Technical Note
ORP/LV-76-7
ENVIRONMENTAL AND SAFETY ASPECTS OF
ALTERNATIVE NUCLEAR POWER
TECHNOLOGIES - FUSION POWER SYSTEMS
MAY 1976
^£DSX
"it PRO^°
U.S. ENVIRONMENTAL PROTECTION AGENCY
OFFICE OF RADIATION PROGRAMS
LAS VEGAS FACILITY
LAS VEGAS, NEVADA 89114
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Technical Note
ORP/LV-76-7
ENVIRONMENTAL AND SAFETY ASPECTS OF
ALTERNATIVE NUCLEAR POWER
TECHNOLOGIES - FUSION POWER SYSTEMS
Bruce J. Mann
May 1976
U.S. ENVIRONMENTAL PROTECTION AGENCY
OFFICE OF RADIATION PROGRAMS
LAS VEGAS FACILITY
LAS VEGAS, NEVADA 89114
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This report has been reviewed by the Office of Radiation Programs
Las Vegas Facility, U.S. Environmental Protection Agency, and ap-
proved for publication. Mention of trade names or commercial pro
ducts does not constitute endorsement or recommendation for their
use.
11
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PREFACE
The Office of Radiation Programs of the U.S. Enyironmental
Protection Agency carries out a national program designed^ to _ eval-
uate population exposure to ionizing and non-ionizing radiation
and to promote development of controls necessary to protect the
public health and safety and assure environmental quality.
Part of this program is devoted to an examination of existing
and proposed energy technologies with respect to radiological
health impacts. This effort includes a review of advanced nuclear
power technologies initiated during the past year. The Office of
Radiation Programs --Las Vegas Facility has conducted a review of
conceptual fusion power plant designs in support of this effort.
This technical note has been published in order to provide a
summary to the agency, to interested professionals in other orga-
nizations, and to members of the public of health and safety as-
pects at an early stage in this developing technology. This report
is not intended in any way to represent EPA policy with respect to
any aspect of fusion power technology or its environmental and
safety issues .
Readers of this report are encouraged to inform the Office of
Radiation Programs --Las Vegas Facility of any omissions or errors
and comments. Requests for further information are also invited.
Donald W. Hendricks
Director, Office of Radiation
Programs - Las Vegas Facility
111
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TABLE OF CONTENTS
PREFACE 11X
LIST OF FIGURES vi
LIST OF TABLES vii
ACKNOWLEDGMENTS viii
INTRODUCTION 1
The Tokamak Concept 2
History of Fusion Research 4
State of Development 5
State of the Technology 5
Limitations to Development 6
Research and Development Support 10
Potential Significance to Energy Sources 13
TECHNICAL DESCRIPTION OF CONCEPTUAL TOKAMAK
POWER REACTOR 15
1 8
Reactor Systems j-°
Plasma and Burn Cycle |°
Divertors ^
Fuel System ^
Blanket and Shield L*
Magnets *~
Cooling System and Steam Generators ^°
Some Comparisons of Fusion and Light Water
Fission Reactor Characteristics *jj
Radioactive Waste Management Systems
SAFETY ASPECTS 35
Design for Safety r,fi
Overall Reactor Design ^°
Coolant System--Blanket ^
Structure and Balance of Plant •5/
Potential Accidents and Engineered Safety
Features
IV
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ENVIRONMENTAL IMPACTS OF NORMAL OPERATION 39
Radiological Effluents 39
Non-Radiological Impacts and Siting
Requirements 42
REFERENCES 43
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LIST OF FIGURES
Page,
Figure •
1. Toroid Geometry 3
2. Main Features of a Tokamak 3
3. Schematic of Tokamak Fusion Reactor
Blanket Zones
4. Tokamak Fusion Power Plant Schematic 16
5. Principal Fuel Flows in Princeton Tokamak 21
6. Princeton Tokamak Cross Section Showing
Blanket, Shielding, and Main Field Coils 24
7. Wisconsin Toroidal Fusion Reactor Coolant
Loops 27
8. Tritium Flow in Wisconsin Tokamak 32
VI
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LIST OF TABLES
Table
1. Plasma Parameters in Tokamaks 7
2. ERDA-Controlled Thermonuclear Research
Funding Projections, 1975-1981 H
3. Tokamak Reactor Reference Design Features 17
4. Reference Design Plasma Characteristics 19
5. Flow Rates in Princeton Tokamak Primary
Fuel Loop Z2
6. Magnet Characteristics 25
7. Wisconsin Tokamak Cooling System Power
Distribution and Operating Parameters 28
8. Comparison of Some Fusion and Light Water
(PWR) Reactor Characteristics 31
9. Summary of UWMAK-I Tritium Extraction System
Characteristics 33
10. Summary of Radiological Quantities Associated
with Tokamak Designs 40
VII
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ACKNOWLEDGMENTS
Information contained in this review was obtained from the
references cited herein and from discussions with a number of
scientists and officials in the fusion community. Much information
and a number of helpful comments were provided by the Energy Re-
search and Development Administration, Division of Controlled
Thermonuclear Research (ERDA-DCTR). in particular, the comments
and assistance of Drs. R. L. Hirsch, M. J. Katz, and the DCTR stair
are appreciated. The fusion safety study group at the University
of California--Los Angeles, headed by Professor W, E. Kastenberg,
was also very generous in providing review comments.
The author also appreciates the contributions of Daphne
Prochaska and Pamela J. Pursey in typing and editorial assistance.
Thanks are due to David Melton for the adaptation and preparation
of the figures. Any errors in fact and interpretation are solely
the responsibility of the author, however.
Vlll
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LIST OF TABLES
Table Page
1. Plasma Parameters in Tokamaks 7
2. ERDA-Controlled Thermonuclear Research
Funding Projections, 1975-1981 H
3. Tokamak Reactor Reference Design Features 17
4. Reference Design Plasma Characteristics 19
5. Flow Rates in Princeton Tokamak Primary
Fuel Loop 22
6. Magnet Characteristics 2$
7. Wisconsin Tokamak Cooling System Power
Distribution and Operating Parameters 28
8. Comparison of Some Fusion and Light Water
(PWR) Reactor Characteristics 31
9. Summary of UWMAK-I Tritium Extraction System
Characteristics 33
10. Summary of Radiological Quantities Associated
with Tokamak Designs 40
VI1
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ACKNOWLEDGMENTS
Information contained in this review was obtained from the
references cited herein and from discussions with a number of
scientists and officials in the fusion community. Much information
and a number of helpful comments were provided by the Energy Re-
search and Development Administration, Division of Controlled
Thermonuclear Research (ERDA-DCTR). In particular, the comments
and assistance of Drs. R. L. Hirsch, M. J. Katz, and the DCTR staff
ar€Va?^ecl?ted; T^e £usion safety study group at the University
of California--Los Angeles, headed by Professor W. E. Kastenberg,
was also very generous in providing review comments.
The author also appreciates the contributions of Daphne
Prochaska and Pamela J. Pursey in typing and editorial assistance.
Thanks are due to David Melton for the adaptation and preparation
+L ™J^hM'/ny/^°rS i? faCJ and interpretation are solely
the responsibility of the author, however.
viii
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INTRODUCTION
the realm of £«£??!« JI? 7Jar - ' f^sion ""arch has moved from
actor design* J?rS? Physics into the stage where power re-
power dev^f™ are beinS seriously discussed. A national fusion
in excess o?P?2enS Z*^™ h*S C°me inJ° being> cu^ently funded
for the «rS f S mijllon Per year- Goals are being established
the operation LdeVel°Pment Sf fusi°n devices ^ich\re aimed at
° demonstrati^ reactor in the
A?C-Ude the anno^cement by the Energy
a fusion Administration (ERDA) of plans to build
the To£L»S J reactor (Science, I975a) . This machine, designated
S10
is Planned to ^ opr
un sienHr J%ex?ected to be the first machine capable of
Project fnr\g?^ J ant -10n ener^' Jt is also the first fusion
Prepared CERSA"C?97?aeVlr0nmental impaCt statement CEIS) has been
nave bepn^f J^f^^eP^al designs for large fusion power systems
? Published in the United States within the past several
Vera f theSe are falrly detailed and include safety
asPe?ts *° Drying degrees. A number of other
ep0 . er
with fSc? 6SS gene^1 Sa£ety and environmental issues associated
c
ball! fS0^' ThSSe a5d °th^ Cited references form the pri-
°r the Present discussion of fusion power systems anrt
environmental "sues. PEven thougT the
not permit assessments in the depth pos-
he^ei C"rrentl7 exist several different major desijm co
or fuef-Shed Prfmarily by the means with which the reading pm
?eac?nr contained. For an excellent review of the major fSsion
'
ebeyo is J?fe"ed to the artcle by
use n-F -i ' 1?75^' The two general confinement schemes are- the
fields to contain ^e plasma and the inertial con-
°£ theSG concePts, in turn, have Sev~eraT
ai?0typ^ of magnetic confinement systems are under
^ achieve conditions for sustained fusion reactions
5r?,are: 1} ^V,1"™ density closed toroidal
system' -' ftellarators^and Tokamaks, 2) high-plasma density
terns »' S'' linear and circular theta-pinches, and 3) open sys-
tems, e.g., magnetic mirrors (ERDA, 197Sa). 7
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confinement "«nt, but highly -publicized concept is the inertial
compression $»\ In this scheme, confinement is achieved by the
t2?S Iuffi?iSnt% "• P^eV° ultra-high density at a tempera-
let An Pvi? • >Sn1??" the constituent atoms of the fuel pel-
conditiorf?^ t°n I5 a£hieved under high temperature and pressure
iS ?his c«f Jjar t0 ^°S| in a thermonuclea? weapon, except that
Sosion remits ?rnrU-°£ifU^ is S° sma11 that ™^ a mi^roex-
is tha? a Si?r* iS f^g^D^terium-Tritium pellet. The theory
is tnat a device could be built wherein the energy is extracted
chamberraPMostTenCe°f,P?lle\eXpl0sions in a luitably-Seslgned
las^rJ ^r ^ost experimental work in this area involves the use of
energy ?o 1^1™^™* tO SUPply the r^uired intense burst of
the discussion will be centered on the tokamak
esons h nk^St?m: J4S choice has been made for several
pincMthe most tho^, ^i (wjth ^ Possible exception of the theta-
ab?e concep?SIl fusIongSiLdeV?loPe? a^d do^mented of the avail-
1974- Frill IQ?!? S10?,?ower Plant designs (Wisconsin, 1974; Mills,
total Federal aov.;™IhiS concePt receives about 60 percent of the
dition ?he" SubliiKS y!.aP°n! explosions (Science, 1975b) . In ad-
do no? 'contain suffic^tTf"-11™ concePtU8l power plant designs
comparison of envfr^io! detai1 *° support more than a limited
systems environmental considerations with other nuclear power
THE TOKAMAK CONCEPT
''''
derived ^om the
s
itself This si2i,?t«ni. induced electric current in the plasma
s V ". m*
finement Iystem7 g f the main feat^es of a tokamak con-
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POLOIDAL
DIRECTION e
MAJOR RADIUS
AZIMUTHAL, AXIAL,
OR TOROIDAL
DIRECTION^
MINOR RADIUS
Figure 1. Toroid Geometry
TRANSFORMER CORE
OHMIC HEATING
& POLOIDAL
FIELD COIL
(Primary Winding(-^
1$ PULSE-
POLOIDAL FIELD Be
TOROIDAL FIELD B
RESULTING SPIRAL FIELD
Figure 2. Main Features of a Tokamak
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In the tokamak, the plasma itself forms the secondary of a
set of transformer cores. In this case, the secondary is highly-
ionized (hence, electrically conducting) gas instead of a solid
conductor, as in a conventional transformer. A strong current
pulse in the primary windings ionizes the gas and generates an
axial (or toroidal) current in the plasma, as shown in Figure 2.
This produces two effects: 1) heating of the plasma, and 2) gen-
eration of a magnetic field, called the poloidal field, perpendic-
ular to the direction of the axial plasma current. A second mag-
netic field is generated in the axial direction (parallel to the
plasma current) by the toroidal field coils. The resulting field
lines from the combination of the toroidal (axial) and poloidal
fields form a helical configuration as indicated by the heavy
dashed line in Figure 2. Actually, due to the large number^of lines
formed in this manner, the resulting field creates a confining_
structure which resembles the crossed plies of an automobile tire
casing, as described by Rose (Rose, 1971).
These features comprise the essential ingredients of a Tokamak
confinement system; however, a number of additional requirements
must be met in order to achieve energy production. In a fusion
power system, conditions of plasma density, temperature (average
ion kinetic energy), and confinement time, sufficient to sustain
the reaction between the plasma components, must be achieved and
maintained. The energy produced in the reacting plasma must be
extracted, the reaction product atoms (ash) removed, and the plasma
replenished, or fueled, on a continuous basis. Cooling is required
and the large plasma chamber must be maintained at a relative vacuum.
HISTORY OF FUSION RESEARCH
The fusion reaction was discovered in the early 1930's and
was identified as the mechanism by which the sun and most stars
produce energy. Achievement of net energy production by fusion
first occurred via the thermonuclear weapon. The first such ex-
plosion took place in 1952. At about this time, research was ini-
tiated in this country, and elsewhere, on controlled fusion, as the
possible advantages of such a means of producing power were realized.
A number of fusion reactions are possible, including the Deuterium-
Tritium or D-T reaction, which can be fueled by deuterium, for
which an essentially inexhaustible supply exists; However, it was
also recognized that no known or conceivable solid material could
be used to contain the reaction at the 100 million degree Kelvin
temperature necessary for its sustainment (ERDA, 1975a) . Hence,
early investigations centered on the use of magnetic fields to con-
fine the reacting fuel of hot plasma.
As knowledge was gained about plasma behavior, particularly
the mechanisms by which plasmas tended to dissipate, early optimism
for achievement of controlled fusion was replaced by efforts aimed
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at gaining necessary further understanding of plasma behavior.
Today, some forty years after the discovery of the principle, sus-
tained, controlled fusion has not yet been achieved. In contrast,
four years elapsed between the discovery of nuclear fission and the
first sustained fission experiment. Fusion research has been, in
general, characterized by a series of small successes in achieve-
ment of conditions leading toward a sustained, controlled fusion
reaction. This has, by no means, been an orderly progression, as
a number of sometimes unexpected setbacks have been experienced
and the solutions sought.
STATE OF DEVELOPMENT
State of the Technology
The current status of fusion power research and development
can be discussed by first examining, quantitatively, the parameters
that describe the conditions required for the production of net
energy in a plasma. The basic requirements are that a reacting
plasma of sufficient density must be confined at high temperature
for a time which allows enough reactions to occur for net energy
production. In 1957, the British physicist, J. D.^Lawson, devel-
oped a simple relationship for the energy balance in a plasma
volume (ERDA, 1975a). Under the assumption that the energy from
the reactions can be recovered at about 33 percent efficiency, the
product of the plasma density and confinement time must be at least
10l* sec/cm3 at an average plasma temperature of about 10 KeV.*
This is the so-called Lawson criterion for minimum conditions re-
quired for a successful fusion reactor (Tuck, 1971).
To date no single device has achieved the simultaneous con-
ditions of temperature, density, and confinement times necessary
for net energy production. A number of machines of different types
have, in recent years, shown a rather steady advance toward the
region of success. For example, some specialized devices have ex-
ceeded the fusion ignition temperature in a plasma for sub-second
durations. Others, such as tokamaks and their near relatives have
pushed the density-confinement time product to about 10 sec/cm ,
but at temperatures of about 1 KeV as opposed to the 10 KeV neces-
sary for ignition of the D-T reaction (Nucleonics Week, 1975).
A statement by R. L. Hirsch, of the U.S. ERDA provides a cap-
sule summary of the present state of development (Hirsch, iy/bj.
Practical generation of fusion energy requires
that the fuel plasma be held at a specified
* The value of 101If sec/cm3 is for the Deuterium-Tritium, or D-T
reaction, which is the most energetically-favorable fusion reac-
tion. The value would be higher for other reactions; for example,
about 1016 sec/cm3 for the D-D reaction (Chen, 1974).
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density and temperature for sufficient time to
result in the desired energy release. In mag-
netic confinement fusion research, minimum
practical plasma densities were first realized
in 1953, minimum temperatures in 1962, and ade-
quate confinement in 1969.' It is necessary
that all three of these conditions be achieved
at one time in one experiment in order to have
a meaningful fusion rate take place. At present
only combinations of two can be achieved be-
cause sufficiently large magnetic "bottles"
have yet to be built.
h^T current ^atus of key parameters achieved
break even n ^ With minimui" requirements for energy
concentua?'deSanf r£n' ValU6S are shown for typical power plant
operation an f^n ^ are currently at least 15 tokamaks in
1975? pVesPnt S!S- r cons^uction in .various countries (Prevot,
1975J. Present devices are fairly small (with manor radii on the
order of two meters or less) , operate with conventional magnets
delg^S
magnets' and
DlasmanDhvsicsyintoS^ Q^rimf^s *re Coving from an emphasis on
plasma physics into the state where engineering development and
technology problems are under active investigation P
Limitations to Development
ther development J^JS1"1 5reaS-haVe been identified in which fur-
overcomrbefo?e T«rarqUlred °r in which Present problems must be
overcome betore lare power systems will be feasible
"
n .
summarlzes ">e major prolems
a. Plasma Engineering
heating and^eac??™**?'^ !t1:einable via P^sma current ohmic
tain ?he fusion r^? "^ fbsorPtion are not sufficient to sus
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TABLE 1. PLASMA PARAMETERS IN TOKAMAKS
CURRENTLY
ACHIEVED
MINIMUM FOR
ENERGY BREAK-EVEN
TYPICAL FOR LARGE
POWER PLANT CONCEPTUAL
DESIGNS CRIBE, 1975)
plasma
Density
(elec/cm3)
6 x 10
1*
10
10llf to 1015
Temp.
(KeV) F 1.6 (b)
5 to 10
20 to 40
Article
Jr?nfinement
Time (Sec) 0.1
.01 to 1
10
(al
Mass. Institute of Technology ALCATOR experiment (Hirsch, 1976).
(b)
French Tokamak Experiment (TFR)(Kinter, 1976)
(c)
Soviet Union T-10 tokamak (Nucleonics Week, 1976).
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"purity" Problem is the necessity for maintaining plasma
purity , i.e., minimizing the number of foreign parasitic non-
duct a^ha°LS Jn ihS ?ia^a' The i-P-ities include Reaction p?o-
chamberPwanfh1Cle? (He .lons) and i°ns knocked off the vacuum
Fuel ions in iLlntraKtl0nS with energetic plasma particles.
order ?Smaiiia?rV° Continuously supplied to the plasma in
for this ?hrnnai ^ reaction rate. Some tokamak designs provide
pole to inject8small SaT ^^ ^ inJecti™, while^thers pro-
ket around tZl hf ^ solid pellets, or provide a fuel gas blan-
miniSSS impu?ities"ng ^^^ tO simul^neously fuel the plasma and
scaled to large?°vn?mainS.t0 be demonstrated that plasmas can be
as S^nTriTS t\attain the Power levels appropriate for
"de0r' Wlth°Ut introducing destructive
b. Tritium Technology
pecked to
10 7 Ci
'
revi us
-
ilo i
kilograms. At approximately
"11? T' 8r?ater ^h
control requirements
Structural -terials and sys-
c.
Material Radiation Damage
heavy fluxeTo? ^^v" °f/ D"T fusi°n reactor "ill be exposed to
the vacuum wa?l Ld ^ ?RS * This includes structural materials,
irom the full fluence by a neutronically-inert shield such as car-
d.
region
Blanket Design
blank^t generally includes the
ld WindlngS 1
serves several
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functions. It provides for neutron moderation and energy removal.
Tritium is generated by neutron reactions in a suitable material,
e.g., Li-6, which is generally a constituent of the main ^olant.
The blanket region must also shield the magnetic field coils from
the neutron and gamma ray fluxes. Cooling must be provided for
the vacuum or first wall, which must be designed to withstand se:
vere thermal and radiation fluxes in addition to its vacuum barrier
function.
In most conceptual designs, the main coolant is a liquid
metal such as a lithium compound. This means that an electrically-
conducting fluid must be pumped through the strong magnetic tieia
surrounding the reactor. Considerable power must be available for
pumping the coolant and cooling passages must be designed to mini-
mize cross-field flow (Lidsky, 1972). The understanding of turbu-
lent liquid-metal heat transfer in the presence of strong magnetic
fields is important for blanket design. However, the current under
standing is incomplete due to the lack of appropriate experimental
data (Lykoudis, 1975). The use of a gaseous coolant would elimi-
nate the problems associated with liquid metal coolants, mis ap-
proach however, would increase the difficulty in attaining a good
tritium-breeding ratio which is associated with liquid lithium.
The problems of blanket design are severe; a number of
designs incorporating a variety of materials have been examined.
VACUUM WALL COOLANT
PLASMA REGION
VACUUM WALL -
REFRACTORY METAL—
MODERATOR - BULK HEAT
TRANSFER & TRITIUM BREEDING-*
THERMAL & RADIATION SHIELD-
MAGNETIC FIELD COILS
Figure 3. Schematic of Tokamak Fusion Reactor
- Blanket Zones
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Many of the unresolved questions must await the construction and
operation of actual test reactors,
e. Magnet Technology
Attainment of the necessary magnetic field strengths for
fusion reactors is feasible through the use of super-conducting
magnets. However, there is no experience in constructing, handling,
and operating magnets of the size required for a fusion reactor
with a plasma chamber several meters in diameter (Darvas, 1975).
The structural requirements to withstand the stresses generated by
the magnetic forces are imposing.
f. Fast Electromagnetic Energy Storage
Several tokamak reactor designs operate in the pulsed
mode. This requires the ability to shift large amounts of energy
in very short times, e.g., on the order of seconds, into magnetic
field windings. This precludes the use of line power directly,
hence rapid inertial or electromagnetic storage systems must be
used. Such systems need to be developed in order for large pulsed
systems to be operable (Darvas, 1975).
In addition to the technical problem areas summarized
above, economic and institutional constraints will control the im-
plementation of fusion power systems into the national energy grid.
An overall constraint on the development necessary to
demonstrate the feasibility of large fusion power systems is cost.
If the technical problems are overcome, but capital or operating
costs of the resulting power systems are not competitive with other
energy technologies, the designs will be unacceptable. Successful
implementation of fusion power will require the development of an
industrial base to produce fusion power systems as the technical
development problems are surmounted (AEG, 1973).
Research and Development Support
The support for fusion power research and development current-
ly comes primarily from the Federal government, administered by the
ERDA. During the fiscal years 1951 through 1974, the ERDA prede-
cessor, the Atomic Energy Commission (AEG), spent $544 million on
magnetic confinement research (GAO, 1975). Table 2 summarizes mag-
netic confinement funding and projections for the fiscal years 1975
through 1980. An additional $116 million has been spent by the AEC
on laser fusion research and development in the period, 1963-1974
(GAO, 1975). The 1975-1980 projections for laser fusion funding
are about one-third of the ERDA funding for magnetic confinement
research and development.
The major ERDA-funded magnetic confinement research and devel-
opment efforts are carried out primarily at four laboratories.
10
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TABLE 2. ERDA-CONTROLLED THERMONUCLEAR RESEARCH
FUNDING PROJECTIONS 1975-1981 (MAGNETIC CONFINEMENT PROGRAM
ONLY--DOES NOT INCLUDE LASER FUSION, DOLLARS IN MILLIONS) t
FY-75. FY-76 ^Period™ FY-77 FY-78 FY-79 FY^SO FY^Sl
105* 192** 62 360 471 567 559 634
From ERDA, "Fusion Power Program Research and Development Program
Projections, May 1975, draft.
* Excludes $6 M deferral.
**Estimate
11
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These are: Lawrence Livermore Laboratory (LLL) , Livermore Cali-
fornia; Los Alamos Scientific Laboratory (LASL) , Los Alamos, New
Mexico; Oak Ridge National Laboratory (ORNL) , Oak Ridge, Tennessee;
and the Princeton Plasma Physics Laboratory (PPPL) Princeton New
Jersey. The remainder of the program is carried out at a number of
private firms and universities. About 75 percent of the current
magentic confinement research and development program is conducted
at the four ERDA laboratories. Additional fusion research and de-
velopment support is provided by private concerns, electric utili-
consortia) ' and the Electric Power Research
The General Atomic Company has supported a fusion reactor de-
velopment program for a number of years and is now receiving major
ERDA support for the design and construction of a large non-circular
cross section tokamak experiment. Westinghouse Electric Company
?5n«irSa 1X1 ^ PPPi Pr°2ram and United Aircraft Corporation
supports some in-house fusion research and development. Estimates
*L ?6 arUnt of support for magnetic confinement fusion research
and development from sources other than ERDA are difficult to ob-
«ni™^ 5 ?' r5S estlmated that other governmental agencies
«S5P!J ! ^ari°SSieuf2rts ^ the level of about $2 million per year,
o? «Sfi %7 6M?-al Y£n Private-industrial effort was in the range
II V V $LnAUl0nJAEC' 1973)* A recent estimate of the total
L^r^ * £ ^Pendl?f? in^uding other government and private
years (Katz,197^n $ " $2° million Per ^ for the next five
nroara^h tfcfP^i tO - forei8n fusion research and development
programs, the following summary is included (AEG, 1973):
Abroad, there are major fusion programs in Europe,
the Soviet Union, and Japan. The European program
is now er EURATOM and is about twice
the U.S. The Japanese effort
°f the U'S* Program, while that
"
and laser fusion.
The USSR has about 50 mai CT
and plasma research experiments in operas on
1 - •*• ° •*• ** civ4.viJLL..lUil. L.IJ.G OO~
viets have an outstanding theory program In mae-
netic confinement research there hasbeSn excellfnt
international cooperation for many years In laler-
theX°rob ere ±S g°°d COIIll!!Vnication on portions of
sion in other parts.
12
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Finally, the ERDA-DCTR objectives and strategy for fusion
power development, as recently announced, are given (ERDA, 1975b):
Objectives
Near-Term (-1985) : Produce reactor level hydrogen
plasmas. Produce substantial
quantities of thermal energy in
the First Fusion Test Reactor
using Deuterium-Tritium fuel.
Mid-Term (-2000) : Produce electrical energy in
substantial quantities in two
Experimental Power Reactors
between 1985 and 1990. Operate
commercial scale Demonstration
Power Reactor (1997).
Long-Term (+2000): Begin supplying a fraction of
the Nation's electrical energy
demand.
Strategy
A series of progressively larger experimental devices
will provide needed knowledge of fusion plasma physics
and engineering under prototypical fusion reactor con-
ditions. This will permit an evaluation of the differ-
ent types of fusion systems and serve as the basis for
the design and operation of fusion power reactors.
A combination of industrial, academic, and National
laboratory resources will be used with funding sup-
port from EPRI and utility consortiums, where possible,
to expand the scope, hasten the pace, and prepare the
technology for full commercialization.
All subprogram efforts are interrelated with the goal
of supporting a successful fusion demonstration by the
most promising concept (currently the tokamak) and al-
so to continue to develop feasible alternatives as
backup.
POTENTIAL SIGNIFICANCE TO ENERGY SOURCES
The main attraction of fusion power is that the fuel supply i.s
essentially inexhaustible. Deuterium occurs in sea water in a ratio
°f one to every 6,500 hydrogen atoms (Lidsky, 1972). Deuterium se-
paration is relatively easy and inexpensive. Tritium is produced
by neutron capture reactions in lithium, of which there appears to
be an adequate supply for a full-scale fusion power economy (Braams,
1975).
13
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£U6i re(luir5m«its are on the order of two kilograms
ol gSra fus^'
*°
in
a ceranpv Power Plant construction depend, to
In addition In * °n ^6 Pfrticular materials used in chosen designs
and coicre?e wMrrf 10nal mat!rial requirements, such as steel
liSh? Ufllfl £• * expected to be similar to conventional
of8a plan? a1?ul?nsc^Ct?r ^ and £°Ssil Plants in the balance
on Lme SateJifV r^nf US1?n economy could pose severe demands
Sn a?e rela?^i v C6S ' /°r exa«Ple, He, Be, Nb , Cr, Mn, and
^ use of large
14
-------
TECHNICAL DESCRIPTION OF CONCEPTUAL TOKAMAK POWER REACTOR
The details of large fusion power systems will be drawn from
the recent Wisconsin and Princeton designs. These represent the
most complete and best documented of the tokamak power system de-
signs currently available.* Referring to the schematic m Figure
4, the main features of a conceptual tokamak power system are pre-
sented. It is convenient to discuss the power plant features in
terms of the following systems: plasma, divertors fuel system
blanket and shield, magnets, cooling system, and steam generators.
Each system will be described in terms of its purpose and
some designyparameters will be presented. Where Appropriate com-
parisons to LWR design parameters will be made. In many instances,
such comparisons will be of limited value due to ^e inherent dif
ferences in fission and fusion physics and the asso"^efr£^nd{red
requirements for each type of power system. It must be acknowledged
tha-t the present discussion i/limited to a very cursory examination
Further, the design reports under Consideration in many cases do
not include details of design and operation of all the systems as-
* This is to be ^din conceptual
sociaed with a ower plant This is to be ^^Jjin concepua
designs which, in many cases, can only discuss alternatives or pos-
sible ranges of design parameters.
*
Both designs use conventional steam turbine plants to generate
electricity.
The duty cycles of both machines are similar in that they are
l
* It should be noted that these designs represent very ^rly at-
tempts to describe systems that may be built 20-30 years in the
future Actual design configurations can be expected to be con-
siderably different, as the technology progresses and new alter-
natives are evaluated.
15
-------
TRITIUM SEPARATOR
WATER
HEAT EXCHANGER
\
n
GENERATOR
ALTERNATING
CURRENT
Figure 4. Tokamak Fusion Power Plant Schematic
16
-------
TABLE 3. TOKAMAK REACTOR REFERENCE DESIGN FEATURES
Power: MWt
MWe (net)
Net Electrical
Efficiency (%)
Fuel Cycle
Toroidal Vacuum Chamber:
Major Radius (meters)
Minor Radius (meters)
Vacuum - First Wall
structural Material
Shielding Materials
Coolant
Moderator--Tritium-
**eeding Material
Magnets: Super-
Conducting Material
power Cycle
Plasma Burn Cycle:
fiurn Time (min)
Recovery Time (min)
Flywheel
UWMAK-I
4,660
1,473
32
(D-T) Li
13
5
316 - stainless
steel
Pb, SS, B^C
Lithium
Lithium
Nb Ti + Cu
Li-Na-Steam
90
6.5
Sodium storage
loop
PPPL
5,305
2,030
38
(D-T) Li
10.5
3.25
PE-16 - austenitic
nickel alloy
high-density concrete,
Pb, B, polyethelene
Helium
(LiF)2BeF2 eutectic
mixture (flibe)
Nb3 Sn
He-Steam
100
3
flibe diverted from
blanket
17
-------
In order to provide continuous electric power output, thermal fly-
wheels are incorporated. During the plasma recovery phase, energy
to the steam generators is provided by a special reservoir of cool-
ant.
REACTOR SYSTEMS
Plasma and Burn Cycle
Reference design plasma characteristics are summarized in
Table 4. Operating cycles of these machines are characterized by
long pulses, i.e., the burn time is much longer than the residence
time of the plasma particles. Consequently, fuel ions must be con-
tinuously supplied, and the reaction products and unburnt fuel re-
moved to maintain plasma conditions during the burn cycle. Each
machine accomplishes plasma management and burn cycle control in a
slightly different manner. The Wisconsin Tokamak burn cycle uti-
lizes neutral deuterium and tritium beam injection following plas-
ma current (ohmic) heating to achieve plasma ignition conditions;
whereas, in the Princeton design, ignition is achieved by ohmic
heating and D-T ice pellet injection. During the burn phase, D-T
pellet injection is used to maintain the plasma in both machines.
After the burn phase is terminated, due to a number of factors
including resistive dissipation of the plasma current and build-up
of plasma impurities, the current is shut down, and the plasma
chamber evacuated. Burn cycle times are given in Table 3. During
the recovery phase, the field coils are re-energized, fresh D-T gas
is injected, and the cycle is re-initiated.
Divertors
As stated by Ribe (Ribe, 1975):
The basic objective of a divertor is to prevent
particles diffusing out of the plasma from hit-
ting the first wall and to provide a means of
removing alpha particle "ash" and impurities,
while maintaining a steady throughput of fuel
from injectors. The divertor also provides a
shield for the hot plasma against impurities
released from the wall, or flowing back from the
divertor chamber.
The divertors consist of a system of specially-shaped second-
ary magnetic fields at the plasma periphery, baffles and flow
channels, and vacuum pumps and manifolds. Particles which diffuse
outward from the active plasma zone through the main confinement
magnetic field are channeled by the divertor poloidal field to par-
ticle collectors. Vacuum-pumping systems sweep the diverted mate-
rials to gas cleanup systems for tritium separation and gas disposal
18
-------
TABLE 4. REFERENCE DESIGN PLASMA CHARACTERISTICS
D + T Ion
Density (cm"3)
Ion Temperature
(KeV)
Average Ion Confinement
Time (sec)
Plasma Volume (m3)
Plasma Current (Amp)
Plasma Power
Density (MW/m3)
UWMAK-I*
0.8 x 10llf
11
14.2
6,400
21 x 106
0.8
PPPL
101"
30
3.8
2,190
14.6 x 106
2.4
* UWMAK plasma parameters are given for operation at thermally-
unstable equilibrium.
19
-------
The divertor vacuum system also purges the main plasma chamber
at the end of each burn cycle. The Princeton design uses two to-
roidal vacuum manifolds connected to the inner circumference of the
main plasma chamber. The vacuum and pumping requirements are met
by 96 mercury diffusion pumps. The UWMAK divertor-vacuum system
also uses 96 mercury diffusion pumps with freon and liquid, nitrogen'
cooled traps. Two sets of two divertor slots provide access to the
main plasma vacuum chamber.
Vacuum requirements for the fusion power plants are well with-
in the capabilities of current vacuum technology, i.e., pressures
down to 10"5 torr (1 torr = 1 mm Hg). Design challenges are posed
by the exceedingly large vacuum chamber volumes and large pumping
speeds required. Pumping speeds within the range of multi-million
liters per second are required to purge the plasma chambers at the
end of each burn cycle.
Fuel System
Fuel ions are supplied to the plasma in several ways. At the
beginning of the burn cycle, deuterium and tritium gas is pumped
into the plasma chamber. After the initial fuel charge is ionized,
heated, and ignited, it is proposed to supply makeup fuel ions ei-
ther by injection of neutral beams or solid pellets. During the
plasma burn phase, both conceptual designs under consideration ac-
complish fuel supply by solid D-T pellet injection. During the
ignition phase, the Wisconsin design uses neutral beam injection,
but the primary purpose here is to supply energy to the plasma.
As the plasma burns, deuterium and tritium ions are lost by
fusion reactions and diffusion out of the active plasma region.
The combined loss rate for deuterium and tritium ions in these de-
signs is about 3 to 4 x 106 ions/second. To maintain the plasma,
fuel ions must therefore be injected at this rate. The UWMAK de-
sign proposes to inject D-T pellets of 20 micron radius into the
plasma by electrostatic accelerators, at the rate of 20 x 106 pel-
lets per second.
Tritium is recovered from the plasma exhaust and the breeding
blanket for reinjectipn into the plasma. Principal fuel flows for
the Princeton design are shown in Figure 5. In this design, cryo-
genic fractional distillation is used for gas separation in the
fuel exhaust purification system. The Wisconsin design plasma ex-
haust system traps the gases on the lithium divertor coolant and
divertor vacuum pump traps. The design proposes to use yttrium ex-
traction beds to remove tritium from the divertor coolant and cryo-
genically-cooled charcoal traps for tritium removal at the pump
traps .
Table 5 gives typical material flow rates in the Princeton
primary fuel loop. The argon is used as the source of impurity
ions which are injected into the plasma for burn control and
quenching at the end of the cycle.
20
-------
HD + H2 Ash to
H20 & Impoundment
COOLANT
HELIUM LOOP
BREEDER
SALT f
LOOP
He Ash to Final T2
Control & Exhaust
Figure 5. Principal Fuel Flows in Princeton Tokamak
21
-------
TABLE 5. FLOW RATES IN PRINCETON TOKAMAK
PRIMARY FUEL LOOP (GRAMS/HR)(MILLS, 1974)
COMPONENT REACTOR FEED VACUUM PUMP EFFLUENT
He 29.3
H 0
HD 0.15
HT 0.10
D2 77.1 77.0
DT 219.8 182.8
T2 0 21.5
Ar 252.8 252.8
22
-------
Blanket and Shield
Blanket design requirements are summarized in the Wisconsin
design report as (Wisconsin, 1974):
-- Tritium breeding.
-- Conversion of D-T reaction product kinetic energy into
heat.
-- Conduction of the heat to an electrical generation system.
-- Provision of: vacuum-tight enclosure for the plasma;
passage's for fueling, heating, and diagnostic equipment;
and coolant channels to conduct the heat away.
Details of blanket and shield for the PPPL reactor are shown
in Figure 6. Due to radiation damage, both reactor designs will
require replacement of the first, or vacuum wall, during the re-
actor operating life. The design goal for the Princeton first wall
is five years and the expected life in the Wisconsin design is two
to three years. Both reactors use a blanket made up of modular
segments, like the sections of an orange, in their placement around
the torus. The PPPL blanket consists of 24, and the UWMAK of 12,
identical segments.
"Each module consists of a super-conducting toroidal magnet,
blanket and shield segment, vacuum pumps, coolant headers, and all
the other necessary plumbing and electrical connections for fuel-
ing and heating the plasma" (Wisconsin, 1974). The modular design
is intended to provide ease of blanket systems maintenance by al-
lowing the removal and replacement of an entire module^ which can
be repaired and maintained while the reactor operates with the re-
placement module.
Materials and coolants for the two blanket designs are given
in Table 3 Shielding is incorporated into the blanket for several
purpose!: to,provide protection to blanket materials from thermal
loads; to protect materials and systems from radiation damage; and
finally, tV provide biological protection from the intense neutron,
gamma/and x?ray fields generated in the reactor. In certain areas,
cryogenic magnets are located close to the plasma, e.g., stabilizer
and divertor coils. These systems in particular require protection
from thermal loading and radiation damage. Most of the shielding
is designed with provision for internal cooling.
Magnets
Magnet characteristics for the two designs are presented in
Table 6 Both reactors use super-conducting, helium-cooled magnets
for all'major magnet systems. As indicated in Table 3, the super-
conducting material for UWMAK is copper-stabilized NbTi, and for
the PPPL design, Nb3Sn.
23
-------
-TOROIDAL FIELD COIL AND DEWAR
POLOIDAL FIELD COILS
DIVERTOR
FIELD
COIL
RADIATION .
SHIELD WALL &/
HELIUM COOLING TUBES
VACUUM WALL-
SPECIAL SHIELDING
MAIN
VACUUM CHAMBER
'FLIBE BLANKET WITH HELIUM
COOLING CHANNELS
INSULATING GAP-
POLYETHYLENE
REFLECTOR
Expanded view showing detail
crossection of radiation
shield wall, vacuum
wall and inner
part of flibe
blanket.
He
HeMHe
VACUUMIWALL*
He He He He
CONCRETE/
SHIELD
He
Figure 6. Princeton Tokamak Cross Section Showing
Blanket, Shielding, and Main Field Coils
-------
TABLE 6. MAGNET CHARACTERISTICS
UWMAK
PPPL
Toroidal Field
Magnet System
Number of Coils
Magnet - Coil
WT (MT) each
Max Magnetic Field
Flux Density (gauss)
Flux Density at
Plasma Center (gauss)
Stored Energy in
Coils (Joules--Total)
Coolant
12
900
86,600
38,200
350 x 109
Liquid He Bath
48
130
160,000
60,000
250 x 109
Forced Supercritical
He Vapor
Max Magnet Structure
Design Stress (psi)
90,000
N. A.*
Poloidal Field System
Diverter Coils
(4 pr)
Transformer Coils
(5 pr)(to supply
plasma current)
Diverter Coils
(2 pr)
Vertical Field Coils
(5 pr)(to heat and
ignite plasma)
Control Field Coils
(21 pr) (to maintain
plasma current)
* Not available.
25
-------
The size of the toroidal magnet-coil systems required impose
a number of design problems. These magnets are incorporated into
the modular segment design of the reactor blankets. The bulk and
mass of the coils pose problems in fabrication, handling, and main-
tenance. Mechanical forces induced by the strong magnetic fields
on coil conductors and support structures require careful design
to keep stresses below yield points.
Cooling System and Steam Generators
The three-loop (Li-Na-steam) cooling system for the Wisconsin
design is shown in Figure 7. This design is used primarily to min-
imize tritium and corrosion product transport from the lithium pri-
mary to the steam generators. There are 12 three-loop systems,
including steam generators, which feed two 1,800 RPM tandem com-
pound turbine generators. Condenser cooling is provided by two
conventional units rejecting heat to the atmosphere via mechanical
draft-cooling towers.
Power distribution and coolant loop operating parameters for
the Wisconsin reactor are summarized in Table 7. In addition to
the main coolant system, one three-loop system is used to cool the
divertor coils. This is a relatively low temperature system pro-
ducing steam at 188°C, which is used to drive the main steam gen-
erator feed water heaters. A third cooling system handles energy
removal from the main blanket shield, which must be maintained well
below the melting temperature of lead.
The PPPL cooling system consists of a helium primary and 12
helium-driven steam generators designed especially for this appli-
cation. These high-temperature, gas-cooled fission reactors (HTGR)
represent state-of-the-art technology. Each steam generator, a
super-critical, double reheat unit with a base-mounted, axial flow
helium circulator, operates in parallel with the main helium cir-
cuit. Helium circulates through the reactor at 55 million kg/hr.
An important part of the reactor coolant system in the UWMAK
is the tritium-handling system associated with the tritium-breeding
main coolant. Tritium breeding in the PPPL does not occur in the
main coolant, but in the flibe blanket. These handling and clean-
up systems will be discussed in the section on radioactive waste
systems.
S°roAB?r^n^2^c°F FUSION AND LIGHT WATER FISSION REACTOR
LnAKAL ihRIoTICS
The core of a light water reactor is a heterogeneous mixture
of fuel, cladding, structural material, and water; whereas, the
core or energy-producing region of a fusion reactor is a dilute
mixture of fuel atoms in the plasma state. Control is achieved
in a fission reactor by regulating the number of neutrons available
to initiate fission reactions. Fusion reactor control (or, more
26
-------
Reactor Main Loop Coolant
(4383 MWt)
Toroidal Magnet Coils
Shield
Blanket
Diverter Coil Cooling Loop
(233 MWt)
" from
Diverter
Coils
Reactor Shield Cooling Loop (47 MWt)*
Time averaged values
Figure 7. Wisconsin Toroidal Fusion Reactor Coolant Loops
-------
TABLE 7. WISCONSIN TOKAMAK COOLING
SYSTEM POWER DISTRIBUTION AND OPERATING PARAMETERS
MASS FLOW LOOP TEMP
NUMBER OF Mw-th IN EACH LOOP HOT/ COLD
SYSTEM LOOPS/UNITS EACH (Kg/hr) (°C)
Blanket :
Li Primary 12
Na Secondary 12
Steam 12
Diverter :
Li Primary 1
Na Secondary 1
Steam 1
Shield:
He Coolant 12
H20 12
Turbine Generator:
Turbine Generator 2
Condenser 2
392 2.7 x 106 489/359
392 9.8 x 106 456/336
392 6.3 x 103 404/218
250 3.3 x 10* 325/
250
250 3.2 x 105 188/57
4.2 1.9 x 101* 200/50
4.2 2.9 x 105 N. A.*
840 MWe 3.79 x 106 399/218
(steam)
1,450 9.4 x 107 13.3 (AT)
(Cooling H20)
(4.1 x 105gpm)
OPERATING
PRESSURE
(psi)
670
35
2,000
20
N. A.*
735
N. A.*
1,900
1.72
(Back
pressure)
* Not available.
28
-------
accurately, reaction rate maintenance) is essentially accomplished
by regulating the rate at which fuel atoms are supplied to the
plasma.
Thus, in contrast to a fission reactor core with a fuel inven-
tory of around 100 metric tons (MT) of uranium, a fusion reactor
plasma contains only a gram or two of fuel at any time and fuel is
supplied continuously to maintain the reaction rate. Nuclear en-
ergy in a fission reactor is primarily carried by the fission re-
coil atoms and converted to heat within a few millimeters of the
individual reaction sites in the fuel pins. On the other hand,
fusion energy is primarily carried by high -energy neutrons (about
80 percent of the fusion energy) and converted into thermal energy
in the reactor blanket within several meters of the plasma or
energy-producing zone.
A rough analogy can be made between the core inside the pres-
sure vessel of a fission reactor and the plasma volume of a fusion
reactor. The primary heat transfer surfaces are the fuel pin clad-
ding and vacuum wall, respectively. In addition, the cladding is
the primary container for the reactor fuel; whereas, in the fusion
device, the primary container is the magnetic field surrounding the
plasma. The power density in a light water reactor is around 100
kw/liter of core volume; in the conceptual fusion plant designs,
it is only one or two kw/liter of plasma volume. Hence , the much
larger reactor size is required for a fusion plant of comparable
power output. The analogy between the cladding and vacuum wall as
primary heat transfer surfaces suffers in that the energy flux
through the cladding of a fission reactor is Costly thermal energy;
wherefs the "wall loading", as it is called in a fusion reactor, is
primarily neutron kinetic energy with relatively little thermal
energy transfer ?o the wall itself. In the Princeton design rough-
ly ^percent of the total thermal energy produced is car ried by
alpha particles from the plasma and the W?TityiQ?d? £ractlon
is deposited as heat in the vacuum wall (Mills, 1974).
in fusion reactors must approach 50 to 100
as .
thermal characteristics between the two systems is t hat
fuel has essentially zero heat capacity while **« Wi.™1.?^
light water reactor has a high heat capacity. The result o± this,
z I ct Hb as ftj^^^
e^Rr^
ding Elusion first wall design temperatures are comparable as
they are both constrained by material properties.
Operating pressure in a pressurized water fission reactor (PWR)
core is in the neighborhood of 2,000 psi and results from thermal-
hydraulic requirements. Pressures at the vacuum wall inner surface
in a fusion reactor are very low, on the order of several microns
29
-------
of Hg. The pressure of the plasma itself is in the range of ten
atmospheres, but is contained by the magnetic fields within the
vacuum chamber.
The radiation environment in a fusion reactor is expected to
be more severe than in fission reactors. Neutron flux densities
in the Wisconsin and Princeton fusion designs, for example, are
several times greater than in an LWR. In addition, the average
neutron energy approaches 14 MeV in fusion plasma environs which
produces more radiation damage than neutrons of fission energies.
Table 8 summarizes a number of comparisons between UWMAK-I
(Wisconsin, 1974), the PPPL design (Mills, 1974), and the Combus-
tion Engineering System 80 PWR (Combustion Engineering Company).
Further comparisons will be made.
RADIOACTIVE WASTE MANAGEMENT SYSTEMS
Fusion power plants, like fission plants, can be expected to
produce gaseous, liquid, and solid radioactive wastes. While fusion
plants will, not produce fission products and actinides, tritium and
a variety of activation products will constitute the sources of
radioactive waste. The majority of wastes associated with routine
operation will be produced in the various tritium-handling and cor-
rosion product removal systems in the blanket and plasma exhaust
clean-up systems.
An idea of the magnitude of the tritium-handling problem can
be gained_from an examination of Figures 5 and 8. Figure 5 shows
the principal fuel flows in the PPPL design and Figure 8 shows the
tritium flows and environmental release sources for the Wisconsin
SSJiP\:i F?T examPle? in the Wisconsin design, there are a minimum
of 36 blanket and shield coolant loops from which tritium is re-
covered. In addition, there are tritium recovery systems in the
vacuum systems and in the divertor system cooling loop. Character-
istics of the tritium extraction systems are given in Table 9 for
the Wisconsin design.
Another major source of radioactive waste will be activated
products in the various coolant loops, since these re-
within the zone of significant neutron fluxes in the
Details of treatment and handling systems and procedures
category of waste are not currently available in the de-
ler consideration. The Wisconsin design analysis includes
on of corrosion product formation in the coolant and the
analysis of activation product formation in stainless
r,-.^ ^oi c^ural materials. What is currently lacking is the cou-
th?q ™« reactor (LMFBR) design experience will be of value in
this area.
30
-------
TABLE 8. COMPARISON OF SOME FUSION AND
LIGHT WATER (PWR) REACTOR CHARACTERISTICS
CONCEPTUAL FUSION
REACTORS
PARAMETER
Power (Mwt)
Plasma/Core Vol. (m3 )
UWMAK-I
4,600
6,400
PPPL
5,305
2,190
COMBUSTION
ENGINEERING
SYSTEM 80 PWR
3,800
40
Plasma/Core Fuel
Inventory (Kg)
Power Density (Kw/1)
Heat Transfer Surface
Area (m2) First Wall/
Cladding
Average Neutron Loading/
Heat Flux at First Wall/
Cladding (watt/cm2)
Max Fuel Temp. (°C)
Max First Wall/Cladding
Temp (°C)
Neutron Flux Density -
All Energies (n/cm*-sec)
2.1 x 10"3
0.8
2,830
0.9 x 10"3
2.4
1,563
1.03 x 10s
95.6
6,410
1.1
N.
177
x 108
470
A.*
211
3 x 108
680
8.7 x 101*
57
1,880
350
2.9 x 10
f
(at wall) (core center)
Not available.
31
-------
To offsfte shipment
(or storage on site for special use)
j\
— MAGNET
SHIELD
BLANKET
VACUUM
i BUILDING CONTAINMENT
Figure 8. Tritium Flow in Wisconsin Tokamak
32
-------
TABLE 9. SUMMARY OF UWMAK-I
TRITIUM EXTRACTION SYSTEM CHARACTERISTICS
(WISCONSIN, 1974)*
COOLANT TEMP. EXTRACTION
SYSTEM RANGE °C METHOD
TRITIUM TRITIUM
ACCUMULATION LEAKAGE
PER DAY (kg) Ci/DAY
TRITIUM CON-
TOTAL Na CENTRATION IN Li
OR Li (kg) OR Na ppm (wt.)
TRITIUM
INVENTORY
(kg)
Primary Yttrium Metal
Lithium 283-489 Bed 1.05** 10.1 1.73 x 106
Secondary Yttrium Metal
Sodium 261-411 Bed
Shield
Helium
0
Diverter
Lithium 200-325 Yttrium Metal
Sodium 190-265 Bed 7.4 T + 5.0 D
7.6 x 105
2-x 10'1* 3.4 x 10"
6 x 101*
Diverter Charcoal-
Vacuum 25 Cooled with 0.3 T + 0.2 D
Liquid He 1 x
50-200 Metal Getter 1.1 x 10
-6
low
8.7 x 10"5
0.24
3 x 10"*
Not Applicable
Not Applicable
In Li 8.7
In Beds 1.0
In Na
6.7 x 10'5
In Beds ~0
In Li 8 x 10
In Beds 3.5
In Na 6 x 10
0.3
low
Total:
10.1
13.5
* Based upon thermodynamic calculations; no kinetic considerations.
** At maximum breeding ratio of 1.49
-------
Discussions of solid waste treatment systems are limited to
general discussions of sources of activated components, such as
vacuum and divertor wall replacement, failed pumps and equipment,
and solids or sludges from the various tritium and corrosion pro-
duct clean-up systems. Estimates are made of amounts of solid
waste expected and several modes of handling are discussed.
Gaseous wastes are expected from various tritium leakage paths,
and neutron activation of air and coolant gas contaminants. Both
conceptual designs include the use of containment-building vapor
barriers around the main torus to control gaseous wastes. In fact,
the Wisconsin design includes an evacuated primary containment
building to limit tritium out leakage. No details were provided
of general plant gaseous waste clean-up and treatment systems, in-
cluding building ventilation systems. No consideration was included
for secondary sources of gaseous wastes, such as cooling system off
gassing, and turbine condenser air ejectors.*
A similar situation exists with respect to liquid waste treat-
ment systems. The balance of plant designs do not include inte-
grated liquid waste treatment systems to handle the various expected
sources of liquid radioactive wastes.
In summary, detailed assessment of the environmental radio-
logical consequences of fusion power plants must await the arrival
of complete power plant system design details. Until such details
are available, plant pathways and environmental release points can-
not be identified and environmental release quantities estimated.
It is obviously unreasonable to expect to have the design details
of radioactive waste systems in hand, as the current conceptual
designs are at a relatively early stage of evolution toward actual
plant designs. The discussion of environmental radiological impacts
that follows will therefore be limited by the amount of design in-
formation available.
Based on the examination conducted herein of sources of radio-
active materials produced in these fusion power plant designs, it
can be seen that radioactive waste treatment will require the de-
velopment of major in-plant systems. It is expected that these
systems may equal or exceed in cost and complexity the radwaste
systems in light water nuclear plants of comparable size.
PrTnceLn f^h. h?le? °f*ases are Produced each day in the
thi? ?« 5* i J b*ank^ ^ neutron transmutations. The bulk of
this is He, but significant amounts of N-16, 0-19, F-18, and
F-20 are produced. With the exception of F-18 all have half
lives of less than 30 seconds. TL equilibrium iiJeSS/Sf 110
minute F- 8
— — — -" — — — — —- -v w v •*» irf «.* ^b«_f a JL li ^ C LJ
minute F-18, however, is 6.6 x 108 Ci.
34
-------
SAFETY ASPECTS
As in any engineering design effort, safety considerations
form an inherent part of the design process. An important benefit
of the current conceptual design exercises for fusion power plants
is the early identification of safety and environmental constraints
that must be factored into the design iteration.
The safety design objectives for fusion power systems are
generally the same as for any nuclear power system: to design a
system that can be built and operated in a manner that poses risks
to the public that are within acceptable limits commensurate with
the benefits gained. The primary risk to the public is expected
to be from the exposure to radioactive materials released from the
facilities.
Some safety aspects are dictated by the nature of the fusion
process itself, while others are highly dependent upon the design
of a particular system (Fraas, 1974). For example, from the con-
ceptual designs examined in the previous section, a number of safe-
ty requirements can be identified, and design requirements for
safety systems can then be considered for reactors with these
characteristics. It is equally important to identify and specify
appropriate design safety criteria for fusion power reactors, e.g.,
design basis accidents, or probabilistic-type criteria, in a timely
manner as designs evolve toward actual systems.
In the following discussion, the main emphasis will be on
safety aspects of tokamak systems represented by the Wisconsin
Princeton designs.
and
DESIGN FOR SAFETY
For fusion reactors designed around the D-T reaction, the
major safety design objective is to achieve a design which will
minimize the release of tritium and other radioactive materials
to the environment. Additional potential hazards to members of
the public may be posed by large inventories of materials, such as
lithium, mercury, lead, boron, and in some designs, beryllium
(Kastenberg, 1975).
A brief discussion of some of the safety considerations asso-
ciated with the various systems of the tokamaks follows. They
will be considered under the general headings of reactor, coolant
system-blanket, and structure and balance of plant. The emphasis
35
-------
here is on public safety considerations, but it is noted that the
design for plant personnel safety also poses a challenge. The
problems of worker protection against radiation exposure from
massive tritium inventories and direct radiation from intense 14
MeV neutron and associated radiation fields will pose challenges
to both designers and operational health physicists. In addition,
electrical hazards will be present and intense magnetic fields over
large areas will raise questions concerning safety and biological
ettects. Liquid metal coolants pose in-plant safety problems and
a variety of toxic materials associated with various systems will
be present. '
Overall Reactor Design
An important safety design problem associated with fission
reactors is not present in fusion devices. Fission reactor cores
are designed around the problem of avoiding a nuclear excursion or
excessive reaction rate leading to uncontrollable energy release.
ine problem in a plasma fusion device is to design a system that
will maintain the delicate balance of conditions required to sus-
tain the reaction. The unlikely instantaneous reaction of the en-
tire plasma fuel inventory would be expected to result in only a
minor increase in blanket temperature (Fraas , 1974).*
A major design objective for a power plant is to "minimize the
°f tritium in every sector of the
'1 inCet°n design' these objectives will be pursued by
(Mills,
(a) processing the blanket so as to achieve the lowest
possible concentration of tritium,
(b) employing a contiguous system of barriers (both
metal and ceramic) that are impervious to the
hydrogen isotopes in any form, and
l.cj operating a processible gaseous purse flow sv^tpm
around and throughout thl plant onTcon?Luous
Coolant System — Blanket
desifinen™- £u?ion reactor coolant system-blanket
inven?o?v ullt^l ^ intesrty> and 2> minimum radioactivity
straints such 2J ?5cJ 6£? ?°als and satisfying other design con-
" hlgh temPerat^e operation for maximum
volume ?he ±sihn .^V^s uniformly throughout the plasma
leading to loyally tilt7 °f non:unifo™ P^sma enlrgy overshoots
is beiL inves^iL^g^energ?: deP°sition and vacuum wall damage
ib oeing investigated (Kastenberg, 1975).
36
-------
The need for maximum integrity is obvious, i.e., to reduce
the leakage of tritium and other radioactive materials under nor-
mal and abnormal operating conditions. Minimization of the radio-
activity inventory from neutron activation of materials is desir-
able to reduce the decay afterheat removal problem, to make the
source of radioactive materials for release to the environment as
low as practicable, and to minimize solid radioactive waste handling
and disposal problems.
Preliminary comparisons of afterheat and induced radioactivity
have been made between various fusion reference designs and with
LWR and LMFBR fission reactors (Dudziak, 1975; Conn, 1975), Com-
parisons of fusion blanket operating average power densities with
fission reactor core power densities indicate that fusion reactors
range from 5 to 20 Mwt/m3 of blanket volume; LWR's are around 100
Mwt/m3; and LMFBR designs range from about 300 to 600 Mwt/m3. Ini-
tial shutdown afterheat to operating power ratios for several years'
operation are in the range of 0.5 to 5 percent (depending upon
blanket design and operating history) for the various fusion
designs and about seven or eight percent for a fission reactor.
The post-shutdown decay afterheat removal problem appears to
be much less severe in fusion reactors than for most fission re-
actors . More important than the fact that initial decay heat
ratios are slightly lower is that the relevant afterheat power
densities are lower by a factor of 50 to 100.
In terms of the radioactive materials present and potentially
available for release to the environment as a result of a disrup-
tive accident, fusion reactors at this stage look favorable in
comparison with fission reactors. There are no fission products
or actinides present, of course, and the major species are isotopes
of Fe, Ni, Mn, Co, Cr, V, and Mb, depending upon the blanket mate-
rials, and tritium. Additionally, relatively short-lived isotopes
will be produced in blanket coolants and breeding materials, e.g.
isotopes of He, Li, N, 0, and F, in flibe, and Na-22 and Na-24 in
sodium.
In accident analyses of large power systems, stored energy in
various systems and the potential for its uncontrolled release must
be examined. Major energy sources in the fusion plants are: the
liquid metal or molten salt and gaseous coolants, the super -conduct ing
magnets, the kinetic energy of the plasma, and the energy stored in
cryogenic systems, e.g., liquified helium. Stored energy in coolant
ior the fusion designs under consideration appears to be comparable
J°. an LMFBR, i.e., on the order of 1012 Joules (Postma, 1971).
eri6?of *° fusi°n reactors is the large amount of stored magnetic
for fLZ5;the suPer -conduct ing magnets, between 101J and 10XZ Joules,
• e- Wlsconsin and Princeton Tokamaks .
of Plant
Several features can be identified in the fusion designs under
37
-------
examination, which have a bearing on safety. The structural design
must accomodate the vary large weights of the magnets, blanket
shield, and supports. The modular blanket designs require that
the segments be transportable and capable of precision placement.
For example, each module of the UWMAK weighs over 3 000 MT which
far exceeds conventional crane capabilities and for which special
transport systems will have to be designed. The removal and re-
placement of blanket sections may have to be handled remotely due
to high radiation levels.
Both the Wisconsin and Princeton designs include containment
vapor barriers surrounding the reactors. The Wisconsin design
includes an evacuated enclosure surrounding the nuclear island and
an additional outer containment building, intended to minimize tri-
tium leakage to the environment.
m»,ii JoecW^C°n?in three" 1°°P. cooling system design with the inter-
mediate sodium loop is primarily intended to reduce tritium and
corrosion product leakage from the primary lithium loop into the
steam loop. The use of 12 of these three-loop cooling systems is
S°e) ds radioac?ivS°ii?SriZl
h .
in the event of a catastrophic, disruptive accident involving the
primary lithium coolant system. j.aivuxvj.ng
POTENTIAL ACCIDENTS AND ENGINEERED SAFETY FEATURES
Both the Wisconsin and Princeton design reports include an
S5iS1S ?f,P£ten£lal accidents' An independent study of safe?y
aspects of tokamaks centering on the Wisconsin design is currently
thfTr^ntf Gnbe-f ' l¥*h What £ollows is a b^ef summary of 7
ing tafe^ol thSl'SSigS? '°"e prelimina^ conclusions concern-
A number of events have been identified which could lead to
reJhaSe -f *rjtiuV° thS envi^ment. Under general cate-
*^«Se inCJ -: -,10SS °^ coolant an
-------
As a result of this analysis, an engineered safety system has
been identified. This consists of a building containment system.
similar to those associated with LWR designs. Additional systems,
such as containment atmosphere clean-up systems, were also dis-
cussed.
Probabilistic safety assessments leading to absolute public
risks associated with fusion reactor operations are not feasible
at this time. These methods generally require complete system de-
sign details and knowledge of operating modes (Hafele, 1974). An
evolution in safety assessments beyond the identification of pos-
tulated accident outcomes is currently being pursued. These efforts
include the examination of event sequences and failure modes which
can lead to various accident end events (Wisconsin, 1974; Kastenberg,
1975).
ENVIRONMENTAL IMPACTS OF NORMAL OPERATION
RADIOLOGICAL EFFLUENTS
Tritium, primarily, and additional amounts of activation
products are expected to constitute the main radiological effluents.
From routine operations, it is expected that both gaseous and liq-
uid releases will occur. While expected release rates are unavail-
able, some discussion has occurred as to what practicable design
tritium release rates might be. Inventories of various isotopes
estimated for Wisconsin and Princeton Tokamak designs are summarized
in Table 10 In comparison with light water reactors, the estimated
tritium release rates for these fusion designs are about one to two
responds to a leakage fraction of about 10 of the reactor cooiant
inventory per day (Pigford, 1973).
tlntial'release includl the large nventory of £££„%»„ &
evolved at a rate of about JO moles per o«/ ± reactive
a-c^^ifi^aHS^r.^^-^.
Will 51T^Tlf*5lT* TTl L IT C liJ-L/v » J ^ .___*_ .1.1_ *, 4- 4-1* A -v« A VTlll DO
Tritiu, diffusion rates throughp-e^
materials are not well known. rr / previously estimated
' be attaint (Young, 1976).
39
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TABLE 10. SUMMARY OF RADIOLOGICAL
QUANTITIES ASSOCIATED WITH TOKAMAK DESIGNS
UWMAK
PPPL
Tritium Inventory (kg)
Estimated Tritium
Environmental Release
Rate (Ci/Day)
Fractional Tritium
Leakage of Total
Inventory Per Day
Major Isotope
Inventories in
Reactor Blankets
(Ci)
Estimated Annual Quantities
of Solid Waste (nT/yr)
First Wall Replacement
Annual Equivalent from
Replacement of Blanket
Corrosion Products
Removed from Coolant
13.5
(a)
10
0.7 x 10
Mn-56 (2.0 x 108)
(b)
Fe-55 (6.5 x 108)
Co-58
Mn-54 (1,
(1.4 x 108)
1 x 108)
Co-60 (2.2 x 107)
V-49 (3.1 x 106)
Ni-57 (5.1 x 10s)
31.4
62.6
0.3
2.6
2.3 x 10
-7
F-18 (6.6 x 108)
(c)
Fe-55
Co-57
Mn-54
Co-60
(1
(2
(1
(1
5 x 109)
5 x 109)
5 x 108)
5 x 108)
V-49 (1.7 x 109)
20.7
(from first wall and blanket)
45
(from flibe and
structural material)
Estimates of tritium inventory for UWMAK vary, other reports indicate
higher values than the design report; 15-30 kg (Conn, 1974) and 30-50
kg (Sehnert, 1975) respectively.
First wall inventory only, based on 10-year operation.
Cc) Entire blanket inventory calculated for 30-year operating history.
40
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large potential sources for tritium release as HTO vapor and T
and D-T gas. Another source of gaseous effluent is A-41 from neu-
tron activation of air and cover or off gases in various liquid
metal systems. In the Princeton design, a major potential source
of this isotope is the several hundred grams per hour of argon that
is circulated through the plasma as an impurity for burn control
and quenching.
The question of exposure from direct and scattered radiations
originating from the fusion reactor itself, to members of the pub-
lic in the environs, has not been analyzed in the design studies.
Potential sources are: the large 14 MeV neutron fluxes, neutron
capture gammas, and gammas from neutron activation products. In
the Princeton design, while no numerical estimates were given, it
appears that the equilibrium inventory of N-16 in the flibe is on
the order of megacuries. Depending on the amount of shielding
around the flibe, this could constitute a source of exposure, i.e.,
the air-scattered secondaries from the energetic N-16 decay gammas.
For example, nearest boundary, off-site annual dose rates for the
TFTR from direct radiation have been estimated to be between 3 and
5 millirem per year (ERDA, i975a). This is based on about 300
equivalent seconds of operation at 15 MWt plasma fusion power.
Sources of solid radioactive wastes from routine reactor oper-
ation include: sludges from various coolant clean-up systems,
activated materials and components from wall and blanket replacement
and failed equipment, and miscellaneous contaminated small tools
and clean-up wastes.
In the Wisconsin design, an estimated 1,500 to 2,000 kilograms
of stainless steel is expected to be dissolved in the lithium cool-
ant per year (Conn, 1974). This material will have to be removed
continuously to avoid fouling of heat transfer surfaces. An addi-
tional source of radioactive materials in coolants is from neutron
sputtering of first wall materials into the coolants. It has been
estimated that this mechanism could contribute up to 50 percent as
much radioactivity as chemical corrosion alone (Conn, 1974).
Equilibrium Li coolant corrosion product inventory total radio-
activity is estimated to be about 5 x 107 Ci (Wisconsin, 1975).
This indicates that the specific activity of coolant-borne corrosion
products will be on the order of curies per gram. These materials
from coolant clean-up systems will have to be handled as high-level
radioactive waste requiring biological shielding, and perhaps cool-
ing as well.
A major source of solid waste generated during routine operation
is the activated structural material from periodic replacement of
first walls and blankets. Some volume estimates obtained from the
Wisconsin and Princeton design reports are included in Table 10.
This material will also have a specific activity on the order of
curies per gram, particularly from first wall material (Wisconsin,
41
-------
1975). The solid waste handling problems posed by these materials
are unique to fusion designs in that large volumes of high specific
activity solid wastes are not associated with fission reactor rou-
tine operation. The conceptual fusion designs include provisions
for remote handling of this material via manipulators and hot cells.
It is likely that the volumes of this material that will actually
have to be handled have been underestimated in the design reports,
as it does not appear to be amenable to compaction. The importance
of minimizing blanket and structural radioactivity has been recog-
nized by fusion reactor designers and it is expected that future
designs will improve on the Princeton and Wisconsin designs in this
regard.
While fusion reactors do not require off-site shipment of
highly radioactive spent fuel as with fission reactors, it appears
that shipments of high-level radioactive waste will be required
unless it is decided to locate fusion plants such that on-site,
long-term storage of these wastes is feasible.
It can also be expected that an additional several thousand
cubic feet per year of solid radioactive waste will be generated
from miscellaneous sources as in any large nuclear power plant.
These could include: filters from the heating and ventilation
system, deactivated coolant traps as in an LMFBR (AEG, 1974) ana-
lytical laboratory and liquid waste treatment residues, contami-
nated small tools and parts, clean-up wastes such as towel wipes
and plastic bags, and protective clothing. Additional materials
unique to fusion plants would include: vacuum system traps, tri-
tium clean-up system beds, tritium getters, tritided metals used
to contain tritium fuel, and failed fuel system components. De-
commissioning of fusion power plants at the end of design lifetime
would be expected to present waste disposal problems similar to
those associated with LWR power plant decommissioning.
NON-RADIOLOGICAL IMPACTS AND SITING REQUIREMENTS
The fusion designs examined herein have conventional steam
power cycles with heat rejection to the atmosphere via evaporative
cooling towers. It can be expected that these plants would have
chemical discharges similar to conventional nuclear or fossil
plants of similar size. The thermal impacts could be estimated by
scaling the various material balances for an evaporation cooling
system to the rejected heat quantities for these plants.
Siting requirements for the fusion designs, at this stage,
appear to be similar to those for conventional large LWR's. In
terms of non-radiological requirements, few differences are ex-
pected from any large fossil or LWR plant. Designs have not
evolved to the stage where accident-based siting criteria have
been developed. These designs appear to have radiological impli-
cations quite similar to LWR's, particularly from routine opera-
tions. It is not reasonable to expect that these designs will
have much less stringent siting requirements than LWR's.
42
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REFERENCES
Braams C. M. , "Controlled Nuclear Fusion and its Potential
Contribution to Future Energy Needs" in Transactions of the
European Nuclear Conference, Pans, France, April 21-25,
1975, published by the American Nuclear Society.
Chen, Francis F. , "Introduction to Plasma Physics", Plenum Press,
'New York, 1974.
Nuclear
Vol. 1, pp. 56-69
^cJ^
Designs", Nuclear Technology, Vol. 26, pp. 391-398, August
1975.
Darvas J "Progress in Assessing the Technological Problems of
^"usion .Reaclors", in Transactions of the European Nuclear
Conference, Paris, France, April 21-25, 1975, published Dy
the American Nuclear Society.
i "Radioactivity Induced in a
" Niclea^jreclmolok, Vol. 25,
pp. 32-55, January 1975.
Fraas A P "Conceptual Design of the Blanket and Shield Region
and Related Systems for a Full-Scale Toroidal Fusion Reactor",
Oak Ridge National Laboratory, ORNL-TM-3096, May 1973.
Fraas A P "The Environmental Impact of Fusion Power", presented
at the'i40th Meeting of the American Association for the Ad-
vancement of Science, San Francisco, California, March 1974.
Hafele, W., and C. Starr, "A Perspective on Fusion and Fission
Breeders", Journal of the British Nuclear Energy Society,
Vol. 13, No7 2, pp. 131-139, 1974".
43
-------
Hirsch, R. L., Director, Division of CTR, ERDA, Remarks at the
Second Energy Technology Conference, Shoreham-Americana Hotel,
Washington, B.C., May 12, 1975.
Hirsch, R. L., U.S. ERDA, Letter to Bruce J. Mann, U.S. EPA.
January 14, 1976.
Kastenberg, W. E., et al., "On the Safety Analysis of Some Tokamak-
Type Fusion Power Reactor Concepts", prepared for presentation
at the American Nuclear Society 1975 Winter Annual Meeting,
San Francisco, California, November 1975.
Katz, M. J., U.S. ERDA, Letter to Bruce J. Mann, U.S. EPA. Septem-
ber 11, 1975. F
Kinter, E. E., U.S. ERDA, Statement to Joint Committee on Atomic
Energy, U.S. Congress, Atomic Energy Clearing House, Vol. 22:12,
March 22, 1976. " & '
Lidsky, L. M., "The Quest for Fusion Power", M.I.T., Technology
Review, pp. 10-21, January 1972.
Lykoudis, P. S., and M. Andelman, "Liquid Metal Heat Transfer in
Pipes with Aligned Magnetic Fields", in Transactions of the
American Nuclear Society, 1975 Annual Meeting, New Orleans,
Louisiana, June 1975.
Mills, R. G., ed., "A Fusion Power Plant", Princeton Plasma Physics
Laboratory, MATT-1050, August 1974.
Nucleonics Week, "Fusion Development Has Had an Unexpected Success",
McGraw Hill, November 6, 1975.
Nucleonics Week, "Preliminary Results From the Princeton Large Torus
Fusion Machine", McGraw Hill, April 1, 1976.
Pease, R. S., and A. Schulter, "The Potential of Magnetic Confine-
ment as the Basis of a Fusion Reactor", in Transactions of the
European Nuclear Conference, Paris, France, April 21-25, 1975,
published by the American Nuclear Society.
Pigford, T. H., M. J. Keaton, and B. J. Mann, "Fuel Cycles for
Electric Power Generation", Teknekron Report No. EEED 101
(EPA Contract No. 68-01-0561), January 1973.
Postma, H., "Engineering and Environmental Aspects of Fusion Power
Reactors", Nuclear News, April 1971.
Prevot, F., "Recent Progress of Tokamak Experiments", in Transactions
of the European Nuclear Conference, Paris, France, April 21-25,
1975, published by the American Nuclear Society.
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Ribe F. L., "Fusion Reactor Systems", Reviews of Modern Physics,
*Voi. 47:1, pp. 7-40, January 1975.
Rose D J "Controlled Nuclear Fusion: Status and Outlook",
'Science, Vol. 172:3985, pp. 797-808, May 21, 1971.
Science, Vol. 187, p. 421, February 7, 1975.
Science, Vol. 188, "News and Comment, Laser Fusion: An Energy
- Option, but Weapons Simulation is First", April 4, 1975.
Sehnert, M. , and W. Kastenberg, "On the Determination of Tritium
Inventories in CTR Power Plants", prepared for presentation
at the American Nuclear Society 1975 Winter Annual Meeting,
San Francisco, California, November 1975.
Tuck J L., "Outlook for Controlled Fusion Power", Nature , Vol.
*233, pp. 593-598, October 1971.
U S. Atomic Energy Commission Fusion Energy Subpanel, "Report to
the Chairman, U.S. AEG", October 1973.
U.S. Atomic Energy Commission, "Proposed Final Environmental
Statement- -Liquid Metal Fast Breeder Reactor Program' , WASH-
1535, December 1974.
U S ERDA, "Final Environmental Statement, Tokamak Fusion Test
Reactor Facilities", Princeton Plasma Physics Laboratory,
Princeton, New Jersey, July 1975.
U.S. ERDA, "A National Plan for Energy Research, Development, and
Demonstration: Creating Energy Choices for the Future", ERDA-
48, Vol. 2, 1975.
U.S. General Accounting Office, "Efforts to Develop Two Nuclear
Concepts that Could Greatly Improve this Country's Future
Energy Situation", Report to the Congress, May 1975.
The University of Wisconsin Fusion Feasibility Study Group, "UWMAK-
The U Fusion Reactor Design", UWFDM-68,
Vol. I, March 1974.
The University of Wisconsin Fusion Feasibility Study Group, "UWMAK-
I, A Wisconsin Toroidal Fusion Reactor Design", UWFDM-68,
Vol. II, May 1975.
Young, J. R., Battelle Pacific Northwest Laboratories, private
communication to Bruce J. Mann, U.S. EPA, April 22, 1976.
45
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TECHNICAL REPORT DATA
(Please read Instructions on the reverse before completing)
1. REPORT NO.
ORP-LV-76-7
3. RECIPIENT'S ACCESSION-NO.
4. TITLE AND SUBTITLE
Environmental and Safety Aspects of
Alternative Nuclear Power Technologies
Fusion Power Systems
5. REPORT DATE
May, 1976 (Issue)
6. PERFORMING ORGANIZATION CODE
7. AUTHOR(S)
8. PERFORMING ORGANIZATION REPORT NO.
Bruce J. Mann
9. PERFORMING ORGANIZATION NAME AND ADDRESS
U.S. Environmental Protection Agency
Office of Radiation Programs-LVF
P. 0. Box 15027
Las Vegas, NV 89114
10. PROGRAM ELEMENT NO.
11. CONTRACT/GRANT NO.
12. SPONSORING AGENCY NAME AND ADDRESS
Same as above
13. TYPE OF REPORT AND PERIOD COVERED
14. SPONSORING AGENCY CODE
15. SUPPLEMENTARY NOTES
Summary presented at 1976 Health Physics Society annual meeting
16. ABSTRACT
An examination of environmental and safety issues associated with
advanced nuclear power technologies is being conducted by the EPA Office
of Radiation Programs. Part of this effort has been devoted to a review
of fusion power systems. In the past several years, progress in fusion
power research and development has led to the production of conceptual
power plant designs for several fusion concepts. These have included
both magnetic and inertial (primarily laser fusion) confinement systems.
One of the more promising concepts is the tokamak magnetic confinement
scheme, and several fairly detailed power plant designs based on this
concept have been produced. The tokamak designs prepared by the Prince-
ton and Wisconsin fusion power plant design groups were used as the
basis for the present review.
The present discussion includes a review of the main features of
tokamak power plants with emphasis on radiological aspects. Both in-
plant and environmental health physics implications of the designs are
briefly reviewed. Environmental control and radioactive waste manage-
ment considerations are included.
17.
KEY WORDS AND DOCUMENT ANALYSIS
DESCRIPTORS
b.lDENTIFIERS/OPEN ENDED TERMS C. COS AT I Field/Group
Thermonuclear Fusion
Nuclear Power Technology
Radiological Aspects of Fusion
Tokamak Reactors
Fusion Power Plants
18. DISTRIBUTION STATEMENT
Release Unlimited
19. SECURITY CLASS (ThisReport)
Unclassified
21. NO. OF PAGES
2O. SECURITY CLASS (Thispage)
Unclassified
22. PRICE
EPA Form 2220-1 (9-73)
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