusion products to es!
cape beyond the connement boundary, either to strike
a wall and produce thermal energy "the steam cycle# or
to enter into a concentric, spherical direct! conversion
grid system.
system because of its strong surface magnetic elds that
extend beyond the outer boundary of the machine, but
can not be accomplished in the IXL system. Figure 67
shows the conversion e$ciency possible for various fuel
combinations and the approximate width of the DCS
grid system required for this e$ciency under the two
extreme assumptions of limiting voltage gradient.
Direct Conversion Process Schematic
KDC = 50 keV/cm 0
1
2
KDC = 12.5 keV/cm 0
4
8
1.0
p11B
3 L (m)
DC
12
DD 1/2 cat.
fDC
D3He
0
5
EDC(MeV)
10
15
Figure 67 — Schematic diagram of direct conversion
process for charged particle fusion product kinetic energy
for several fusion fuel mixtures; note high efciency possible for p11B
Note that nearly all of the energy of the alpha particles
produced in p11B fusion can be converted with a modest
2 MeV convertor that may be only 0.4!1.6 meters or so
in width. This is because these alphas are at xed and
predictable energies, none exceeds 4 MeV, and their
charge state is always Z = 2. In contrast, it is much more
di$cult to convert the energy of the proton produced in
the D3He reaction because it has very high energy
"nearly 15 MeV# and only Z = 1. This problem appears in
the DD!half!cat system, however nearly 60% conversion
e$ciency of the energy of all of the charged particles
involved in this cycle can be recovered directly by a 5
MeV convertor over a span of 1!4 meters. The remainder
of the charged particle energy must be removed by a
thermal cycle from the external walls of the system.
If the allowable bias voltage gradient that can be sus!
tained in this direct conversion system "DDS# is KDC
kV/cm, the stando& width of the DCS grids must be
LDC = Ecp/Z KDC cm for the charged particle energy Ecp in
kV, with a charge state of Z. The practical range of KDC
has been found to be about 12.5 < KDC < 50 kV/cm for
many high voltage systems. If the voltage gradient is
made too high, then secondary electrons emitted by
energetic ion impact on structures "grids, etc.# will be
accelerated to cascade into a breakdown arc. Suppres!
sion of such electron acceleration is inherent in the EXL
43
B. Impurities and “Ash” Production
In conventional M&M/LTE fusion concepts, the genera!
tion of high!Z ions in the plasma mixture will cause
greatly increased radiation losses and cause the plasma
fusion system to shut down, being unable to sustain igni!
tion operation under such lossy conditions.
Sputtering Yield from Deuteron Impact
summarizes the situation for IEF systems in respect to
impurity generation by plasma ion impact on system
structures.
High Z Impurities in IEC Systems
Electric potential removes high Z ions and prevents
buildup in dense core region.
Impurity ions from EXL outer wall removed by:
1
•
•
Cusp losses
Wall structures
10-2
•
•
DD/D3He in
IEC Systems
Be
Fe
C
0
-E
0
Mo
W
10-3
10-4 1
10
10
2
EXL
(r/R)
1
0
IXL
-E
0
(r/R)
Impurity ions removed by
internal grid collisions
1
Internal
grids
Figure 69 — Electrostatic potential well distribution and
high-Z impurity motion, trapping in and removal from IXL
and EXL IEF systems; note trapping impossible in core region
10
3
Energy (eV)
10
4
10
Figure 68 — Sputtering yield from deuteron impact on various rst wall coating materials as a function of D energy; D
is at high end or very low end of energy scale in IEF systems
In IEF systems these same problems never arise, for the
system is not an ignition system, and the central core
region is both very small at high density, and positioned
in the potential well in such a way as to exclude high Z
materials generated at surfaces external to the core. The
sputtering yield from structures in IXL and EXL sys!
tems is also low, because the ions involved are either at
very low energies "EXL surface region# or very high en!
ergies "IXL grid/conductors#. As can be seen from Fig!
ure 68, the sputtering yield per ion impact for D ions, is
low at either end of the energy scale, especially if mate!
rials such as carbon or beryllium are used as coatings
towards the plasma. But even if sputtering does occur,
the shape of the potential well within the IEF system
e$ectively excludes the particles from any signicant
residence time in the core "EXL # or traps them in a re!
gion far from the dense core, where they can not give
rise to any signicant radiation losses "IXL#. This is
shown in Figure 69 for each IEF system. Figure 70
Internal grids
Wall structures
Electric Potential (E)
10-1
Electric Potential (E)
Sputtering Coefficient
Impurity ions from IXL outer wall removed by:
5
The production of “ash” is an inevitable consequence of
the fusion event for “ash” is nothing more than the high!
Z fusion products themselves. In most cases of interest,
this “ash” is simply 3He, which is trapped and collects
within the plasma of M&M/LTE machines and increases
bremsstrahlung radiation losses due to the Z depend!
ence of bremsstrahlung output on charge state. The
production rate thus is simply the fusion rate. However,
unlike M&M/LTE fusion devices, the probability for
trapping of fast fusion produced alphas in the core of
IEF machines is very small because it depends on the
collision of the fast alpha with an in!situ core ion. This
collision cross!section for scattering is very small at core
alpha energies, and the path length available in the con!
verged core is also very small. These two factors to!
gether always lead to the result that the maximum pos!
sible buildup of core!trapped fusion product alphas is
limited to less than 5 x 10!3 of the core ion density.
44
Sputtering and Impurities in IEF Systems
•
•
•
•
•
Not a problem in either type
EXL ions cold at wall !never “see” wall", few eV
IXL ions hot at grid but sputtered atoms removed
by Gig = 10#20x recirculation factor # never reach
core
Fusion product sputtering from outer !envelope"
wall, removed by:
• Grid collisions, Gig in IXL
• Outer wall collisions in EXL,
inter#polar collector plates, if needed
Out#surface generated high#Z atoms cannot ac#
cumulate in core
Figure 70 — Summary of effect of high-Z impurity atom
generation in IEF systems
Ash Buildup/Removal - I
Ash Production: nas ! bij n 2f , f v f k f
N
- *r 0
Trapping Probability: pt 5 N 6 ! / c 2 ! rc n f , ash
. 2 9ac 1
5
6
N
- * rc n f , ash 0
Ash Buildup: nat ! nas pt 5 N 6 t ! bij n 2f , fus v f k f /
21 t
.
2
N
- * rc n f , ash 0
n
or at ! bij n f , fus v f k f /
21 t
.
nf
2
5
6
N
Figure 71 — Helium “ash” production; trapping and buildup
in IEF systems
Ash Buildup/Removal - II
Slowing down: N min
-E 0
ln / a 2
. Ef 1
!
; let Ea ! 2 MeV, E f ! 50 keV
-E 0
ln / 1 2
. E2 1 max
-E 0
- 40 0
for 4 He on D; / 1 2 ! / 2 ! 0.95
. E2 1 max . 49 1 max
-E 0
- 40 0
for 3 He, T; / 1 2 ! / 2 ! 1.08
. E2 1 max . 9 1 max
Take rc ! 2 cm; R ! 350 cm; v f ! 7 x10 9 cm/sec
, fus ! , ash ! 0.16 b, K f ! 1.5, n f ! 1019 / cm 3 , then
5
nat
! 5 x1013 10 "6
nf
6
N
t; assune N ! 1, then
nat
! 5 x10 7 t
nf
- 2R 0
nat
"4
"3
now if Gi ! 10 3; C i ! Gi /
2 ! 10 sec, and n ! 5 x10
v
. f 1
f
•
No signicant ash buildup is possible
Figure 72 — Helium “ash” collisional slowing-down, removal
rate and equilibrium level in IEF
This level of ash buildup can have virtually no e$ect on
system operation. Figures 71 and 72 show calculations of
this trapping probability, the ash production and buildup
rates, and the collisional slowing#down and system re#
moval rates, and nal equilibrium density for a typical
IEF system.
C. System Safety
All IEF systems are inherently safe with respect to “run#
away” power excursions; these simply can not happen, as
the plasma will merely “go out” if any signicant devia#
tion from operation at design conditions is experienced
!see later discussion of control". This is a direct result of
the fact that these systems are not % and indeed, can
not become % critical, as in a ssion reactor, or reach
ignition, as in a Maxwellian fusion system of the conven#
tional type. They are all simply power ampliers, driven
by electrical input, and producing much more output
than input. If the input falters, the system will stop
without any means of self#destruction. However, exoge#
nous sources of failure may cause an element of the sys#
tem to fail, such as an earthquake, falling objects, struc#
tural failure of a vacuum shell, etc. If such events occur,
they could rupture the system and lead to dumping of its
contents, or of its stored energies, or to dispersal of the
unavoidable radioactive isotopes that have been formed
in its structures by neutron interaction !for the DD#
halfcat cycle" over their life in the system. It is thus of
some interest to estimate the magnitude of these poten#
tial hazard sources, and compare them with other com#
monly known hazards in nuclear power plants, to esti#
mate their threat level.
The most obvious hazard is that of the radioactive tri#
ton gas produced directly in the T branch of the DD
reaction that drives the DD#half#cat cycle. This will all
be extracted from the vacuum system by direct cryo#
genic separation from the residual He gas components,
in order to assure that no tritium is recycled back into
the IEF system, where it would otherwise react with in#
situ D and produce only
undesirable DT neutrons. The extracted tritium will be
stored in a safe place or shipped to another site for use
as may be required. The level of tritium required to be
so stored is about 2 MCi per MW of fusion power pro#
duction. While this is a large amount, it is very easy to
store, cool and protect as tritiated water. From this wa#
ter storage it is also easy to recover the fusion fuel 3He
gas which is the product of the tritium beta#decay over
its 12.7 year half#life. The amount of tritium present i!
the IEF power system will then be determined almost
entirely by its once through density in the vacuum sys#
tem and its transient density from production in the
IEF fusion device, itself. It is almost impossible to esti#
mate the former amount but, since it must be fully#
extracted as it passes through the vacuum system, the
amount must be small. The amount found in the IEF
device can be estimated from the known density and
45
fusion rate distribution; this is found to be in the range
of 5!10 ngm "0.5!1 x 10!8 gm# for a typical DD!halt!cat
system. This is only about 100 micro!curies, roughly 1/10
of the maximum permissible human whole!body inven!
tory. This is in considerable contrast to the tritium in!
ventory found in large M&M systems, which may run
into many tens of grams.
Other radioactive materials will be found in neutron!
activated structures. However, since the neutron pro!
duction rate is only about twice that of a PWR and
since no e$ort is sought to conserve neutrons "as for
PWR criticality#, the structural activation of metals in
the system is expected to be no greater than the yield of
radioisotopes from structural radiation found in a PWR
system. External activation in the blanket can be made
to be much less, in fact, because the blanket function is
principally to remov! the neutrons "ithou# creating
active species. Furthermore, the IEF unit itself can be
removed and recycled at any time, thus the radioiso!
topes species can be removed and a clean new unit be
installed.
Radioisotopes Inventory
•
•
•
•
•
•
IEF/FPC units recycled out at end!of!life
Removes radioisotopes from in!plant inventory
Principal inventory is blanket/shield
Number of radioisotopes atoms ! neutron fraction
captured in n,y reactions in blanket
10% of neutrons captured in structure, 90% in
10B
Relatively small volumes of FPC and blanket ma!
terial minimize volume of active material pro!
duced
Figure 73 — Qualitative considerations of radioisotope inventory in IEF systems
By all of these means, the radioisotopes inventory in an
IEF plant can be kept well below that of the activated
materials in a PWR plant and, of course, there are NO
ssion products in the IEF plant, at all. Figure 73 sum!
marizes these qualitative considerations of radioisotopes
inventory in IEF systems. Quantitative estimates re!
quire studies for particular, specic plants and design
conditions, and these have been beyond the scope of
this work. A more detailed discussion of the issue of
neutron capture!removal and structure activation is
given in a paper in Appendix B, herewith, prepared after
the conclusion of this e$ort.
have a recirculating electron power 10,000 or more
times the fusion power, but its energy content is only
that over a single electron transit time of about 60 nsec.
In such a system, with Gj = 5 x 105, for example, the total
particle lifetime will be 30 msec or so.
For a typical system with 100 MWe injection power, this
means that the maximum stored energy can be only
about 3 MJ. If all of this could be dumped instantane!
ously into only 1% of the system structure mass, its
temperature rise would be only about 30!40°C; but there
is no known way to achieve such a localized dump, as
IEF systems do not su$er from the sort of “disruptions”
that are found in the powerful drive currents of toka!
maks. Figures 74a,b show details of these calculations for
a 2000 MWfus system with 100 MWe injection power,
at typical ion and electron recirculation ratios.
IEF System Internal Energy Storage and Power I
Precirc = Pdriv!"Gi,j#; Gi = ion recirculation ratio "IXL#, Gj =
electron recirculation ratio "EXL#
Estored = Precirc #trans$i,j%; #trans$i,j% = 3R/&i,j
Typical values: Gi = 104; Gj = 5 x 105,
#transi = 3 x 10!6 sec, #transj = 6 x 10!8 sec
Yield: "Gi,j #trans$i,j%# = !lif! = 3 x 10!2 sec
Figure 74a — Ion and electron recirculation and lifetime in
IEF systems
Finally, there is the question of the dumping of stored
magnetic eld energy in systems that use superconduct!
ing magnets. This can happen if the superconducting
loses cooling, heats up and “goes normal” so that it no
longer super!conducts. In this situation, the free!owing
magnet current that supports the B eld will decay by
ohmic heating of the conductor material in the region in
which it has lost cooling and now has a nite resistivity.
As this happens, the conductor heats up and propagates
the failure condition along the conductor material. This
further raises the net resistance of the conductor and
increases the rate of ohmic heating, and the entire proc!
ess exponentiates with considerable speed. If the B
Another potential hazard element in IEF systems is that
of stored energy, both in the B eld of EXL systems and
in the plasma energy found in the recirculating ion and
electron streams. The recirculating power levels in IEF
machines are very large, however so little particle mass
is found to make up these currents that the total energy
involved is quite small. An operating IEF system may
46
IEF Systems - Internal Energy and Power II
Example: Pdriv! = 102 MWe, Pfus = 2 x 103 MWth
Estored = 102 x 10!2 = 3 MJ "0.7 kg high explosive#
•
If dumped into 1$ of IEF structure, the resulting
temperature rise is ca. 30!40 oC
Precirc! = 102 x 104 = 106 MW
Pfus/Precirci = 4 x 10!5
•
Pfus"total# is only a very small disturbance to Precirc,
thus contributes little power/energy to drive in!
stabilities
Figure 74b — Stored energy and power in recirculating ions
and electrons in IEF systems
eld energy is su%cient, it can overheat and vaporize
the superconducting coil. And, if very large, can lead to
explosive destruction of the superconducting coil in the
region around the initial failure, before the thermal wave
has propagated very far from its initial site. This prob!
lem has been addressed in this work, with the result that
IEF systems are found to have only small B eld energy
storage, relative to that found in large M&M/LTE to!
kamak machines. This is due largely to their much
smaller elds as well as the very much smaller volumes
occupied by these elds.
Hazard Potential in IEC/FPC
•
•
•
•
•
4 MJ = 1 kg high explosive
Plasma energy ca. 3!10 MJ
Tritium inventory "FPC# ca. 5!10 nano!grams
B eld energy "EXL# ca. 100 MJ
Hazard 100!1000x below large machine systems
Figure 75 — Summary of selected hazard potentials in IEC/
IEF systems
More detailed calculations of this problem are given in
the paper included herewith in Appendix B. Generally,
results show that IEF systems will have stored B eld
energy in the range of 100!1000 MJ. Large!scale conven!
tional M&M/LTE systems "e.g. the ITER tokamak# may
have as much as 0.25!1 x 106 MJ; up to 1,000!10,000
times the energy storage in IEF systems. As an aside, it
is worth noting that such a system can & if its magnet
cooling fails & release up to 250 tons equivalent explo!
sive yield by vaporization of a fraction of the supercon!
ducting coils in the magnet system. Since this will inevi!
tably involve a large amount of neutron!activated radioi!
sotope material, it represents a hazard of high order in!
deed. In contrast, the IEF stored eld energy can
scarcely heat the magnet material above a few hundred
degrees. And, if normal conductors are used the entire
problem vanishes. In general, as summarized in Figure
75, it is found that IEF system hazard potential from the
sources discussed above is less than that from PWR "in!
cluding ssion products# and large M&M fusion ma!
chines by two to three orders of magnitude.
D. Neutron Production and Materials
Damage
As already discussed, the neutrons produced in the DD!
half!cat cycle are all of relatively low energy, comparable
to those found in ssion, and very much below those
found in the DT fusion process. This has the conse!
quence that the damage to materials that can be done by
these neutrons is comparable to that from ssion neu!
trons and very much less than from fast DT neutrons.
Figure 76 shows this comparison for displacement dam!
age in niobium metal and for helium gas production by
"n,!# reactions in various metals. The energy distribution
of neutrons from ssion is sketched in on the right hand
chart. Note the overlap of DD!half!cat cycle neutrons
with ssion neutron mean energy.
This comparison is extended in Figure 77, which shows
the relative materials damage potential per unit source
power for neutrons from ssion, the 3He fusion reaction
"the neutrons arise from unavoidable DD side reactions
because of the D present to react with the 3He; the
numbers cited in the gure are for an optimized 3He!
rich mixture#, the DD!half!cat reaction "here labeled as
“DD”#, and the DT fusion reaction. The great di'erence
between the DT reaction and the others arises because
the DT neutron is so energetic "14.1 MeV# that it can,
and will, generate more neutrons by "n,2n# reactions in
its rst collisions. Furthermore, a DT system must be
designed so as to conserve the single neutron produced
in the fusion process, in order to breed the tritium
needed to sustain the process itself. Thus, design choice
of materials in the blankets of DT systems can not be
made for the purpose of disposing of the neutrons with
minimum activation; this must be secondary to their
conservation to breed tritium.
47
Neutron Displacement Damage
DT
2250
150
DT
130
2000
110
1750
1500
Fission
1250
DD/
D3He
1000
Total
(n,2n)
750
Inelastic
500
Elastic
250
0
(n,p)=10
(n,!) cross-section in barns
Displacement cross-section in barns
2500
Fast Neutron Helium Production
Fission
90
Al
DD/D3He
70
316SS
50
Ti
30
Mo
(n,e)=10
2 4 6 8 10 12 14 16 18
Neutron Energy in MeV
Figure 76a — Comparison of neutron displacement damage
and helium production in metals by fast neutrons from various sources; note DD-half-cat neutrons similar to ssion
neutrons
In the DD!cat system, this is not the case, and the ma!
terials choice can be made for minimization of activa!
tion; no e"ort is needed to provide conservation in scat!
tering collisions. Taking these factors into account leads
to the relative weights of damage potential shown in the
gure. Note that the DD!cat system is almost as good as
the D3He and ssion system, and 10x better than DT.
10
V
8 10 12 14
6
2
4
Neutron Energy in MeV
Figure 76b — Comparison of neutron displacement damage
and helium production in metals by fast neutrons from various sources; note DD-half-cat neutrons similar to ssion
neutrons
Relative Fast Neutron Damage Potential
Fission D3He
DD
DT
Average neutron
energy
2.0 MeV
2.45
MeV
2.45
MeV
14.1
MeV
Neutron power
fraction
0.03
0.02
0.09
0.65
Relative materi!
als damage
potential per
unit power
1
1
2
20
Figure 77 — Relative fast neutron damage potential from
various neutron sources
48
IEF System Startup - IXL
E. System Control and Stability
All IEF systems are easy to control, given means of con!
trolling ion and electron current inputs and voltages.
These involve the development of ion and electron
sources whose output can be controlled by external elec!
trical means, to achieve desired operating states within
the bounds of limiting emission or ionization physics.
With these, control of the complete system is straight!
forward because thermal power system time constants
are all very long compared to the ion and electron life!
time and ion transit time constants in the IEF plasma
system. Typically as shown in Figure 74a, these latter
may be in the range of 0.03 sec and 3 x 10!6 sec, respec!
tively, while the thermal time constants of the system
may be 1!5 seconds for the IEF unit and 100!500 sec!
onds for the blanket of a DD!half!cat system, as shown
in Figure 78. In this circumstance it is simple to devise
system controls for close load!following, and for smooth
startup and shutdown.
System Control - IXL/EXL
Load following, startup and shutdown are simple "
limiting thermal time constants are:
•
•
Sequence:
1. Inner accelerating grid cooling on
2. Apply full accelerating voltage to grid
3. Turn on low voltage ion guns to initial level
4. Turn on external e! supply guns
5. Ramp up ion current on thermal time scale to
make virtual anode
6. Turn o# e! guns and control e! input by thermi!
onic grid emission after startup
Fueling:
1. Pump system to low background to avoid charge
exchange recirculation
2. Use external injection guns to supply ions and
electrons until in!situ collision
3. Switch to neutral gas fueling in power range
Figure 79 — Startup sequence required for lXL/IEF systems
for optimum operation
IEF System Startup - EXL
IEF/FPC unit, !FPC = 1.0 second
Blanket, !BLANKET = 100 seconds
Sequence:
Both are long CF. to !E, !B in IEF systems
1. B eld on to initial minimum value, Bmi!
Thus all drive controls can be very slow relative to
IEF fusion reaction control times
2. I" on at $a% initial WB condition !K$W%E0.5, $b% at
1/3 to 1/2 design voltage, E0
Figure 78 — Basic system control considerations for IEC/
IEF systems
Control during startup is a matter of applying the
proper sequence of operations to the system. This is
di#erent for the two types of IEC/IEF. The IXL system
can be turned on and driven to operating conditions
directly, while the EXL system must be started well be!
low the desired end state, and moved to this point along
the rb1 line $or close to it% to minimize drive current re!
quirements. Figures 79 and 80 summarize these se!
quences for the two approaches.
3. Ramp B up on timescale long CF. !" life
4. Increase E ➝ E0, and decrease # ➝ #mi! as W ➝
Wmi! $design point%
Fueling:
1. Start on preset background gas, with e!guns at
initial voltage
2. Transition to low!voltage i+ guns as background
burned out by e! guns and core density increases
3. turn o# i+ guns as W ➝ 100!10 x Wmi!, run on
neutral ion gas input
Figure 80 — Startup sequence required for EXL/IEF systems for optimum operation along rb1 line in WB mode; Eo,
B and Ie can all be varied to achieve minimum power consumption
Analysis of the dynamics of control and the control sys!
tem transfer function $here denoted by G; for power
“gain”; G = Pfus/Pinj = Pf/Pi!% of IXL systems shows that
this depends on the energy!dependence of the fusion
reaction rate per fuel ion. This is simply the product of
the ion collision speed and fusion cross!section, thus
depends strongly on the slope of the cross!section curve
with energy, d!fus/dE. In order to attain a zero!order es!
49
timate of the transfer function behavior it is convenient
to specify the cross!section energy!dependence as a
simple exponential form, !fus"E# = !fo"E#S; so that with
this form "1/!fus#"d!fus/dE# = dLN!fus/dE = S. The fuels DD
and p6Li have exponents of 1!2 in the general energy
range of most usual interest for IEF systems, while DT,
D3He, and p11B are much more strongly energy!
dependent, with S = 3!5.
dimensional parameter space determined by the global
design parameters W = E0"BoR#2 and Z = 8%r!crc2. Be!
cause of the multi!parametric nature of this system, the
electron connement physics thus inexorably couples
the parameters E, B, and the fusion power generation
and injection power of the system. The gure given pre!
viously "Figure 39, reproduced here# showed the varia!
tion of Gj with these parameters.
This di$erence is found to reect itself in the gain
transfer function for system control of the IXL system
in such a way that the initial scaling of G at small power
"i.e. at startup# for the low!S fuels is simply proportional
to the input power G ! Pi!, while that for large!S fuels
varies also as the square of the drive voltage, G ! Pi!E2.
At high power, with strong ion trapping in the system
"i.e. when the well depth has been fully formed; it starts
outt with no e$ective well depth# and saturated ion and
electron losses "e.g. by collisional upscattering, grid colli!
sions, cusp escape, or other means#, the gain transfer
function becomes less dependent on the system energy,
so that G ! Pi!/E for S = 1!2 and G ! Pi!E for S ! 3!5.
This behavior means that the drive system control gain
will change as the system is started from zero initial well
depth and density to full operating conditions, about as
indicated in the G vs. Pi! curves of Figure 81, for the
IXL system. This sort of behavior is, of course, com!
pletely stable over the entire operating range.
3D Plot of EXL System Operation
IXL System Control
Decoupled "no feedback# power amplier, Pf =
G"Pi!, E# Pi!
At low power "startup#, with !c small,
G"DD, p6Li# = Pi!; G"DT, D3He, p11B# = Pi!E2
At high power "operation#, with ion trapping, losses:
G"DD, p6Li# = Pi!/E; G"DT, D3He, p11B# = Pi!E
E3
DT, D3He, p11B
E2
E1
E1
E2
E3
G
DD, p6Li
Pin
FigFigure 81 — IXL system gain control transfer function,
showing functional dependence on power and drive voltage
for fuels with differing cross-section energy-dependence
107
kL = 2, " = 0.
9,
r
"q = 0.995, #
= 0.995,
m = 3, N =
NMR = 8
WB
106
5
10
Gj
104
107
106
105
104
3
10
102
103
<rb> = 1
Gj
102
1
10
101
1020 19
10 18
Z= 10 10171016
8!
n 1015
c
rc
2
(1/
1014
cm
1013
)
101
1
10-1
10
-3
10-4
10-5
10-6
2
2R
E 0/B 20 cm2 )
=
W kG
/
(keV
10-2
Figure 39 — Three-dimensional plot of EXL system operation over the parameter range of Gj, Z and W. Note that
transition from MR behavior on the at, lower-right-hand
plane to growing WB operation on the slope, must occur
through a “valley” in the plot surface
As already explained, the desirable means of startup for
EXL is to follow the rb1 WB “ridge” line in this parame!
ter space, to minimize drive power required. Startup
along the rb1 WB line forces .the gain transfer function
to vary approximately linearly with W, thus dGj/dW =
constant "but see below#. The total derivative of fusion
power generation with respect to W "dPf/dW# then de!
termines the control system gain transfer function in
terms of both B and E. Since the optimum method of
startup involves variation of both B and E, this is neces!
sary if the actual dependence of the gain transfer func!
tion on real design parameters is to be found, from G =
"dPf/dW# = ""Pf/"B#""B/"W#"E + ""Pf/"E#""E/"W#"B.
Holding E xed "rst term above#, the partial derivatives
of Pf show no dependence on system cross!section
energy!dependence, but variation of E while holding B
xed "second term# does give such dependence. These
partial derivative terms are summarized in Figure 82 for
the EXL system. There is yet another complexity to
consider in this system.
In the EXL system, startup operation is considerably
more complex to analyze, because the system power gain
is set entirely by electron "rather than ion# losses and the
parameters that may be accessed cover a three!
50
EXL System Control I
•
•
EXL System Control II
Electron connement physics couples Pf, E, B, Pinj
Interactive control in system parameter space
W!
E0
2
2 , G j , Z ! 8* nc rc
5 BR 6
E S ) 0.5
! E S "1.5 B 4 R 4 ; (, f ! , f 0 E S )
W2
E Pf
- P0
- P 0 E Pf
! 5 S " 1.5 6 / 2
! "2 / 2 ;
.W 1
. W 1 EW B
EW E
Pf D
E Pf
EB
E
- P0
! 4/ 2;
.W 1
E Pf
EE
B
- P0
! 5 S " 1.5 6 / 2
.W 1
System inherently stable over range
FW
! 5.0 Below design point at W ! Wmin
W
Figure 82 — Partial derivatives in gain transfer function of
EXL system, showing dependence on the parameters W, E,
B, and fusion cross-section energy dependence (exponent)
This arises because the cusp losses of electrons !which
determine gain" will begin to be reduced by electrostatic
reection back through the cusp loss cone from the
guns/grids that supply the electrons initially into the
system, when the cusp loss hole size approaches that of
the electron#emitting surface from the injection guns or
emitters. This happens as the B eld is raised towards its
nal design operating point. The fractional electron
losses then will be reduced to a value less than for opera#
tion far from this condition by a factor !R determined
by the internal reectance of the e#guns. Typically, !R $
0.1, giving an increase in gain of up to 10x as the nal
design point is approached. In terms of the parameter
W this means that as W ➝ Wmi! !the design point W",
the gain will rise more rapidly than linearly with system
power !drive or fusion". This has the happy consequence
that stability will be assured over a considerable range of
operation !i.e. drive current" below that required to be
“on” the WB line. For, note that the system has a basic
instability if operation is attempted just on this line.
Since the drive current required to be just o" this line,
but below it !in the 3#d parameter space this means be#
ing on the sloping region stretching down to the MR
region plane", is greater than that to be on the line, any
fallo% from the exact W path will result in an unstable
runaway down to very small power levels !well below the
Z value point at which dGj/dZ = 1".
Design Point
Control by changing
e-injection drive
voltage, E0, or B field
drive current
Startup Range
LN(Gj)
Power Control
Range (5X)
LN(Z)
12
10
Figure 83 — Effect of e-gun reectance on WB-line dependence of Gj as design operating point is approached in
EXL system; note increased slope over nal 10x increase in
W
While this does not pose any safety or hazard problem
!the system can not run away to higher power" it is a nui#
sance and requires drive operation during startup that is
marginally higher than that actually needed for WB op#
eration. All such excess drive power is simply wasted,
thus it is fortunate that the gun reection e%ect de#
scribed above exists to allow exact operation a! the full
power design point. The major features of this are
shown in Figure 83, that shows this e%ect. Note that
operation in the power control range is inherently stable
over a variation of W such that "W = 5W can be handled
without reaching the “fallo% ’ instability just described.
As the EXL system is driven on the WB line it is impor#
tant to note that this line may take the WB internal ra#
dius to the critical radius r#, but no further with stability.
Driving with currents appropriate to higher W values to
reach rb1 at r = R will not yield any greater power output.
51
EXL System Control - III
Design Point
1.0
1.0
Ggr
0.5 Pf
Ggr 0.5
Pf
(a)
(b)
0
•
•
0
0.5
Wmin
W
1.0
!!stable power balance; r! < r < R
Within "R "W/Wmi"# e$ect
Figure 84 — Stability region around full power design point
in EXL system, showing both beta stability in rk < r < R and
gun reectance effect with approach to design point operation
The system is therefore stable over the range of W cor!
responding to rb variation from r! to R. This is not a
large range, because r! may be only about 0.93R at the
fully!diamagnetic operating state. Nevertheless, this
provides another stable region of operation in which
variations in W can not result in system unstable shut!
down. The e$ect of these two phenomena "r! ➝ R and
"R# is to broaden the region of W!space in which the
system is inherently stable around its full power design
point. Figure 84 shows this region and the way in which
both G and Pf vary as Wmi"/W is varied. Thus, operation
at full power can be held stably down to 10!20% of full
power by no more complex controls than simple varia!
tion of the drive voltage or drive current over a 5x range.
Below this, the B eld must also be varied to maintain
W along the WB line. This is easy to do but not neces!
sary for full power load!following control over a range 5!
10x: below the full!power design point in the EXL sys!
tem.
52
7. Development and
Deployment
A. Critical Physics Issues.
As mentioned earlier, each of the two IEF approaches
has its own critical physics issues that must be resolved
to determine if that approach will eventually prove to
have merit for practical power systems.
Critical Physics Determine IXL Feasibility
IXL needs:
•
•
Ion core trapping or
Electrostatic wave scattering
Either e!ect allows:
•
•
Small machines
D"fuels only
Figure 85 — Major critical physics issue requiring resolution
for IXL system
Critical Physics Determine EXL Feasibility
EXL needs:
•
Diamagnetic electron ow
This allows:
•
•
Small machines using D"fuels
Mid to large machines using p"elds
Figure 86 — Major critical physics issue requiring resolution
for EXL system
In the IXL system, the critical issue is how to raise the
core ion density while maintaining low losses and well
stability. In the EXL system, it is the essential require"
ment that the electrons behave diamagnetically so as to
restrict their losses to those from simple cusp escape
holes in the surface B eld of the system. These are
summarized in Figures 85 and 86, which indicate the
nature of the nal fusion power systems that could be
attained if these issues are resolved favorably. Each of
these has been analyzed, with the results previously
cited; neither has been unequivocally “proven” from the
analyses and computer simulations. In reality, neither
can be so proven; this can be done only by experiments
conducted to study these issues directly. Thus the most
essential steps to take to test these concepts must in"
volve such experiments.
For the IXL system, this means rst devising tests to
determine if electron two"stream instability waves actu"
ally are or are not generated at high electron currents. If
they are, then the next critical issue is the ability of ion"
acoustic or other electrostatic wave phenomena to pro"
duce ion wave"group"trapping in the core. If this is the
path followed and WGT/ICC proves not possible, the
IXL system would seem to be hopeless for useful fusion
#although may be of interest for study of core physics or
for low level neutron source generation$. For the EXL
system, the most important issue #indeed, almost the
only important issue$ is that of electron diamagnetic
WB behavior. If WB behavior can not be achieved, then
the system is reduced to that faced by the IXL concept
and must rely on WGT/ICC core trapping for success.
If WB behavior proves feasible, then all succeeding
physics issues are important to limiting system perform"
ance but none can render the concept unworkable.
53
IXL Physics Decision Tree
End
N
ICC
Effect
Y
IXL
Mantle
2-Stream
<rc>
Small
Y
Classical
Trapping
N
End
N
Y
N
Small, LTE
D-Only
Machines
End
Figure 87 — Decision “tree” for IXL system, showing major physics issues that must be resolved for useful fusion power systems; note system always limited to D-only fuels
EXL Physics Decision Tree
N
N
!e
high
N
Y
EXL
End
ICC
Effect
WB
Diamagnetic
N
Y
N
End
Small, LTE
D-Only
Machines
Y
<rc>
Small
Y
N
<rc>
Small
Y
Small-size
non-LTE Machines
D-Fuels Only
N
End
Y
Mid, Large Size
Machines
All Fuels (p11B)
ICC
Effect
!e
low
Y
End
<rc>
Small
Figure 88 — Decision “tree” for EXL system, showing major physics issues that must be resolved for useful fusion power systems; note multiplicity of paths to success
Figures 87 and 88 show very simple R&D decision
“trees” for the two concepts. Note that the achievement
of small core convergence ratios is shown as essential to
both systems. Later work suggests that these devices can
be made to yield net power even if the core convergence
is not so small as previously thought; it now appears that
<rc> as large as 0.3 may still be able to give net fusion
power, at least from DT fuels. DD fuels may require
smaller < rc> perhaps as small as 0.1, but this is very
much larger than the 0.01 and less required for practical
use of p11B, for example.
IEF in Electric Power Plants"
B. R&D Paths and Programs;
Proof-of-Feasibility
These next steps in understanding of IEF concepts do
not require large programs, long times, or large sums of
money. The critical physics issues can all be tested at
quite modest scale and cost and, consequently, need not
take excessive time. The next step studies could be con!
ducted with systems whose radius is less than one meter,
for example, with drive power of only a few megawatts
at currents of a few hundred amperes.
54
Next Step Critical Physics Studies I
WB Diamagnetic Electron Flow in EXL
He/D diluted or DD, R = 50 cm, <rc> = 10!2, !a = 0.2
I! = 500 A at E0 = 4 keV, P! = 2 MW steady!state
Truncated cube system, 6 e! guns, Zg = 50 ohms
Wmax = 15; Bmi" = 10.3 G; "mi" = 1.5 x 1013/cm3
Wmi" = 4 x 10!3; Bmax = 0.63 kG; "max = 6.8 x 1016/cm3
•
e! gun development is critical
Figure 89 — Critical physics experimental system for test of
EXL concept
Figures 89 and 90 show an example of such experiments,
designed to test the critical physics in IXL and EXL
systems. Note that even these small tests can show en!
hanced performance of 100!1000x by the relevant phys!
ics of the IEF concepts, above that expected if these
concepts do not work.
Next Step Critical Physics Studies II
Wave Group Trapping of Core Ions in IXL
He/D diluted or DD, R = R1 = 30 cm, R2 = 50 cm
Ii = 500 A at Ei = 4 keV, P! = 2 MW steady!state
Dual spherical grids, Gi = 10; <rc> = 10!2, !a = 0.5
"c = 2.5 x 1014/cm3, "0 = 3.5 x 1016/cm3 "with ICC#
Grid bucking voltage and heating control are critical
Figure 90 — Critical physics experimental system for test of
IXL concept
The content of these experiments and their general
goals are summarized in Figure 91, that also gives an es!
timate of the time and cost for their conduct. This is, at
most, $12M over 3 years, or $4M/year; about 1% of the
current U.S. magnetic fusion budget. Since the payo& of
success here is so great, it seems obvious that this small
investment should be undertaken, especially when
viewed against the highly!improbable and very!long!
term future faced by the impractical non!economic
characteristics of the huge tokamaks that currently form
the sole basis of U.S. "and most of the world# fusion re!
search. Small IEF machines of inherently low!cost and
modest power could, if successful, be developed many
decades sooner than these M&M/LTE behemoths that,
so far, seem to be of little interest to any practical utility
system application.
Next Step Goals, Activities, Costs and Time
EXL
Develop e! guns
Build truncated cube
system
IXL
Develop/build grids
Build dual grid system
Build, assemble, setup power supplies, diagnostics
Experimental development/testing of system controls
Conduct critical physics experiments
Test WB/Diamagnetic
ion ow, mantle electron
core/edge collisions
Test "ICC# wave group
trapping, mantle electron
two steam instability,
potential well stability
Time = 2!3 years, cost $8!12M "total#
Figure 91 — Goals, activities, costs and time scales of the
next step small-scale experimental research required for
IEF systems proof-of-principle research studies
The nature of these two approaches "M&M/LTE vis a
vis the IEC concepts# is such that they require funda#
menta$y very di%eren& development paths. The very large
M&M tokamak systems su&er from the fact that they
cannot be described ab initio by classical physical models
or theories; all research and study of these must be done
be a succession of scaled experiments, ever larger in size
and ever more costly. This is because the critical issue
for tokamaks is that of transport losses of the neutral
LTE plasma across the high magnetic elds used for
their connement. This transport loss is a collisionally!
dominated anomalous process, and can be determined
only by experiment. Loss rates "i.e. di&usion or trans!
port coe'cients# are so large that experiments at pro!
jected nal sizes could not and can not now be projected
with any degree of accuracy. This necessitated studies at
small scale, learning transport coe'cients that could
only be extrapolated to the next larger scale device, at
which time the corrected transport coe'cients and scal!
ing could then be extrapolated to the nex& larger device,
ad innitum. Development of this sort then is based on
scaling laws and ever larger test devices. Since the cost
of these goes roughly as the cube of their dimensions,
the cost and time to reach any nal useful result "and the
existence of such a “useful” result is not even clear# be!
comes exponentially larger with continuing research.
This is summarized in Figure 92, that also shows the
comparative R&D approach available to IEC systems.
55
Fundamental Differences
Between Approaches
IEC
M&M
Concept feasibility
depends on
Physics Issues
Scaling Laws
Requires experimental
testing at
Projected final size
cost ! fixed; R ! small
Ever larger size
cost ! R3; R increasing
Figure 92 — Fundamental differences between R&D approaches for IEe/IEF and M&MIL TE fusion power systems;
conventional approach requires ever-larger scaling studies
of empirical physics; IEC needs only small-scale research
of classical phenomena
In striking contrast to this situation, the IEC ap!
proaches can both be tested at small, xed scale. They
are both governed by classical physics, in simple geome!
tries but over parametric conditions that make for com!
plex physics " but still only classical physics, not em!
pirical scaling laws " not governed by instability or col!
lisional transport phenomena.
C. Long-Term Development to Commercial Power; Strategy and Schedules
Because there is no need to conduct R&D over ever!
larger sizes of experiments to test or prove scaling be!
havior of IEC systems, it has been found possible to
devise an elementary and simple plan for fusion power
development. This uses a straightforward R&D program
to full!scale power plants without changing the size or
driving conditions of the lEF devices from those used at
the beginning of the research e#ort. Analyses of DD!
half!cat $and full!cat% cycles done under the auspices of
others $NASA/SDIO% have shown the size and cost of
such a development program to be very much less than
that projected for large tokamak M&M/LTE systems. A
summary of this work is included herewith in Appendix
B; this was prepared for presentation at the Second
Wisconsin Symposium on 3He and Fusion Power, to be
held at the University of Wisconsin in July, 1993. This
shows that a system with radius 3 m and B eld about 14
kG can be used to develop DD!half!cat power, starting
with research on HD and DD alone $all at the same
size%, in about 12 years at a cost of less than &1B, including
a full scale prototype demonstration power plant at
about 600 MWe. Subsequent commercial plant devel!
opment could be accomplished easily within another 6!8
years.
A strategy for IEC/IEF fusion power development and
deployment can then be adopted that ts this rather
short time scale, starting with the demo plant R&D de!
scribed above, and leading to a subsequent commercial
DD!half!cat fusion plant based on thermal $steam% con!
version as its rst deployable product system. Each of
these steps could take as much as ten years, thus com!
mercial fusion power could certainly be available within
20 years in the utilities market. Beyond this time, rapid
deployment of such plants should take place, simply
because they will be cheaper, environmentally benign,
and easier to build and maintain than any other energy
source. Construction of nuclear ssion plants should
cease, as should that of fossil!fuel!red systems. Nuclear
ssion product waste accumulation should end, as
should further additions to greenhouse gases in the at!
mosphere $at least from power plants%.
On a longer time scale, the path for further R&D and
deployment is clear, given success with the rst of these
IEC/IEF systems. This is suggested in Figures 93a,b,c,
which outline the strategy, schedule and logic of con!
tinuing research, development and deployment over a
one hundred year time scale. With a burgeoning DD!
half!cat thermal cycle economy, the logical next step is
R&D of the full!cat $T ➝ 3He catalyzed% system to up!
grade initial half!cat systems. In parallel, R&D on clean
fuels can be undertaken, leading to development of p11B
systems in another decade or so. And, the development
of direct electric conversion should continue along with
both of these e#orts, so that the DD!cat systems can be
upgraded by this means to give larger net power with
still less thermal pollution.
IEF Fusion Power Development Strategy
•
•
R&D to prove IEF physics, select system type
$IXL/EXL%
Develop/deploy power plants
$1% Simplest thermal conversion $TC% $DD 1/2
Cat%
$2% Partial direct conversion $DC% upgrade for
new plants $DC 1/2 cat%
$3% Simplest all DC system $p11B%
$4%Advanced fuel partial DC upgrade $D3He%
$5% Advanced fuel all!DC $D3He, 3He3He%
Figure 93a — Mid- to long-term IEF fusion power development strategy; from R&D on commercial DD-half-cat thermal power plants, to direct conversion plants and clean
fuels
56
IEF Fusion Power Development Strategy
DD
TC
(1)
Partial
DC
(2)
DD
(4)
p11B
DC
D3He
He3He
3
(5)
(3)
0
10
20
30
50
40
60
Calendar Years
70
80
90
100
Plant Development
Retrofit/New Plant Development
Figure 93b — Schedule of mid/long-term RD&D for IEC/IEF
power; from DD-half-cat to DD-full-cat, to p11B and 3He3He;
and from thermal to direct conversion systems
Eventually, clean p11B plants can begin to be deployed,
using direct conversion, to supply new power demands
and replace aging DD!cat plants as these reach their
economic!end!of!life. And, nally, work can be under!
taken on clean 3He3He fueled systems, against the day
when the 3He resources of space "Earth’s Moon, but #
much more plentiful # the atmosphere of Jupiter$ be!
come available for use on Earth and elsewhere in the
solar system. The gures show general time scales for
such long!term developments and deployments, all of
which are possible if the rst one can be made to work.
IEF Fusion Power Development Strategy I
•
•
•
•
•
•
Uses lowest cost available fuels "DD$
No DT burning, minimum neutrons "% ssion
spectrum$
Starts simply, no exotic cycles "DD 1/2 cat, TC$
Converts towards direct conversion on plant up!
grade time scales
Employs advanced available fuels "p11B$ only from
prior development experience
Transition to more advanced fuels and cycles
"D3He, 3He3He$ as needed and available from base
of full plant/system experience.
Figure 93c — Logic of mid-long-term IEF/IEC research,
development and deployment program
57
8. Conclusions and
Recommendations
A. The Promise of IEF Systems.
The overwhelming promise of IEF systems almost needs
no elaboration and, indeed, was the principal nding of
the EPRI IEF Review Panel reported in Appendix A. In
this the Panel said that, “... the IEF approach to fusion
power is more promising than any other fusion concept ye!
studied ...,” if the critical physics can be proven e!ective.
In considering this promise and potential it is of utmost
importance to grasp the fact that these systems are no!
like other approaches to fusion, either in their physics,
or their R&D requirements, or in their stunning eco"
nomic and operational potentials. A" fusion systems co#$
cepts ar% no! alik%. The large scale paths of the main
U.S. governmental program o!er no realistic hope for
any useful solution for power from fusion. The IEF con"
cepts studied here o!er not only such hope, but promise
power systems of such desirable performance and fea"
tures as to render all other forms of power production
uneconomic, unacceptable and obsolete. Success of IEF
power systems development promises to yield process
steam production at costs 1/3 to 1/2 of those experienced
today. This can be used to generate electrical power with
conventional thermal conversion means at lower cost
than from any other system, or to drive thermal power
requirements of a wide variety of commercial industrial
plants from paper"drying to desalination to alcohol pro"
duction. Use for desalination promises potable water
from the sea at costs less than #1.00 per thousand gal"
lons, an order"of"magnitude less than other means of sea
water conversion.
Use for production of anhydrous ethanol from two"
crop"a"year sugar cane bio"sources in semitropical re"
gions can yield auto"fuel"grade alcohol at a net cost of
about #0.25/gallon, after credit for byproduct synthetic
wood production is accounted. A single cane eld 30
miles square could provide enough alcohol fuel to supply
1$ of the U.S. auto economy; 100 such elds could
power the entire U.S. auto eet. Since ethanol burns
clean and smoothly %several million cars currently run on
it in Brazil& at lower"than"gasoline temperatures, the
problem of auto"generated smog would largely disappear
with its deployment and use in the U.S. And, it ts all of
the current marketing and distribution infrastructure in
place for the gasoline industry. Petroleum reneries
could shut down, and the oil companies could make new
fortunes from low"cost alcohol sold at market prices in
the auto"developed countries. And, the international oil"
based political and economic power of the oil"bearing
states would vanish overnight.
Other applications to which IEF systems can uniquely
apply include the use of DT"fueled versions operating at
very high rst wall ux to provide large neutron uxes
for transmutation burn"up of ssion product wastes.
Study of this possibility has been done by EMC2 over
the past 10 years. Combining this with results of the
current analyses of DT fusion output from IEF devices
shows that such systems could even transmute the usu"
ally inaccessible short"lived noxious isotopes of Sr and
Cs that are produced in quantity in the ssion process.
IEF Transmutation Waste Burner %TWB& plants thus
seem feasible, given the basic IEF unit itself. A paper
describing this TWB system and its burn"up perform"
ance has been prepared for presentation at the Global ‘93
Conference on the Nuclear Fuel Cycl%, to be held in Seattle
in September, 1993. A copy of this paper is given in Ap"
pendix B.
And still other applications of clean p11B systems seem
possible, that can provide enormous reductions in the
costs of space"ight, by the development of aerospace
power and propulsion engine systems of very advanced
performance capabilities based on IEF fusion source
systems.22,23 If successful, these engines could realistically
yield practical spaceight, at costs not much greater
than current high"speed long"range air transport.
But, of course, the principal application of immediate
concern, with IEF development success, would be to
central station power of the sort required by electric
utilities to run urban centers; indeed to run America.
Here the promise of IEF systems is %eventually& that of
completely clean power, with direct conversion at up to
80$ e'ciency, and with corresponding reductions in
thermal pollution and power cost by factors of at least
2x from those at present. The fusion fuel for plants that
use these units will be essentially free in terms of cost
per unit energy output, and the fuel resources are %as
noted earlier& very nearly innite relative to the life of
the Earth, itself. This enormous promise would seem to
argue that the Electric Power Research Institute, which
has supported this study, should logically undertake the
development of this new means of fusion power genera"
tion at the earliest possible time.
58
B. Needed Further Study and Analysis
Most of the near!term research needs and directions
have been discussed in the prior section of this report.
It is su"cient here to note that further progress in prov!
ing or disproving the potential of these IEF/IEC con!
cepts absolutely requires the modest!scale experiments
that have been described above. Further analytical and
numerical computer simulation work can, of course, be
useful and shed new light on second and third!order
issues, but the critical questions of basic system function
can be answered only by experiment; theory and calcula!
tion can not provide these answers.
Of course, many experiments can be devised and carried
out that relate in secondary ways to the IEC/IEF issues
of importance, that would still give useful and interest!
ing information. These include such tests as those in!
volving larger scale experiments at small currents to rep!
licate Hirsch’s early results, studies of exotic geometries
such as the Miyamoto E x B systems proposed 40 years
ago that o#er interesting small!scale physics but no
hope of fusion power, etc. In general, a great many ex!
periments could be done that would be interesting but
not of importance to the central questions of IEC con!
cept feasibility. Until these are addressed and answered,
it is not logical to consider further development. All
such non!denitive research is fundamentally simply
digging randomly in the garden of physics for the en!
joyment of the study. What is needed now is a tightly
focussed program of connected research e#orts, all
aimed at the most critical “show stopper” physics issues,
at the soonest possible time. Without this, no real pro!
gress can be made; with this, progress can be very rapid
indeed and, if successful, can yield commercial fusio!
power in less than two decades.
$4% Undertake limited power plant conceptual de!
sign studies based on the assumption of success
of IEF research and development, in enough
detail to assess and estimate their considerable
advantages vis a vis other types of power plants.
In addition, it seems time to begin to discuss the pros!
pects of IEF/IEC systems with both the nancial mar!
kets and utilities managements to develop a dialogue for
future communication of the relevant business issues
that confront their eventual use. In particular, a major
education campaign is needed to allow both groups to
understand and discriminate between the two di#ering
approaches to fusion power represented by the heavily!
advertised, large!scale concepts of the huge U.S. gov!
ernment program as contrasted with that of the small!
scale IEF/IEC approach.
If the EPRI & as the ombudsman and technology / sci!
entic conscience of the utilities can provide support
for these actions, and provide the leadership needed to
initiate them and the dialogues suggested above, then
there may be hope for practical, economical, clean fu!
sion power. If not EPRI, then who else?
C. Recommended Actions
These are only four in number:
$1% Support further parametric scoping and design
studies of IEC systems and concepts to dene
all possible critical experiments in greater detail,
and to greater level $i.e. beyond those that char!
acterize the rst!order issues%.
$2% Support experimental studies now, at modest
scale, to test the most critical rst order physics
issues in the best possible fashion, under the
present state of knowledge and experiment de!
sign.
$3% Begin examination of both the controls and
electrical conversion aspects of IEF power sys!
tems, to develop improved understanding of
their inherent features, stability, load following
capabilities, fail!safe mechanisms, et al.
59
9. References
13 Robert W. Bussard, “Some Physics Considerations of
Magnetic Inertial$Electrostatic Connement: A New
Concept for Spherical Converging$Flow Fusion,” Fusion
Technology., Volume 19, 273 "1991#
Nicholas A. Krall, “The Polywell: A Spherical Con$
verging Ion Flow Device,” Fusion Technology, Volume
22, 42 "1992#
14
Philo T. Farnsworth, “Electric Discharge Device for
Producing Interactions Between Nucleii,” U.S. Patent
No. 3,258,402 "June 28, 1966#
1
Robert L. Hirsch, “Inertial$Electrostatic Connement
of Ionized Fusion Gases,” Journal Applied Physics, Vol$
ume 38, 4522 "1967#
2
R.W. Bussard, “Method and Apparatus for Controlling
Charged Particles,” U.S. Patent No. 4,826,626 "May 2,
1989#
3
Irving Langmuir and Katherine B. Blodgett, “Currents
Limited by Space Charge Between Concentric Spheres,”
Physics Review, Volume 24, 49 "1924#
4
15 J. Luce, “Collective Field Acceleration of High$Energy
Ions,” in Inertial Electrostatic Connement and the
Physics of Relativistic Electron Beams, editor H. Sahlin,
New York Academy of Science, 1975
Robert W. Bussard and Katherine E. King, “Phe$
nomenological Modeling of Polywell / SCIF Multi$Cusp
Inertial$Electrostatic Connement Systems, paper 2T10,
Annual Meeting Division of Plasma Physics, APS,
Tampa, FL, November 4$8,1991, Bulletin American
Physics Society, Volume 36, 2319 "1991#
16
Katherine E. King and Robert W. Bussard, “EKXL: A
Dynamic Poisson$Solver for Spherically$Convergent
Inertial$Electrostatic Connement Systems,” paper
2T11, Annual Meeting Division of Plasma Physics, APS,
Tampa, FL, Nov. 4$8, 1991, Bulletin American Physics
Society, Volume 36, 2319 "1991#
17
W.W. Salisbury, “Method and Apparatus for Producing
Neutrons,” U.S. Patent No. 2,489,436, "November 29,
1949#
5
W.C. Elmore, J.L. Tuck and K.M. Watson, “On the
Inertial$Electrostatic Connement of a Plasma,” Physics
Fluids, Volume 2, 239 "1959#
6
7 Philo T. Farnsworth, “Method and Device for Produc$
ing Nuclear Fusion Reactions,” U.S. Patent No.
3,386,883, "June 1968#
D.C. Baxter and G. W. Stuart, “The e%ect of charge
exchange on ion guns and density in inertial electro$
static connement devices,” Journal Applied Physics,
Volume 83, 7 "1982#
8
9 Robert W. Bussard, “Potential and Density Distribu$
tions in Inertial$Electrostatic Connement Systems,”
paper 1D12, International Sherwood Theory Confer$
ence, Santa Fe, NM, April 6$8, 1992
Robert W. Bussard, “Ion$Acoustic Waves and Wave$
Group$Trapping in IEC Systems.” paper 8S32, and, with
K.E. King and L.W. Jameson, “Particle Trapping and
Electron Two$Stream Instability in IEC systems,” paper
8S31, Abb. Meeting Division of Plasma Physics, APS,
Seattle, WA. Nov. 16$19, 1992, Bulletin American Physics
Society, Volume 37, 1581 "1992#
10
11
op cit ref. 3
Robert W. Bussard, “A New Physical Process, Method
and Apparatus for Creating and Controlling Nuclear
Fusion Reactions,” U.S. Patent No. 5,160,695, "Novem$
ber 3, 1993#
12
F.F. Chen, Introduction to Plasma Physics and Con$
trolled Fusion, Second Edition, Volume 1, Plenum Press,
NY, 1988, Chapter Six.
18
M. Rosenberg and Nicholas A. Krall, “The e%ect of
collisions in maintaining a non$Maxwellian plasma dis$
tribution in a spherically convergent ion focus,” Physics
Fluids, Volume B4"7#, 1788 "1992#
19
R.A. Krakowski, “Progress in Commercial Magnetic
Fusion Energy Reactor Designs,” Fusion Technology,
Volume 20, 121 "1991#
20
21 R. Moir and W.L. Barr, “Venetian Blind Direct Energy
Convertor for Fusion Reactors,” Nuclear Fusion, Volume
13, 35 "1973#, and “Test Results on Plasma Direct Conver$
tors, Nuclear Technology and Fusion, Volume 3, 98 "1983#
Robert W. Bussard and L.W. Jameson, “The QED En$
gine Spectrum: Fusion$Electric Propulsion for Air$
Breathing to Interstellar Flight,” AIAA ppr 93$2006,
29th Joint Propulsion conference, Monterey, CA, June
28$30, 1993
22
Robert W. Bussard, L.W. Jameson, and H.D. Froning,
Jr., “The QED Engine: Fusion$Electric Propulsion for
Cis$Oort/Quasi$Interstellar "QIS# Flight,” 44th Con$
gress of the International Astronautical Federation,
Graz, Austria, October 16$22, 1993
23
60
A Proposal for Analytical,
Phenomenological and Performance
Studies of Electrically!Driven
Non!equilibrium Fusion "ENF# Devices
and Systems
Submitted by EMC2, 9100-A Center Street, Manassas, VA 22110 on 12 February
1992 to the Electric Power Research Institute
Statement of Work
of attainment of critical technical performance levels
within each plant system.
Energy/Matter Conversion Corp. "EMC2# , will under!
take a ten month study of systems applications and of
analytical and phenomenological parametric models of
electrically!driven non!equilibrium fusion "ENFl sys!
tems. The ENF concepts and devices to be studied will
be those embodying spherically!convergent ow devices,
based on ion acceleration by inertial!electrostatic
means. The proposed program will be carried out at its
o$ces in Manassas, Virginia.
Alternate thermal and electrical power systems uses will
be identied, and summary estimates made of the po!
tential performance/costs of each of these. These will
include water desalination with cogeneration plants,
nuclear fuel production in DT systems, hybrid power
and ssion product waste disposal burners, low!cost
process steam for synthetic fuels "e.g. ethanol produc!
tion, and other applications as may appear of interest.
Systems Studies
Parametric studies of potential power” systems will be
carried out with emphasis on technology tradeo%s and
R&D requirements for economic viability. Power sys!
tems concepts will be chosen that t EPRr criteria for
use and utility in central station power generation, and
these criteria will be assessed and ranked among the
various systems concepts analyzed. Economic and tech!
nology modeling codes will be developed for both direct
electric and thermal!conversion power systems, using
several candidate fusion fuels. Special attention will be
paid to radiation!free aneutronic systems.
Optimal candidate power systems will be selected from
those studied, and estimates will be made of the capital
and installation costs of each such candidate system in
nal commercial power!generating congurations. The
plants will include both thermal!electrical conversion
systems and direct electrical conversion from charged
particle fusion products. Insofar as possible, these will
be based on EPRI standards for plant costing and will
be used to analyze the costs of power as a%ected by level
&
An intermediate summary report and review of this
work will be completed within 6 months of program
initiation "September 15, assuming March 15 start date#
and will be given to EPRI program management at that
time, as desired. Feedback from this reporting/review
period is expected to sharpen the focus of further stud!
ies which will be concentrated on the specic directions,
issues and plant concepts that are seen as of most inter!
est by the EPRI at that time.
Physics Studies
Supporting physics and analytical studies will be made
during this period to provide a solid scientic underpin!
ning for the systems concepts. This analytical and phe!
nomenological modeling research study activity will
start from a denition of the critical physics issues that
are inherent in the ENF concepts of interest. These will
be determined early in the program, from the 5+ year
background work of EMC2 in study and modeling of
such systems.
An initial review of these will be made with EPRI pro!
gram management, and alternate issues added "or de!
A!1
leted! at that time. Critical areas are seen to be those of
ion core convergence, electron losses, presence of high"
Z ions in the, central core, ion/grid collisions, wall sput"
tering, bremsstrahlung losses #versus fusion power gen"
eration!, core fusion reaction products and multiple
ionization e$ects, reaction control physics, ion and elec"
tron convergent ow stability, et al.
Engineering Studies
Analytic models of the interactive physics involved will
be prepared and used to develop a systems performance
modeling computer code, capable of use for studies of
the e$ects of physics parameters on plant system out"
put. Preliminary results of these studies will be used to
determine critical engineering issues that are found in
promising plant systems. These will be modeled in algo"
rithmic form and used to extend the parametric physics
code to encompass engineering constraints and limita"
tions. The joint code thus achieved will be used for
parametric studies to test the e$ects of engineering
technology level on system performance.
Critical engineering issues will include radiative thermal
and particle loadings on walls, grids and ion acceleration
%
structures #e.g. ion guns!, particle ux levels\ stored en"
ergy within the ow system, energy storage in driver
systems, neutron ux, hazards and material damage ef"
fects #for those systems that use neutron producing fu"
els, e.g. DT, DD, 50:50 D3He!, protonic radiation haz"
ards #e.g. X"rays!, direct electric conversion voltage gra"
dients, convertor arcing and breakdown fueling and con"
trol of core reaction rates, et al.
Plant/System Development Plan
Using the critical physics and engineering issues and
requirements identied by the physics, engineering and
plant systems parametric studies, the research and de"
velopment required for achievement of successful plant
systems will be laid out and put into programmatic
form. Cost estimates will be made for the conduct of
this needed R&D work, and plans made that show the
time scales on which this might be accomplished.
Schedules and costs will be given as from “optimistic” to
“probable” to “pessimistic” to illustrate the range of
costs and time scales that might be found in a realistic
development situation.
A"2
Outline Schedule — Proposed Research Program
Systems Studies
Physics Studies
Engineering Studies
Define candidate systems
Develop first codes
Run first parametric studies
Improve codes with physics
Add engineering constraints
Run parametric studies
Dev. plant economic models
Est. system/cost performance
final plant cost studies
Define critical physics issues
Develop physics modlg. algorithms
Include poissors, ICC effect, et al
Run early para. studies, normalize models
Add new algorithms, run para. studies
Define pwr. sys. models
Inc. analytic algorithms
Test in early code
Parametric engineering studies
Define critical R&D needs
Outline R&D plans and costs
Model development process
Estimate development costs
Lay out coherent plant
Plant/System
Development Plan
Program Reviews
Reports/
Documentation
0
1
2
3
4
final
briefing
4-mo
4-mo
5
6
7
8
9
Months from Program Start
!
A"3
Report to EPRI on the Inertial
Electrostatic Fusion Project
Presented at EPRl Headquarters, Washington, DC, October 15, 1992
Panel Members
II. Main Features of IEF Concepts
Stephen O. Dean
Robert A. Gross
Robert A. Krakowski
Gerald L. Kulcinski, Chairman
Dale M. Meade
Dennis Papadopolous
Robert S. Symons
The lEF class of concepts are fundamentally di$erent
than either magnetic !connement" fusion or inertial
fusion. The basic concept utilizes a deep spherical po#
tential well to attract cold ions from the edge inward.
This results in the formation or a core region in the cen#
ter or the sphere where high energy ion beams converge
and collide. resulting in fusion.
I. Introduction
The EPRI panel !see Appendix A" on Inertial Electro#
static Fusion !lEF" met in Washington DC on October
15. 1992 to hear a presentation of the IEF program e$ort
funded by EPRI. The current IEF program was pre#
sented by Doctors Robert W. Bussard, Nicholas A.
Krall, and Richard Nebel !see Appendix B for the
agenda".
The specic charge to the panel !with respect to the
EPRI sponsored research" is summarized below:
1.
Has the analysis focused on the proper physics
issue?
2.
Does the sum total of the physics analysis point
to reasonable promise for the concept?
3.
Are there any obvious reasons why the concept
might not work other than those already iden#
tied?
4.
Are there attractive engineering features to
these concepts?
5.
Does the limited engineering !analysis" to date
appear reasonable?
6.
How well do IEF concepts promise to meet the
EPRI criteria for useful fusion power systems if
they can be proven technically feasible?
The organization of this report is as follows. Following a
brief description of the IEF concept in Section II, the
specic response or the panel to the EPRI charge is
given in Section III. The overall conclusion from the
panel are presented in Section IV.
&
Two generic congurations have been proposed:
•
•
Ion acceleration !IXL" where radially converging
energetic ion beams, produced by spherical elec#
trodes !or grids". create a virtual anode in the cen#
ter.
Electron acceleration !EXL" where high energy
electrons are injected along a magnetic eld cusp
to produce a negative potential which accelerates
low energy ions from the periphery.
Fusion energy is produced from the highly non#
Maxwellian unconned plasma in a small region at the
center or the sphere which is isotropic and essentially
mono#energetic. Since the scattering cross#section is
much larger than the fusion cross#section. the ions must
be recirculated and refocused many thousands of times.
There is no “connement” at the center, nor thermonu#
clear bum at the focus. Rather, the spherically colliding
“ion beam” results in fusion reactions at the center.
However. the electrons near tho outer region of the
sphere must be conned by magnetic cusps or, 1n the
case of ion guns. the ions must be conned by electro#
static grids. The e%cient connement of electrons
!main energy loss" and the strong focus~ of directional
ions !main energy source" arc tho key physics issues to
be addressed.
III. Specic Responses to
Panel Charge
Following the technical presentations for nearly half of a
day and discussions of the lEF project with the propo#
nents, the panel members have jointly drafted the fol#
lowing responses.
A#4
Panel Member Report on the
Inertial Electrostatic Fusion Project
November 2, 1992
1. Has the analysis focused on the
proper physics issue?
Physics analysis have thus far concentrated on the fol!
lowing issues:
A. The formation of the spherical potential well.
B. E"ectiveness of the spherical electrostatic well
against the loss to grids or along magnetic eld
cusps.
C. The formation and control of self!consistent
ion and electron density distributions.
D. An evaluation or e"ects #e.g., scattering and
instabilities$ that would prevent ion conver!
gence
It is the opinion of the panel that most of the important
physics issues have been identied #with items B & D
given the highest priority$. The limited theoretical
analysis done to date has focused on the appropriate
parameters with the exception of start!up scenarios.
However, identifying and solving problems are two dif!
ferent issues. Examples of the current state of under!
standing in some of the physics areas listed above are
given below.
Formation of the Spherical Potential Well
The IXL!concept was originally devised by Farnsworth
and Hirsch and early experiments generated interest by
producing 1010 neutrons/second using 100 keV ion
beams. The performance of this device was believed to
have been limited by losses of particles to the grids.
The EXL concept, developed in the 1980’s by Robert W.
Bussard, uses high order multipole magnetic elds to
generate a grid onto which electrons are injected to
form the potential wells. Experiments carried out on the
device at Directed Technologies have shown the forma!
tion ora 18 kV electrostatic well as measured by probes.
'
Sustainment of the potential well
The grid system used on the original IXL was thought
to limit performance of this type of system and prevent
reactor applications. However, the performance was
alway, substantially better than predicted by any theory.
The EXL was designed to avoid the grid loss problem. A
critical issue %or EXL is the size of the electron loss
channel through the magnetic cusp. This issue is being
addressed with analytic and computer models which
suggest that the electron leakage rate may be su&ciently
low that an interesting reactor may result. However,
there is no denitive experimental data base on high!
beta cusp particle loss rates. This is a critical physics
question that needs to be resolved.
Evaluation of Effects that Would Prevent Ion
Convergence
A critical issue is whether, near the center or the sphere,
a dense, steady. highly non·Maxwellian plasma can be
obtained and sustained. Theoretical analyses raise the
hope that this can be accomplished but again, there is
no experimental data base to verify the theoretical pre!
diction. Central to this issue of refocusing is the degree
to which ions entering and decelerating in tho edge re!
gion can be “specularly” #i.e, isotropically$ returned to
the reaction core radius without sampling the edge
magnet eld, that in turn. could impart undesirable an!
gular momentum.
2. Does the sum total of the physics
analysis point to reasonable promise
for the concept(s)?
The lEF concepts that have been analyzed thus far sug!
gest that relatively small, and e&cient, fusion power
reactors #10’s to 1oo’s of MWe may be possible. In addi!
tion, because of the spherically focused ion beam, these
concepts may employ advanced #neutron lean$ fuels,
thereby reducing or eliminating many of the material
and radioactive waste problems associated with “conven!
tional” thermonuclear fusion power plants. Further!
more, the unique characteristic of the non!Maxwellian
A!5
reaction core !i.e., isotropic but mono"energetic# may
allow “tuning” of the ion energy to optimal regions of
high fusion cross sections.
4. Are there attractive engineering features to these concepts?
Whether the idealized, spherically symmetric, steady"
state IEF conguration, as proposed by Bussard and
analyzed by Krall can be obtained and sustained, with
acceptable energy to the system !recirculating fractions
<20$#, is a major question. If so, and the reactors can be
obtained in small unit size !%100 MWe or less#, then very
attractive fusion reactors using fuels which produce only
low levels of neutrons may be possible. Such a combina"
tion would result in an exciting and very di&erent di&er"
ent fusion concept with respect to high power density.
separation / elimination of plasma support systems. re"
duced radiation damage and radioactive wastes, and re"
duced R&D e&ort toward a commercial system.
The reduced levels of synchrotron radiation !because of
the relatively low magnetic eld in the Polywell concept
or the reection of ions by electric elds of grids in ion
injection schemes# suggests that it may be possible to
use advanced fuels !D3He, p11B, 3He3He# in IEF reactors.
The fraction of fusion energy in neutrons from these
cycles varies from a few percent to zero, thus nearly
eliminating radiation damage to reactor materials and
greatly reducing the volume of radioactive waste to be
handled during maintenance and decommissioning.
3. Are there any obvious reasons why
the concept might not work other than
those already identied?
As stated previously, the reactor potential and engineer"
ing appeal of IEF depends mainly on two factors; the
electron losses from the surface of the device and tho
extent of ion convergence to a high density core. Failure
to obtain the necessary performance in either of these
areas could result in an uneconomical system. It is also
important to include these loss mechanisms in the
startup scenario; many concepts have failed before be"
cause they could not achieve the desired operating de"
sign point.
The electron loss rate controls the reactor energy bal"
ance. The losses at a “point” magnetic cusp !at high
beta# is an important, yet unresolved issue a&ecting the
electron loss"rate. Furthermore, the extent to which
high"beta diamagnetic e&ects reduce the electron loss
through the open cusp regions needs to be understood
theoretically and experimentally. Success in this latter
area would improve the attractiveness of the IEF con"
cepts. It is also not clear how the grids would be cooled
especially if they must be designed with a small cross
section to avoid intercepting an unacceptable fraction
of the particles.
While there have been insulator problems with electron
and ion guns in some experiments to date, there seem
susceptible to solution. We believe there is an adequate
technology base for the charged particle guns necessary
for the various proposed reactor concepts.
Other issues such as radiative losses, electron"ion colli"
sionality, reaction product !mass# asymmetry and collec"
tive e&ects, while important factors in the reactor de"
sign are not expected to result in “go"no go” problems
for the concept.
(
The current studies indicate that lower magnetic eld
levels !% a few Tesla# are needed in IEF reactors. Super"
conducting magnets have not been proposed !or maybe
not even needed# for the concepts reviewed. Both of
these “low tech” requirements could eliminate costly
magnet development programs. The smaller size possi"
ble in such reactors !compared to current magnet fusion
congurations such as Tokamaks, Stellarators, RFC’s,
RFP’s. etc.#, would allow many units to be built and im"
proved before the next larger device was constructed.
Furthermore. the steady"state operating mode for lEF
systems would avoid the cyclic thermal fatigue of metals,
which is currently a major concern for fusion reactor
designers.
Other attractive features from a reactor view point are
the ability to use all the charged particle energy for
direct"energy collection and the elimination of any ex"
ternal ion heating requirement.
5. Does the limited engineering
(analysis) to date appear reasonable?
A reasonable e&ort has been made !by EMC2 scientists#
to consider the many characteristics of an IEF fusion
power system. These include overall power balance and
subsystem e'ciencies, plant capacity, fuel cycles, impu"
rities and ash removal techniques, structural materials,
coolant systems, blankets, energy conversion systems,
hazards, startup, and control scenarios. The modest ef"
fort to date indicates that a reasonable design window
may exist for an attractive power plant conguration,
provided the reactor physics is favorable. More detailed
analysis would be required to achieve a realistic and self"
consistent quantitative engineering design that would
identify the most critical engineering issues associated
with the extraction of energy from the advanced fuels
and permit cost estimates for pre"conceptual designs.
The nature of the engineering issues is intimately con"
nected to how the physics issues are resolved; for exam"
ple, the power balance may or may not, turn out to be a
critical issue; materials, safety, and energy conversion
systems will depend on the fusion fuel cycles that are
A"6
permitted by the physics. The near!term e"ort should
include a continuing modest e"ort on engineering de!
sign. but an enhanced engineering design e"ort should
await a rmer physics basis.
6. How well do lEF concepts promise
to meet the EPRl criteria for useful fusion power systems if they can be
proven technically feasible?
The recommended reactor characteristics from the 1992
EPRl Fusion Panel Report are listed below.
•
•
•
•
•
•
Simplicity of concept.
Power plant designs without tritium burning be!
cause of the very serious problems associated with
14 MeV neutrons.
Low!activation materials.
High overall energy conversion e#ciency, e.g.,
combined direct electrical and thermal conver!
sion.
Reduction in the outage and waste disposal. prob!
lems of changing out large volumes of fusion reac!
tor core materials every few years.
The importance of e"ective ash removal from
fusion plasmas.
From the presentation to the panel, we draw the follow!
ing conclusions with respect to the desired reactor char!
acteristics above.
Simplicity
The small physical size, the spherical geometry, and lack
of interlocking coils and the separation of the fusion!
power!producing!core should make the maintenance
and construction of an IEF reactor much easier than
most $if not all% known magnetic fusion concepts. The
fact that only electrons need to be conned in a MHD
stable conguration, the lack of external heading and/or
current drive power sources, and the absence of ash con!
trol components should all contribute to a simple ge!
ometry for the power producing cavity.
negligible levels $<<1g%, which means that the release of
the entire tritium inventory in a DD IEF will cause
much less than a 10 mrem absorbed dose to the most
exposed individual at the fence of the plant site. Reactor
licensing will be considerably easier with the virtual
elimination of induced radioactivity and tritium inven!
tories characteristic of the D3He and p11B fuels.
Low Activation Materials and Reduced Radioactive Waste
Ai the present time, “conventional” alloys in DT fusion
reactors will have to be periodically replaced because of
radiation damage caused by neutrons. These radioactive
components will probably. have to be stored under!
ground in the U.S. at costs to a power plan owner com!
parable to those for ssion reactor wastes. Lower activa!
tion alloys $for long!lived isotopes% could be developed,
but only after long and expensive research to verify the
radiation performance of such alloys. Furthermore, it is
not clear that the U.S. will continue with near surface
burial and may require even low!activation alloys to be
stored underground $as in Europe%. Any reactor that can
use neutron!lean advance fuels and avoid periodic re!
placement of its rst wall and blanket should be able to
save on waste disposal costs.
The ability to use “conventional” alloys in IEF reactors
could also reduce the capital construction and R&D
costs. In addition to a drastic reduction in radioactivity
from the D3He and p11B fuels. there will also be a corre!
sponding reduction in after!heat which should allow
such reactors to qualify for an “inherent” safety classi!
cation.
High Overall Efciency
The possibility to use direct conversion of the reaction
product energy to electricity could double the plant e#!
ciency, thus reducing capital costs and alleviating envi!
ronmental concerns about thermal pollution. The lack
of massive and sometimes unreliable heat exchangers
and turbine equipment should also contribute to reli!
ability.
High Availability
Avoidance of 14 MeV Neutrons and the Tritium Handling Problem
The ability to use neutron!lean fuel cycles such as DD,
D3He, p11B, and 3He3He will have a very positive e"ect
on lowering both the nancial and time barriers to
commercial fusion. The fact that one does not have to
develop new materials to withstand 14 MeV neutron
damage could save considerable R&D costs and substan!
tially shorten the time to develop commercial fusion
power plants. The tritium inventory can be reduced to
&
The most important feature of neutron lean fuels will be
the great reduction in radiation damage and tho avoid!
ance of periodic shutdowns to replace the highly radio!
active rst!wall and blanket components. The “solid
state” approach to electrical conversion versus the rotat!
ing machinery approach of all other MFE concepts
should increase the plant availability.
The use of relatively low technology magnets, the acces!
sibility of geometry, and the lack of interlocking mag!
A!7
nets will also contribute to a relatively higher availability!
than achievable in toroidal systems.
Importance of Ash Removal
In Maxwellian conned plasmas, one of the most di"#
cult problems is keeping the plasma free from impurities
and reaction products. This is typically done by divert#
ing part of the plasma to a collection plate and pumping
out the volatile ash. However, the redirected plasma
imposes very high heat uxes and stresses to the reactor
components and is arguably the most di"cult problem
faced by traditional DT magnetic fusion reactor design#
ers today. The IEF concept avoids this problem by al#
lowing the fusion products escape over the potential
“hill” and collecting their kinetic energy well outside the
reactor. Some concern has been expressed by the possi#
ble buildup of impurity ions, sputtered from the grids or
formed by the ionization of background gas, which
could collect in the core region.
IV. Conclusions
The overall conclusion of the panel is that IF the plasma
physics questions addressed earlier can be solved in re#
actor congurations described in the review, the IEF
concepts could meet the EPRI requirements for a desir#
able reactor better than any other magnetic fusion con#
cept proposed to date. More detailed engineering design
of a self#consistent cost#e$ective system is crucial to
verify this conclusion.
The conclusions of the panel listed below, should be
considered in the context that the review conducted was
relatively short and focused on the work funded by
EPRI. The panel did not review the experimental pro#
gram of Directed Technologies nor the past work on the
IXL concept, Nevertheless. it was felt that su"cient
information was presented to allow some broad re#
sponses to be made to the EPRl charge.
1. Has the analysis focused on the
proper physics issue?
The electron cusp losses and ion convergence character#
istic are certainly the most crucial physics issues that
need to be addressed in the near term IEF concepts.
These issues have been identied, along with several
others and a vigorous theoretical program was in place
to analyze them. There is essentially no experimental
program in place to verify or refute the theoretical re#
sults.
!
2. Does the sum total of the physics
analysis point to reasonable promise
for the concept(s)?
The analysis thus far suggests that relatively small %#100
MWe&, spherically symmetric, low neutron yield, ad#
vanced fusion fuel powered reactors are possible. If the
physics can be demonstrated %experiments to date have
been inconclusive& this concept represents a truly revo#
lutionary approach to fusion energy.
3. Are there obvious reasons why the
concept might not work other than
those already identied?
If the desired plasma conditions can be obtained, there
are no other obvious reasons why the IEF approach
would not work outside those already identied. Issues
brought up at the review such as radiative losses in high
Z fuels, electron#ion collisionality and collective e$ects
appear amenable and not of the “go#no go” variety. One
issue that could be in the “make#or#break” category is
the demonstration that ions can be simultaneously ac#
celerated by the electrostatic eld in a high beta mag#
netic cusp and shielded from the magnetic elds. An#
other is the buildup of low energy impurity ions, formed
by sputtering of the grids or ionization of the back#
ground gas, in the core region. Although these issues
were identied, more analytical work is needed.
4. Are there attractive engineering features to these concepts?
The main attractive feature of the IEF concept is its
ability to make e$ective use of fusion fuels which have
low neutron yields. The reduction or even elimination of
neutrons from the plasma greatly alleviates the radiation
damage in and radioactive waste from a fusion power
plant. The small size allows a relatively inexpensive de#
velopment path to be pursued. Similarly, the relatively
low “technology magnets and lack of a breeding blanket
should contribute to a much more robust reactor that
will satisfy safety concerns of regulators.
5. Does the limited engineering
(analysis) to date appear reasonable?
Given the level or funding devoted to engineering, it is
impressive how much has been accomplished to date.
However, much more detail will be required before the
commercial attractiveness of IEF systems can compete
with the more established toroidal concepts on an equal
“knowledge” basis.
A#8
6. How well do IEF concepts promise
to meet the EPRI criteria for useful fusion power systems if they can be
proven technically feasible?
IF the plasma physics questions addressed earlier can be
solved in a reactor conguration close to that presented
to the panel, the IEF concepts have the potential to
meet the EPRI requirements for a desirable fusion reac!
tor better than any other magnetic fusion device. A self!
consistent physics/engineering power plant design will
be required to assess the commercial attractiveness of
the IEF concept.
In addition to the responses above, the committee felt
that it should add the following recommendations with
respect to the IEF concept:
A. The IXL concept be “revisited.” A major part of
the recommended e"ort in this area should be
to develop theories to explain the results ob!
tained in the 1960’s.
B. Start!up physics should be addressed and a
start!up scenario developed to the degree that
one can be assured that there is no road block
to an attractive power generation scenario.
#
A!9
Members of IFE
Panel
Professor Robert A. Gross
Columbia University
Seeley W. Mudd Building
500 West 120th Street
School of Engineering & Applied Science
Plasma Laboratory
New York, NY 10027
!212" 854#2967 phone
!212" 854#8257 fax
Mr. Robert Symons
Technical Director
Litton Systems, Inc.
960 Industrial Road
San Carlos, CA 94070
!415" 591#8411, ext. 327 phone
!415" 591#5623 fax
Dr. Stephen O. Dean
President
Fusion Power Associates
2 Professional Drive
Suite 248
Gaithersburg, MD 20871
!301" 258#0545 phone
!301" 915·9869 fax
Dr. Dennis Papadopolous
Astronomy Department
University of Maryland
Stadium Drive
College Park, MD 20742
!301" 405#1526 phone
!301" 405#9966 fax
Dr. Robert Krakowski
Los Alamos National Laboratory
Post O$ce Box 1663
Los Alamos, NM 87545
!505" 667#5863 phone
!505" 665#5283 fax
Dr. Dale Meade
Princeton Plasma Physics Laboratory
Post O$ce Box 451
Princeton, NJ 08543
!609" 243#3301 phone
!609" 243#2749 fax
Professor Gerald L. Kulcinski
Professor Nuclear Engineering
University of Wisconsin
1500 Johnson Drive
Madison, WI 53706
!608" 263·2308 phone
!608" 263#4499 fax
%
A#10
Preliminary Agenda ! EPRI Inertial
Electrostatic Fusion "IEF# Review
October 15, 1992; EPRI Headquarters, Washington DC
8:00 a.m. $
Welcome and Charge to the Panel ! Robert L. Hirsch
8:15 a.m. $
Panel Chairman's Opening Remarks ! Gerald Kulcinski
8:25 a.m. $
Technical Presentations ! Robert W. Bussard; Nicholas Krall; Richard Nebel
1. Introduction and Summary
a. Background, Concepts and Baseline Design
b. Plasma Physics Characteristics
c. Fusion Engineering Features
II. Physics Characteristics
a. Particle and Potential Distributions
b. Spherical Counterow, Stability and Electrostatic Waves
c. Colllsionality Distribution and E%ects
d. Particle Recirculation and Losses
e. Numerical Simulations of IEF Plasma Systems
III. IEF Device Engineering Features
a. Plasma Particle and Radiation Power Losses
b. Structural Conguration, Thermal Loads and System Sizing
c. Fusion Power Distribution, Ash and Impurity Generation
d. Radiation Hazard Potential, X&Rays and Neutrons
IV. Power Systems
a. Baseline System Design Summary
b. Fusion Fuels, Power Balance and Scaling
c. System Hazards, Magnetic, Materials, Fields and Structures
d. Fueling System Stability and Control
e. Baseline and Alternate Plant Systems Performance
$
A&11
V. R&D Program to IEF Power
a. Critical Physics Issues, Status and Results
b. R&D Theory and Experiment to Proof!of!Feasibility
c. Prototype Plant Development and Demonstration
12:00 p.m." Working Lunch, includes Questions and Answers
1:00 p.m."
Continue Questions and Answers
2:00 p.m.! Panel Executive Session
* Discussion
* Conclusions and preparation of a letter report by Panel, EPRI personnel and DOE guests
4:00 p.m."
Adjourn
• Report is to be two to three pages providing answers to questions and qualitative judgments.
"
A!12
Fusion Technology Institute
Nuclear Engineering and Engineering Physics Department
University of Wisconsin-Madison
November 3, 1992
Dr. Robert L. Hirsch
Electric Power Research Institute
2000 L Street NW. Suite 805
Washington. DC 20036
Dear Bob:
On October 15, 1992 the undersigned met at EPRI Headquarters in Washington. DC, to review the
Inertial Electrostatic Fusion project. The charge to the committee was as follows:
1. Has the analysis focused on the proper physics issue? Does the sum total of that physics
analysts point to reasonable promise for the concept? Are there any obvious reasons why the
concept might not work other than those already identied?
2. Are there attractive engineering features to these concepts? Does the limited engineering
!analysts" to date appear reasonable?
3. How well do lEF concepts promise to meet the EPRI criteria for usefuL fusion power sys#
tems $%they can be proven technically feasible?
We were given an intense and highly informative brieng by Drs. Robert W. Bussard, Nicholas A.
Krall, and Richard Nebel on the current status of theory and reactor design activities pertaining to
the IEF concept. Our response to the charge and our conclusions are in the attached report. A
brief summary of our deliberations is given below.
• The IEF concept does represent an intriguing fusion alternative to the present U.S. and inter#
•
•
•
•
national DT tokamak program.
The promise of small, reliable, inexpensive, low radiation damage systems must be balanced
by the concern over the lack of a broad theoretical evaluation and the small plasma physics
experimental data base currently available.
Considerable attention does need to be paid to the problem of electron losses from the cusps
and the ion convergence characteristics. The concept will not be fully accepted by the plasma
physics community While these areas are experimentally established and better understood in
the context of a self#consistent model of the plasma.
It is recommended that if EPRI is interested in continuation of this project, the next step
should involve additional experimentalists who can help to design a small. but denitive ex#
periment to address some of the key plasma physics questions.
The ultimate promise of the IEF concept is su&ciently attractive that continued funding in
this area could be of great benet to the electric utilities and the fusion community at large.
These conclusions are discussed in more detail in the attached report.
Please feel free to contact me or any member of the panel should you wish more elaboration on its
contents.
'
A#13
Gerald L. Kulcinski
Grainger Professor of Nuclear Engineering and
Director of the Fusion Technology Institute
University of Wisconsin!Madison
Committee Members
Professor Robert A. Gross
Columbia University
Mr. Robert S. Symons
Litton Systems
Dr. Robert A. Krakowski
Los Alamos National Laboratory
Professor Dennis Papadopolous
University of Maryland
Dr. Stephen O. Dean
Fusion Power Associates
Dr. Dale M. Meade
Princeton Plasma Physics Laboratory
"
A!14

IEC Electric Powerplants

Transcription

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