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

-------
      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

-------
               £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

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             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

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         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

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     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

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     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

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             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

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 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

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                   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

-------
                    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

-------
 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

-------
                          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
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     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-
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Kinter, E. E., U.S. ERDA, Statement to Joint  Committee on Atomic
     Energy, U.S. Congress, Atomic Energy  Clearing  House, Vol.  22:12,
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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,
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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,
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Pigford, T. H., M. J. Keaton, and B. J. Mann, "Fuel Cycles  for
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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.
                               44

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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
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                               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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