CELEX: 51987PC0302
Language: pt
Date: 1987-07-24
Title: PROPOSTA DE REGULAMENTO DO CONSELHO que adopta um programa de investigação e formação (1987 a 1991) no domínio da fusão termonuclear controlada#PROPOSTA DE DECISÃO DO CONSELHO que aprova alterações aos estatutos da Joint European Torus (JET), Joint Undertakinhg#RELATÓRIO sobre "Impacte Ambiental e Perspectivas Económicas da Fusão"#(apresentado pela Comissão)

ARCHIVES HISTORIQUES
DE LA COMMISSION
COLLECTION RELIEE DES
DOCUMENTS "COM"
COM (87) 302
Vol. 1987/0181
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 ---pagebreak---      COMISSÃO DAS COMUNIDADES EUROPEIAS
                                                    COM87 ) 302 final
                                                    Bruxelas , 24 de Julho de 1987
               PROPOSTA DE REGULAMENTO DO CONSELHO
       que adopta um programa de investigação e formação
  ( 1987 a 1991 ) no domínio da fusão termonuclear controlada
                                                  '     j /  ..     r
                                                                      Vi'­
                                                                - ■        ■;
                                                                      is
                 PROPOSTA DE DECISÃO DO CONSELHO
             que aprova alterações aos estatutos d a j'Si /
         Joint European Torus ( JET ), Joint Undertakinhg
                              RELATÓRIO
sobre " Impacte Ambiental e Perspectivas Económicas da Fusâo "
                    ( apresentado pela Comissão )
 ---pagebreak---        COMISSÃO DAS COMUNIDADES EUROPEIAS
                                                     COM87 ) 302 final
                                                     Bruxelas , 24 de Julho de 1987
                 PROPOSTA DE REGULAMENTO DO CONSELHO
         que adopta um programa de investigação e formação
    ( 1987 a 1991 ) no domínio da fusão termonuclear controlada
                   PROPOSTA DE DECISÃO DO CONSELHO
               que aprova alterações aos estatutos da
           Joint European Torus ( JET ), Joint Undertakinhg
                                RELATÓRIO
  sobre " Impacte Ambiental e Perspectivas Económicas da Fusâo "
                      ( apresentado pela Comissão )
C0M(87 ) 302 final
 ---pagebreak---                               TABLE DES MATIERES
PROGRAMME FUSION 1987-1991
                                                         Page
A)   EXPOSE DES MOTIFS                                     3
     Annexe   :    Examen des réalisations scientifiques
                   et techniques 1984-86 au sein du
                   programme fusion européen              20
B)   PROPOSITION DE REGLEMENT DU CONSEIL arrêtant
     un programme de recherches et d' enseignement
     ( 1987-1991 ) dans le domaine de la fusion
     thermonucléaire contrôlée                            41
C)   FICHE FINANCIERE                                     49
D)   AVIS DU COMITE SCIENTIFIQUE ET TECHNIQUE             66
     AVIS DU COMITE CONSULTATIF DU PROGRAMME FUSION
 ---pagebreak---                                                                            3.
                                A) EXPOSE DES MOTIFS
I.     JUSTIFICATION
       A l' article 3 de sa décision^ du 12 mars 1985 arrêtant un programme
       de   recherches   et  d' enseignement    dans  le  domaine de    la  fusion
                                                                                !
        thermonucléaire contrôlée     ( 1985-1989 ), le Conseil des Ministres a
       déclaré :
       "Au cours de la deuxième année , le programme sera réexaminé . Sur la
       base de ce réexamen , la Commission présentera au Conseil une
       proposition de révision en vue de remplacer en 1987 le programme
       actuel par un nouveau programme quinquennal ."
       La Commission soumet ci-dessous au Conseil une proposition pour un
       nouveau programme quinquennal couvrant la période 1987-1991 . Le
       réexamen des activités en cours sur lesquelles la proposition se base
        est indiqué dans l' annexe de l' Exposé des Motifs .
        En parallèle avec cette proposition de programme , la Commission
                                                                              \
        soumet egalement au Conseil une proposition pour la prolongation de
        la durée de l' Entreprise Commune JET jusqu' à la fin 1992 ( voir
        Section V) .
      LlLes
       m
              deux propositions sont , du point de vue programmatique ôt
        financier , cohérentes avec la décision concernant le Programme Cadré
        des activités communautaires de recherche et de développement
                                                                   ( 2) 1
        technologique ( 1987-1991 ) prise par le Conseil le
                                                                        J
( 1 ) JO L 83 du 25.3.1983
(2)   JO • « • du « • •
 ---pagebreak---                                                                     4.
II . LA FUSION EN TANT QUE PROGRAMME COMMUNAUTAIRE
     Conformément à des décisions réitérées du Conseil , " le programme
     fusion de la Communauté constitue un élément de collaboration à long
     terme couvrant la totalité des activités entreprises dans le domaine
     de la fusion thermonucléaire contrôlée dans les Etats Membres .     Il
     tend à aboutir , en temps voulu , à la réalisation en commun de proto¬
     types en vue de leur industrialisation et de leur commercialisation ."
     Le potentiel à long terme de la fusion , à savoir un nouveau mode de
     production d' énergie ayant un impact modéré sur l' environnement et
     utilisant un combustible pratiquement inépuisable , justifie la
     poursuite vigoureuse de son développement , quelles que puissent être
     les fluctuations à court terme du prix du pétrole .         La fusion
     pourrait apporter une contribution essentielle à la réduction de la
     vulnérabilité économique , écologique et politique de l' Europe au
     cours du siècle à venir .
     Dès aujourd'hui , la fusion est une technologie de pointe : le JET ,
     les dispositifs spécialisés en construction ou en service dans les
     laboratoires associés et le développement de composants destinés au
     NET sont en eux-mêmes le témoignage d' une haute technologie dont les
     retombées ( en particulier dans le domaine de la technologie des
     aimants supraconducteurs , de la robotique et des systèmes à
     micro-ondes haute puissance ) sont bénéfiques pour d' autres secteurs
     scientifiques et industriels en Europe . Le rôle de l' industrie
     devrait croître sensiblement une fois que le NET en sera à la phase
     de projet détaillé .
     Les raisons principales pour effectuer des travaux R et D sur une
     base communautaire dans le domaine de la fusion sont :
     -    la dimension nécessaire en ressources humaines et financières ,
          qui suggère qu' un tel développement ne pourrait probablement pas
          être effectué au niveau d' un pays ;
 ---pagebreak---                                                                               5.
      -     la durée de l' effort ( allant jusqu' au siècle prochain)
            nécessaire pour arriver à la construction du réacteur ;
      -     l' existence d' un besoin collectif ,       commun à   tous    les   Etats
            Membres ;
      -     la   réalisation     d' un  marché    européen   pour   les   industries
            européennes dans les domaines de haute technologie ;
      -     l' accès , en cas de succès , à un marché communautaire étendu pour
            le réacteur européen ;
      -     de procurer un partenaire potentiel de dimension comparable aux
            3 autres programmes " fusion" mondiaux , favorisant de ce fait la
            coopération internationale dans le domaine de la fusion ;
      -     la qualité du programme fusion européen dont la position de
            premier plan est reconnue dans le monde entier , et auquel la
            Suède et la Suisse sont totalement associées .
      En conséquence , la fusion est en accord avec               les   critères    se
      rapportant aux programmes communautaires de R & D.
III . OBJECTIFS DU PROGRAMME FUSION 1987-1991
      La voie qui conduit aux réacteurs de fusion pour la production
      d' énergie peut être schématiquement et quelque peu arbitrairement
      subdivisée en trois étapes : démonstration de la faisabilité scienti¬
      fique , de la faisabilité technologique et de la faisabilité écono¬
      mique . Actuellement , avec le JET , les Tokamaks de taille moyenne et
      leurs équivalents étrangers , nous nous trouvons encore essentielle¬
      ment au stade scientifique . Le "Next European Torus " ( NET ), qui en
      est actuellement à la phase d' avant-projet , doit constituer un
      dispositif qui devrait confirmer pleinement                  la faisabilité
      scientifique de la fusion dans une première phase et , dans une
      seconde , affronter les problèmes de la faisabilité technologique .
      Dans le cadre de la stratégie du programme " fusion" communautaire
       (JET  et   les  autres   tokamaks    - NET -    réacteur   de  démonstration
      DEMO , ...) les principaux objectifs du programme fusion 1987-1991
      sont les suivants :
      -     établir la base physique et technologique nécessaire au projet
            détaillé du NET ; dans le domaine de la physique et la technique
            du   plasma ,   cela    implique   exploiter   pleinement    le   JET   et
 ---pagebreak---                                                                       6.
           plusieurs tokamaks spécialisés de taille moyenne qui existent
           déjà ou qui sont en construction , et dans le domaine de la
           technologie renforcer le programme en cours sur la technologie
           de la fusion ;
     -     entreprendre le projet détaillé du NET avant la fin de la
           période du programme si la base de données nécessaire existe à
           ce moment-là ;
     -     explorer le potentiel , en vue du réacteur , de certaines configu¬
           rations alternatives (principalement stellarateur et pinch à
           champ inversé ) .
     La proposition de programme a été préparée avec le concours de toute
     la communauté de la fusion , par le biais du système d' évaluation par
     des groupes d' experts de même niveau institués par le comité
     consultatif pour le programme " fusion" ( CCPF) et par le Conseil du
     JET pour le JET .
IV . SITUATION ACTUELLE
     Le programme "fusion" européen a pu être concentré sur la
     configuration la plus prometteuse , le confinement magnétique toroïdal
     et , dans le cadre de cette approche , il permet de garder la marge de
     manoeuvre nécessaire .     Les résultats scientifiques et techniques
     placent l' Europe en tête de la recherche mondiale sur la fusion
     magnétique :
     -     JET est l' expérience de fusion la plus importante au monde , qui
           a atteint ses objectifs initiaux pour les performances de base
           dans les délais et avec les crédits alloués , et l' extension aux
           performances élargies est en bonne voie ; pendant ses premières
           années de service ( début en 1983 ) , il a permis de faire un
           important pas en avant vers la démonstration de la faisabilité
           scientifique de la fusion , produisant déjà une quantité appré¬
           ciable de réactions de fusion dans le deutérium .
     -     les tokamaks européens de taille moyenne contribuent puissamment
           au progrès de la fusion et au succès futur du JET , grâce à des
           expériences avec différentes configurations , à l' exploration de
           nouvelles méthodes de chauffage et au développement de nouveaux
           diagnostics ;
     -     l' Europe vient aussi en tête en ce qui concerne la recherche sur
           les Stellarateurs et les Pinches à Champ Inversé , des configura¬
           tions alternatives au Tokamak ;
 ---pagebreak---                                                                   7.
       l' industrie européenne a construit toutes ces machines (à titre
       d' exemple , plus de 98% en termes financiers des contrats JET ont
       été passés à l' intérieur de l' Europe ) et s' est déjà vu confier
       certains travaux de pointe à long terme .         Sa participation
       devrait faire un bond qualitatif et quantitatif lorsqu' une
       décision sera prise au sujet du démarrage du projet détaillé du
       NET ;
-      le NET se trouve au stade de l' avant projet . Les principales
       spécifications      de   performance   ont    été   provisoirement
       sélectionnées et ont donné un jeu cohérent de paramètres
       utilisés actuellement pour l' optimisation et l' orientation du
       programme technologique ;
-      la bonne exécution du programme technologique est l' une des
       réalisations importantes de ces dernières années . La majeure
       partie des travaux est orientée vers le NET , mais concerne aussi
       des activités à plus long terme . Ces efforts sont concentrés
       dans les domaines des aimants supraconducteurs , du tritium , de
       la couche fertile , de la télécommande , des matériaux , de la
       sécurité et de l' environnement .
En dehors de la fusion magnétique , une activité " de contact " est
exercée dans le domaine de la fusion par laser , tandis que la fusion
catalysée par muons est suivie de près .
L' approche communautaire , qui a permis la création de l' Entreprise
Commune JET ( 1978 ) et de l' équipe NET ( 1983 ), a aussi conduit à la
mise en oeuvre d' une intense collaboration entre laboratoires fusion .
La plupart des Associations travaillent pour une autre Association et
toutes travaillent pour JET et NET par le biais de différents types
de contrats et d' accords . Le programme " fusion" européen a contribué
efficacement à la réalisation d' une véritable communauté scientifique
et technique de petits et grands laboratoires , toujours disposés à
accueillir les nouveaux venus et poursuivant un objectif commun .
Cette situation fait de l' Europe un partenaire attrayant pour la
collaboration internationale , à la fois dans des cadres bilatéraux
 (Canada , Japon , Etats-Unis ) et dans des organisations multinationales
 ( OCDE , AIEA).
 ---pagebreak---                                                                      8.
   Parmi les nombreuses dispositions prises pour garantir la nature
   véritablement   communautaire   du programme  fusion , la mobilité    du
   personnel mérite une mention spéciale : chaque année , plus de 200
   chercheurs ( sur un total de 1200 chercheurs environ ) vont travailler ,
   par le biais de contrats dits de mobilité , en dehors de leurs
   laboratoires pendant des périodes allant d' un mois à un an . Le JET
   représente un cas extrême dans ce domaine : ce projet spécialisé est
   exécuté par un personnel qui a son "billet de retour", c'est-à-dire
   que les organisations nationales se sont engagées à reprendre leur
   personnel après son détachement auprès du JET . Depuis le démarrage du
   projet , environ la moitié de l' équipe est rentrée dans les
   Associations à l' achèvement de ses travaux et a été remplacée par
   d' autres personnes ayant les qualifications nécessaires pour
   entreprendre les nouvelles tâches prévues .
   Un aperçu plus détaillé des activités en cours est donné en annexe .
V. CALENDRIER
   Le calendrier des différentes machines et leurs systèmes de chauffage
   est représenté schématiquement à la figure 1 .
                          Légende de la figure 1
             Plan de développement des principales machines
   Les   différentes  méthodes  de chauffage sont représentées par des
   couleurs différentes :
   noir :     chauffage ohmique ( OH )
   jaune :    injection de neutres (NI )
   rouge :    chauffage par résonance cyclotronique ionique ( ICRH )
   vert :     chauffage par résonance hybride inférieure (LH)
              ou excitation de courant ( LHCD )
   bleu :     chauffage par résonance cyclotronique électronique ( ECRH)
   violet :   ondes d' Alfven (AW)
   L' épaisseur de chaque trait est proportionnelle à la puissance de
   chauffage à travers les fenêtres (1 mm par HW , sauf pour JET où la
   puissance totale est environ 50 MW)
   La phase de construction est indiquée en pointillés .
 ---pagebreak---                    1986    1987      1988              1989      1990                 1991
     JET            JET
Grenoble                   PETUI.A
Fontenay (+ FOM) _ TF: R
                                                       .           ■■■  - -     -
                                       l~   ■ ' · ···
                   TORF-RUPR ;
Cadarache                                                      –
                 K
Garching           ASDEX                                 r
                                                                            :     .
                        ALSDEX- upgrad e
Frascati            FT                                                                –        –
                                     FTU
                          _ DU Έ
Culham
                   COMFASS
                         r
                   TFYTHR
Jülich (+ ERM )  r-
                                   . .    .
                                                                              ■
                                                                                                   ui
                                                                TCA
Lausanne
                                                       TÇV ^                        – -  -  ■ - · –
Garching           W7AS
                             L - .                             _L–_I                    ...
Madrid                      TJII                           -T    -     :-           –
Culham                                                - HBTX  f
                                                              K
Padova                               RFX
Stockholm           EXTR AP
                         I-:
                                                                                            CR86.148
 ---pagebreak---                                                                   11 .
JET   :  Les  résultats   scientifiques  obtenus   ces  dernières  années
indiquent que , afin d' exploiter pleinement le potentiel du projet JET
et atteindre ses objectifs déclarés ( par exemple , s' approcher le plus
près possible des conditions requises dans un réacteur ) en utilisant
au mieux les possibilités de la machine , il sera nécessaire d' ajouter
des équipements complémentaires . Pour cela , il faut plus de temps et
plus de crédits que par le passé . Le Conseil du JET a donc proposé
de prolonger la vie statutaire de l' Entreprise Commune JET , dont
l' échéance était prévue pour le 31 mai 1990 , jusqu' à la fin de 1992 ;
ceci permettrait de garantir une utilisation optimale des équipements
existants et de ceux devant encore être installés , conduisant à une
base plus solide pour le projet détaillé du NET . Parallèlement à la
présente proposition de programme , la Commission présente au Conseil
et au Parlement européen pour approbation ( article 50 du traité
Euratom) un amendement des statuts du JET , en vue de la prolongation
du projet en faveur de laquelle les arguments scientifiques sont
présentés dans ce document .
NET : Conformément à la décision du Conseil de mars 1985 , l' activité
NET s' est ralentie , de sorte que , comme hypothèse de travail , 1990
constitue maintenant la date de décision concernant le projet
détaillé du NET et     1993 / 1994 celle de  la d écision concernant sa
construction . Ces dates sont adaptées au nouveau calendrier du JET
et permettent d' obtenir plus d' informations sur la performance du
plasma au moyen des machines de taille moyenne .
Autres_Tokamaks : Les quatre tokamaks spécialisés de taille moyenne
actuellement en construction au sein des Associations ( Tore-Supra ,
Asdex-Upgrade , FTU et Compass ) entreront en service vers 1988 et
pourront donc apporter d' importantes contributions au projet détaillé
du NET . La construction d' un autre tokamak (TCV en Suisse ) destiné à
explorer les limites de bêta a été récemment approuvée . Le projet
d' une machine compacte à champ élevé en vue de l' ignition ( IGNITOR ,
en Italie ) est également prévu . Les tokamaks actuellement en service
seront pleinement exploités (Textor , Asdex) ou mis à l' arrêt (Dite ,
FT , ...), en fonction de leur potentiel et de la disponibilité
d' équipes de recherche suffisamment fortes .
 ---pagebreak---                                                                       12 .
     Autres   machines   :  En  ce  qui  concerne   les  deux  configurations
     alternatives ,  des machines sont en construction      (W7 AS , RFX)  ou
     projetées (TJII , W7 X), de sorte que le choix du dispositif le mieux
     adapté au DEMO devrait pouvoir être basé en temps utile sur des
     données expérimentales ayant fait leurs preuves ; les machines
     existantes (HBTX . ) seront mises à l' arrêt après complète
     exploitation . Un dispositif plus petit (Extrap , en Suède ), destiné à
     l' exploration d' une conception différente , est en service .
     Technologie : Le programme de technologie est adapté aux nouveaux
     jalons du NET , afin de fournir d' abord la base de données
     technologiques nécessaire aux décisions concernant le NET . Quand la
     décision de commencer le projet détaillé du NET sera prise , un
     programme de R , D & D renforcé , axé essentiellement sur la production
     industrielle et l' essai de prototypes de composants du NET , devra
     être mis en route .
VI . STRUCTURE
     La Commission est responsable de la mise en oeuvre du programme . La
     structure consultative prévoit un seul organisme , le Comité
     Consultatif du Programme Fusion ( CCPF ) , assisté de deux sous-comités ,
     le Comité des Programmes ( CP ) pour les questions concernant la
     physique et la technique du plasma et le Comité de Gestion de la
     Technologie de la Fusion (CGTF ) pour le NET et la technologie . Pour
     l' Entreprise Commune JET , les responsabilités incombent au Conseil du
     JET et au Directeur du Projet . Le Conseil du JET est assisté par le
     Comité Exécutif du JET et peut demander l' avis du Conseil
     Scientifique du JET . Le Programme Fusion sera également soumis à une
     évaluation externe indépendante : en particulier durant la troisième
     année du programme 1987-91 , la Commission demandera une évaluation
     par une Commission d' experts de haut niveau , qui fournira la base
     d' une révision du programme selon le concept du plan glissant .
     Le programme est exécuté au moyen de contrats d' Association conclus
     entre EURATOM et les organisations nationales actives dans le domaine
     de la fusion , ainsi que par l' Entreprise Commune JET , et par le biais
 ---pagebreak---                                                                              13 .
      d' un accord multilatéral concernant le NET . De plus , le Centre Commun
      de Recherches consacre une partie de son programme à la technologie
      de la Fusion : ces activités de fusion sont coordonnées avec le reste
      du programme de technologie via le FTSC . On compte 12 Associations
      réparties dans 10 pays (y compris la Suède et la Suisse ) ; des
      discussions préliminaires sont en cours avec la Grèce et le Portugal
      en vue     de  la  création  de   2  nouvelles   Associations .   L' industrie
      participe par le biais de contrats de développement , ainsi que par la
      construction d' équipements .
      Cette structure est considérée comme bien adaptée aussi pour
      l' avenir , lorsque le rôle des Associations , actuellement orientées
      vers la physique ( et dont les programmes de recherche donnent à
      l' effort européen l' ampleur nécessaire ) sera repris par des
      institutions nationales à vocation technologique et , plus tard , par
      l' industrie .
VII . COLLABORATION INTERNATIONALE
      La coopération internationale a été toujours           très active dans le
      domaine de la fusion . Dans le passé , elle a          fait essentiellement
      l' objet d' accords sur des points spécifiques .        A l' heure actuelle ,
      des formes plus larges et plus substantielles           de coopération sont
      mises en oeuvre ou explorées .
      -     Accords-cadres bilatéraux
            Canada : déclaration commune d' intention ( décision du Conseil du
            20.1.86 ) signée le 6 mars 1986 .
            USA : Accord de coopération ( décision du Conseil du 15.09.86 )
            signé le 15 décembre 1986 .
            Japon    :  un  projet   de   décision   du   Conseil   autorisant    la
            Commission à négocier un accord de coopération a été proposé par
            la Commission au Conseil le 26 février 1987 .
      -     Conventions_d^exécution_dans_le_cadre de 1^AIE_(0CDE)
            Tokamaks : TEXTOR , signé le 5.10.1977 , pour une durée de 15 ans ;
                         ASDEX et ASDEX-UPGRADE , signé le 31.7.1985 , pour une
                         durée de 10 ans ;
                         LES TROIS GRANDS TOKAMAKS ( JET , JT-60 et TFTR) ,
                         signé le 15.1.1986 , pour une durée de cinq ans .
 ---pagebreak---                                                                           14 .
           Configurations alternatives : STELLARATEURS , signé le 31.7.1985 ,
                       pour une durée de cinq ans ;
                       FINCH A CHAMP INVERSE , en préparation .
           Technologie    de  la   fusion    :  "LARGE   COIL  TASK",   signé   le
           6.10.1977 ,
                       l' installation est en service .
                       MATERIAUX DE FUSION , signé le 21.10.1981 : annexe I
                       supprimée , durée de l' annexe II , dix ans .
      -    Coopération dans le cadre de l' AIEA
           Participation d' EURATOM , avec les trois autres grands programmes
           de fusion (Japon , USA , URSS ), aux séminaires d' INTOR depuis
           1978 .
      -    Groupe_de_travail _" fusion"_ ( groupe " technologie , croissance et
           emploi " - sommet de Versailles )
           Consultation entre les programmes " fusion" dans le cadre de la
           participation au sommet économique , notamment en ce qui concerne
           l' étape suivante ( Next Step ).
      -    Initiative de coopération quadripartite concernant un réacteur
           thermonucléaire expérimental international ( ITER) sous les
           auspices de l' AIEA
           On explore , au niveau technique , la possibilité pour les quatre
           grands programmes " fusion" mondiaux (CE , Japon , USA , URSS ) de
           coordonner leurs efforts en vue d' un but spécifique : établir
           vers 1990 , par un effort de collaboration des quatre parties
           fournissant à statut égal des contribution égales , un projet
           conceptuel d' ITER , et coordonner les activités de recherches de
           support . Un groupe technique de travail a été constitué afin de
           préparer en 1987 des propositions concrètes sur les objectifs
           détaillés d' ITER et sur les modalités d' organisation de la phase
           de projet conceptuel 1988-1990 . L' activité du groupe NET , qui
           continuera comme prévu jusqu' à ce qu' une solution internationale
           possible offrant des garanties convaincantes ait été trouvée en
           ce qui concerne le Next Step , pourrait former le noyau d' une
           telle collaboration .
VIII . ENVELOPPE FINANCIERE
      La présente proposition de programme ne concerne que le JET et le
      Programme Général . Les activités de fusion du CCR , qui du point de
      vue technique et scientifique sont entièrement intégrées dans le
      programme fusion global , sont cependant régies par une autre Décision
      de Programme .
 ---pagebreak---                                                                           15 .
        En argent courant ( on a considéré qu' à partir du 1.1.1985 l' inflation
        représentait 4% par an ) , le montant des crédits communautaires requis
        pour la proposition de programme 1987-91 (à l' exclusion du CCR , de la
        Suède et de la Suisse ) est estimé à :
        Programme Général              533 MioECU
        JET                            378 MioECU ^ 1 ^
                  Total                911 MioECU
        Une ventilation des ressources entre les différentes activités est
        faite au tableau 1 .
        Cette estimation est déduite de l' hypothèse sur laquelle repose la
        présente proposition , à savoir que le progrès scientifique et
        technologique sera de nature à permettre le démarrage de la phase de
        projet détaillé du NET avant la fin de la période du programme (voir
        paragraphes III et V ). La décision de commencer le projet détaillé du
        NET aura une importance majeure et la Commission fera une proposition
        dans ce sens au Conseil en temps utile .
        Le tableau suivant montre la ventilation entre JET , le Programme
        Général et le CCR des nouveaux crédits prévus pour la fusion dans le
        contexte du Programme Cadre 1987-91 , ainsi que les crédits reportés
        des programmes en cours .
  MioECU                Nouveaux crédits     Crédits reportés    Dotation
                        correspondant au     de 1985-89          totale pour
                        Programme Cadre                          la période
                         1987-91                                  1987-91
Programme Général            362                   171               533
JET                           169                 209                378
TOTAL - PROGR . FUSION       531                  380                911
CCR                            60                    15               75
TOTAL                        591                  395                986
 ( 1 ) Voir note N° 8 en bas de la page 18 .
 ---pagebreak---                                                                          16 .
      A l' article 4 de la proposition de Règlement du Conseil , il est
      stipulé que la Décision du Conseil concernant le programme 1985-89
      est abrogée avec effet au 1er janvier 1987 . En référence à cet
      article , la Commission fait observer que les montants ayant été
      autorisés aux postes correspondants des budgets 1985 et 1986 au titre
      de la Décision 85 / 201 /Euratom et qui , au 1er janvier 1987 , n' étaient
      pas encore engagés ou étaient engagés mais non encore liquidés ,
      seront utilisés pour l' exécution du présent programme .
IX . EFFECTIFS
      Le nombre d' agents EURATOM autorisé par la décision antérieure du
      Conseil est :
            165 employés temporaires pour JET
            105 fonctionnaires pour le Programme Général .
      Pour la période 1987-91 , aucune modification n' est proposée pour le
      Programme Général , mais un renforcement du personnel JET ( 191 au lieu
      de 165 ) est indispensable afin de rendre possibles la mise en oeuvre
      et la pleine exploitation des améliorations techniques pendant la
      durée prévue du projet . Lorsque le NET passera de la phase
      d' avant-projet à la phase de projet détaillé , de nouvelles
      propositions seront présentées au Conseil .
X.    CONCLUSION
      En raison de ses importants objectifs , des excellents résultats
      obtenus , de son intérêt technologique et de son caractère
      communautaire absolu , la fusion continue à constituer l' un des
      programmes de R & D les plus importants bénéficiant du soutien de la
      Commission . Ainsi qu' il a été annoncé au moment de la décision
      concernant le programme 1985-89 et qu' il a été noté par le Conseil ,
      la Commission a exécuté le programme en 1985 et 1986 dans les limites
      financières indiquées dans la proposition de programme 1985-89 . La
      Commission estime que le niveau de financement indiqué dans sa
      proposition actuelle est nécessaire pour sauvegarder l' élan du
      programme qui est entièrement axé sur le "Next Step" et pour tenir
      compte de l' adhésion des nouveaux Etats Membres en 1986 et de la
      participation croissante de l' industrie . Selon le concept du
      programme glissant , la Commission établira en 1989 une proposition de
      révision de programme , en vue d' un nouveau programme quinquennal
      commençant le 1.1.1990 .
 ---pagebreak---                                                   (1)
        Tableau       Participation communautaire     pendant la période 1987-1991 , en MioECU , en argent courant ( 2 )
NET
    Salaires , indemnités , missions                               27
    Travaux au sein des Associations                               10
    Support par l' Association h3te                                15
    Conception industrielle                                        28
                Sous-total                                         80 - 3 ( 3) -          77
TECHNOLOGIE
    Travaux de base au sein des Assoc .                            65
    Actions prioritaires                                           35
    RD /D industriels                                              37
                Sous-total                                        137 - 13 ( 3) *=       124
PHYSIQUE ET INGENIERIE DU PLASMA
    Dépenses courantes au sein des Assoc .                        231 (4)
    Actions prioritaires normales                                  26  5
                                                                   93 ^;
    Grandes machines avec chauffage
    Support au JET ( article 14 )                                  10
    RD /D industriels                                               9
                Sous-total                                        369 - 67 ( 3) -        302
MOBILITE/GESTION ^ (y compris bourses et évaluation)                                      30
Total GENERAL PROGRAMME                                                                  533 ( 7 )
JET                                                               425 - 19 ( 3) - 28 -   378 (8)
GRAND TOTAL                                                                              911
          CCR (non inclus dans la présente proposition )                                                                 a
          Activité de fusion totale                                                      986
 ---pagebreak---                                                                        18 .
Notes concernant le tableau 1
(1)  Sans la Suède et la Suisse , mais y compris l' activité dans les
     nouveaux Etats Membres .
(2)  A partir du 1.1.85 , une inflation de 4 % par an est prise en compte .
(3)  Crédits engagés en 1985-86 pour 1987 .
(4)  Y compris les crédits pour , éventuellement , une nouvelle machine à
     Madrid .
(5)  Y compris les crédits pour démarrer , éventuellement , la construction
     d' un nouveau stellarateur W-VII.X à Garching .
(6)  Y compris les crédits nécessaires pour financer au niveau de 42 % le
     personnel de la Commission dans les Associations .
( 7) Auquel doit s' ajouter tout solde positif provenant des contributions
     de la Suède et de la Suisse dans le cadre du programme hors-JET .
(8)  La totalité des contributions des Membres , requise pour financer les
     paiements du JET pendant la période 1987 à 1991 du programme , est
     estimée à 531 MioECU (voir " Plan de développement du projet et
     estimation du coût du projet ", tableau 16 de l' annexe , approuvés par
     le Conseil du JET le 26 mars 1987 ). De ce montant 80% , équivalant à
     425 MioECU , sont financés via le budget communautaire . De cette
     dernière somme , 19 MioECU ont été engagés par la Commission avant
     1987 . Le reliquat , s' élevant à 406 MioECU , sera financé comme suit :
     .     378 MioECU de la dotation de programme pour JET ;
     .      28 MioECU des contributions , pour JET , de la Suède et de la
           Suisse au budget communautaire .
 ---pagebreak---                                                                    19 .
( 9) Couvre les activités courantes au CCR dans le domaine de la
     technologie de la fusion , à savoir les études de réacteur et
     l' évaluation des risques , la sécurité de la technologie du tritium ,
     l' intégrité des matériaux structurels et les études sur la couche
     fertile .
 ---pagebreak---                                                                           20 .
ANNEXE
           EXAMEN DES REALISATIONS SCIENTIFIQUES ET TECHNIQUES
                          DURANT LA PERIODE 1984-1986
                     AU SEIN DU PROGRAMME FUSION EUROPEEN
I. INTRODUCTION
Au moment de la soumission de la précédente proposition de programme
1985-1989 , la situation scientifique était la suivante : l' évolution des
programmes de fusion dans le monde avait mis en évidence les perspectives
favorables du confinement magnétique en comparaison avec le confinement
inertiel , ainsi que le rôle prépondérant de la filière tokamak sur
laquelle les machines "Next Step" devraient être fondées . L' Europe avait
joué un rôle de premier plan dans l' amélioration de la compréhension de la
physique du confinement magnétique dans les machines toroîdales et des
progrès substantiels ont été réalisés en matière de chauffage du plasma :
-     le JET (Joint European Torus ) avait commencé à fonctionner et les
      premiers résultats ( en régime ohmique ) étaient très prometteurs ;
-     des    systèmes de chauffage mégawatts-multisecondes devenaient
      disponibles pour les machines de taille moyenne ;
-     la dégradation du temps de confinement avec l' augmentation de la
      puissance de chauffage était un sujet de préoccupation , mais la
      découverte du " régime H" à Garching permettait d' espérer que de tels
      effets nuisibles du chauffage du plasma pourraient être évités ou
      tout au moins réduits .
Sur   cette    base ,  les   objectifs  du  programme    1985-1989   étaient   les
suivants :
-     établir la base physique nécessaire au NET ( Next European Torus ) :
      l' accent était mis plus particulièrement sur le chauffage du plasma ;
-     fournir la base technologique nécessaire au NET ;
-     explorer le potentiel de certaines configurations alternatives .
Suite    à  la  décision   du   Conseil  de mars  1985 ,   il  a  fallu  ralentir
l' activité NET , et le programme de technologie a été en conséquence
remanié pour l' adapter aux nouveaux jalons du NET . L' évaluation des
réalisations scientifiques et techniques dans les sections ci-après est
faite à la lumière des objectifs fixés dans la proposition de programme
1985-1989 , mais compte tenu aussi des contraintes résultant de la dernière
décision du Conseil .
 ---pagebreak---                                                                             21 .
II . TOKAMAKS
       L' Europe consacre la majeure partie de ses efforts à cette
configuration qui , sur le plan mondial , est la plus avancée . Les
principaux problèmes auxquels la recherche Tokamak a été confrontée au
cours des dernières années ( et reste confrontée dans une large mesure )
étaient les suivants :
       les effets du chauffage additionnel sur le comportement du plasma ,
       tels que la dégradation du temps de confinement de l' énergie et du
       degré de pureté du plasma en fonction de l' accroissement de la
       puissance de chauffage ;
       le comportement du plasma au voisinage de limites opérationnelles ( de
       la densité du plasma n , du facteur de sécurité q ou du rapport entre
       la pression du plasma et la pression magnétique        ).
       Les   résultats  obtenus  sur  le JET   et dans   les  tokamaks   de  taille
moyenne améliorent la compréhension des phénomènes concernant le plasma et
donnent un aperçu des effets " de structure fine" (par exemple , cohérence
du profil ) : cela suggère de nouveaux moyens de remédier aux effets
nuisibles auxquels les tokamaks sont exposés en présence d' un chauffage
additionnel puissant .
       Les progrès réalisés en matière de construction de quatre nouveaux
tokamaks spécialisés de taille moyenne devant entrer en service en 1988
sont également présentés ; la contribution de ces machines sera très
importante pour l' élaboration du projet détaillé du NET . Un autre tokamak
spécialisé sera mis en service en 1989 .
Iia_JET
       JET est l' expérience de fusion la plus importante du monde ; il a
déjà apporté une contribution appréciable à la démonstration de la
faisabilité scientifique de la fusion ; ses objectifs initiaux ont été
atteints pour la phase des performances de base dans les délais et avec
les crédits alloués et son extension aux performances élargies est en
bonne voie .
II . 1.1 Régime en chauffage ohmique (OH). La première phase opérationnelle ,
jusqu' à fin 1984 , avait pour but la réalisation de plasmas propres se
prêtant aux études de chauffage additionnel des phases ultérieures :
       On a constaté que JET se comportait comme les tokamaks de taille plus
       modeste .
       Le contrôle stable de la position , de la taille et de la
       configuration du plasma de section en forme de D , avec élongations
       jusqu' à 1,7 , a été réalisé .
       Des décharges jusqu' à 15 s ont_^é réalises sans perturbation , tant
       que la limite de densité IL (m        = 1.10    B(T) /R(m)q   - n' était pas
       dépassée .
 ---pagebreak---                                                                           22 .
-     Des courants de plasma jusqu' à 3,7 MA ont été réalisés pendant plusieurs
      secondes ( longueurs d' impulsion de 15s ), en présence d' un champ
      magnétique de 3,45 T. Des températures des électrons et des ions pouvant
      atteindre 3 et 2,5 keV respectiv^ent^ ont été produites , avec des
      densités pouvant atteindre ^3 10          m , en un temps de confinement
      record de       = 0,8s . Chacun des paramètres - température , densité et
      temps de confinement énergétique - se situait à l' intérieur d' un facteur
      deux ou trois des valeurs nécessaires dans un réacteur de fusion .
-     Le niveau des impuretés a posé un problème , car elles réduisent le
      nombre d' ions du plasma disponibles pour la fusion et provoquent des
      pertes par rayonnement . Des expériences avec des " tuiles" à faible Z
      ( carbone) sur la paroi intérieure et une enceinte carburée ont conduit à
     une réduction des taux d' impuretés métalliques et d' oxygène .
11.1.2      Etudes   sur   le   chauffage    additionnel .  La   deuxième    phase
opérationnelle a commencé au début de 1985 après l' installation de deux
antennes HF dans le tore alimentées chacune par un générateur de 3 MW . La
puissance a été couplée au plasma à la fréquence de résonance cyclotronique
                                                 3
ionique ( ICR) des espèces minoritaires (H , He ) injectées . Les expériences de
tokamak ont repris dans JET en novembre 1985 , après une fermeture où ont été
ajoutés de nouveaux systèmes dont : le premier caisson d' injection de neutres
(NBI) , des protections de carbone dans l' enceinte à vide , une troisième
antenne d' ICRH et un injecteur d' un glaçon de deutérium . Pendant 1986 :
-     Le champ magnétique toroïdal a été porté couramment à sa valeur maximale
      projetée de 3,45 T. Le courant du plasma , la position et la forme du
      plasma sont tous les trois contrôlés par des systèmes d' asservissement .
      Des courants de plasma de 5 MA ont été obtenus couramment , avec une
      durée de " fiat top" de 4,5 s . On a obtenu un contrôle stable
      d' élongations jusqu' à 1,8 . Néanmoins , le courant est resté limité dans
      un domaine d' opération dépendant de cette élongation .
-     Les 3 antennes haute fréquence (HF) ont fonctionné régulièrement à une
      puissance combinée de 7,2 MW pour des impulsions de 2s . Des expériences
      avec une impulsion de 8s ont été faites , qui ont fourni 40 MJ au plasma .
      Un injecteur de neutres en impulsion longue ( 10s ), avec 8 sources de
      faisceau , fonctionne depuis le début 1986 . Une puissance totale de
      faisceau de 5,5 MW en hydrogène neutre (H° ) , ou de 9 MW en deutérium
      neutre (D° ) a pu être injectée dans le tore . Une énergie allant jusqu' à
      40 MJ a été fournie au plasma .
-     Des premières expériences d' injection de glaçons ont été effectuées ,
      avec un injecteur produisant un glaçon unique de 3,6 ou 4,6 mm de
      diamètre à une vitesse allant jusqu' à 1,2 km/s , sous différentes
      conditions de configurations magnétiques . Cela permet d' augmenter la
      densité limite dans JET , et de réduire la charge ionique effective Zeff
      du plasma .
 ---pagebreak---                                                                       23 .
Alors que le temps global de confinement de l' énergie a pu atteindre des
valeurs allant jusqu' à 0,9s dans les décharges ohmiques , la dégradation
du confinement a été confirmée avec la HF , la NBI et le chauffage
combiné   (     (VP         )   dans le mode L d' opération avec limiteur
            u E  ^ tôt                                            _,
matériel . De manière typique , à courant de plasma maximum , c- K tombe de
0,9 à 0,4 s avec P tôt = 10 MW dans ce mode d' opération .
Le mode à séparatrice magnétique a été démontré dans JET (à la fois dans
les points de stagnation X simple et double ). L' opération en mode H a
été obtenue avec un point X simple et a toutes les caractéristiques des
décharges en mode H obtenues dans d' autres tokamaks (profiles Tg plus
plats avec gradients accentués aux bords , seuil de puissance pour
parvenir au mode H , amélioration d' un facteur 2 environ du temps de
confinement par rapport à celui obtenu en mode L avec la même puissance
de chauffage ...). Cependant , même en mode H , il n' y en a pas moins
dégradation du temps de confinement lorsqu' on augmente la puissance de
chauffage .
L' amélioration du temps de confinement lorsqu' on augmente le courant du
plasma a été observée à la fois dans le mode avec limiteur et avec point
X. Les modifications en cours sur le système poloïdal devraient
permettre en 1987 d' atteindre 7 MA en opération de limiteur , et 4 MA en
opération avec point X simple .
En opération combinée avec la NBI , des densités électroniques de crêtes
                  20  –3
supérieures à 10 *" m    ont été obtenues , durant 0,5 s après injection de
glaçons , avec une température électronique correspondante tombant à 1
                                                            –          19 3
keV . Pour une densité linéique moyenne des électrons ne - 3 x 10 m- ,
la charge ionique effective Zeff est couramment entre 2 et 3 , mais elle
peut tomber à une valeur voisine de 1 ( pendant 0,5 s ) après injection de
glaçons . La compatibilité , qui a été observée , de l' injection de glaçons
avec l' ICRH permet d' espérer obtenir l' injection multi-glaçons en 1987 .
Des dents de scie " géantes " purent être obtenues avec l' ICRH seule , en
général pour une déposition au centre de la puissance . Des dents de scie
"monstres" ont pu durer 1,2 s (avec Te = 7 keV) , qui étaient en relation
avec des profils de q plats . Des oscillations "de serpent" (m = n = 1 )
se développent après injection de glaçon (/\n /n = 100% , ^Te /Te = 20%).
                                                   e  e
Des températures ioniques de crête supérieures à 12 keV , à faible
                              19 -3
densité de plasma (2 x 10        n ), ont été obtenues avec l' injection de
 faisceaux de neutres .
                        A^
Le produit de fusion n T . (,     varie peu  avec la puissance dans le mode L
                         D 1 E 20 -3
 ( la meilleure valeur , 1 x 10      m  . keV.s , ayant été obtenue en régime
 ---pagebreak---                                                                                 24 .
                                                                            20
         ohmique à 5 MA). Une telle valeur a pu être doublée (2 x 10 ) dans le
         mode H ( 10 MW de chauffage additionnel , opération en points X). Un
         facteur supplémentaire de 4-5 est encore nécessaire pour obtenir le
         "breakeven" , qui maintenant semble être un objectif " raisonnable".
II . 2 . AUTRE S TOKAMAKS EN SERVICE :
Les tokamaks européens de taille moyenne contribuent puissamment aux progrès
de la fusion et sont pour beaucoup au succès du JET , grâce à la mise à
l' essai de différentes configurations ( telles que l' écorceur . magnétique ,
conduisant à la possibilité de confinement favorable du plasma en "mode H") ,
à l' examen de nouvelles méthodes de chauffage ou d' excitation de courant et à
la mise au point de nouveaux diagnostics .
II . 2.1 . PETULA (Grenoble ). Les travaux des dernières années ont porté sur
différents scénarios d' excitation de courant par ondes hybrides inférieures
( LH ) :
-       Le courant du plasma était entièrement excité à faible densité de plasma
               19 -3
         (^10 m ).
                                                                                  19 _3
-       A forte densité , mais inférieure à une densité limite n^ = 8 10 m ,
         le courant a été partiellement excité (à 3,7 GHz ) ;
         La montée en courant était de 0,25 MA / s avec P_,_ RF
                                                                = 0,35 MW (à 1,3 GHz ).
L' influence du profil radial de courant de plasma sur l' activité MHD a
                                                                                    19-3
également été démontrée ( suppression des dents de scie pour n L X' 6 10 m
avec 0,25 MW à 3,7 GHz ) : un résultat prometteur pour l' application du
contrôle du profil de courant dans les grandes machines telles que JET et
TORE SUPRA . L' exploitation de PETULA s' est terminée en juin 1986 lorsque
l' équipe a déménagé à Cadarache . Le transfert de PETULA à Nieuwegein
(Association Euratom-FOM) est en cours d' examen .
II . 2 . 2 . TFR ( Fontenay ) . Le chauffage par résonance cyclotronique électronique
( ECR) , un programme conjoint des Associations néerlandaise et française , a
commencé en 1985 sur le TFR : la pleine puissance de 0,6 MW a été disponible
en septembre 1985 . Des températures électroniques atteignant 5 keV ont été
obtenues avec n e = l,5.10^m '-
                                          A P RF = 0,5’ MW,' on obtient Ÿf
                                                                         ^e
                                                                              = 1 /2    E
 ( 0H ). On a fermé TFR en juin 1986 , après 13 ans d' exploitation pleine de
succès , du fait que le personnel devait être transféré sur TORE SUPRA à
Cadarache .
II . 2 . 3 . FT ( Frascati) . Le programme expérimental concernait les études des
limites de q et de n dans les décharges ohmiques et la physique de base du
chauffage LH .
 ---pagebreak---                                                                               25 .
         Limites de q et de n ( 1984 ) : plusieurs phénomènes qui limitent le
         fonctionnement des tokamaks ont été étudiés y compris pour la limite de
         densité , la propagation en dents de scie , les précurseurs de disruption ,
         la radiation de l' hydrogène et les pertes d' échange de charge ;
         Chauffage LH ( 1984-1985 ) : le chauffage LH (f = 2,45 GHz ) a été étudié
         avec deux types différents de structures de couplage . Les meilleurs
         résultats ont été obtenus en régime électronique (P              = 0,45 MW
                                                                   2 RF
         correspondant à une densité de puissance de 6 KW/ cm à la bouche du
         grill ;       > 0,5 keV et        > 1 keV) avec aucune dégradation du temps
         de confinement de l' énergie .
        Pour P Kr = 0,2 MW , n = 4 10 ^ m        I = 0,35 MA et B = 6T , le temps de
         répétition des dents de scie s' est accru d' un facteur 3 environ , tandis
         que la propagation vers l' extérieur de l' onde thermique à partir de la
         surface q = 1 s' est ralentie , suggérant de meilleures conditions de
         transport . Le chauffage LH des plasmas de haute densité à 8 GHz ( en vue
         d' une application au FTU ) est envisagé .
III . 2 . 4 . THOR (Milan). Dans l' expérience de chauffage ECR (P Rr jusqu' à 0,2
MW , f = 28 GHz ), une partie de l' onde ordinaire injectée du côté du champ
faible est absorbée au cours du premier passage et le reste est réfléchi dans
le mode extraordinaire par un miroir . Durant l' impulsion HF , la densité
diminue ( 60% ) , la température électronique globale reste constante mais le
contenu énergétique double du fait de la formation d' une population
électronique non thermique .
II . 2 . 5 . ASDEX (Garching ) . Le bon fonctionnement d' un écorceur magnétique
combiné avec un chauffage NBI puissant a conduit au confinement favorable en
"mode H". Maintenant , avec l' application d' ondes LH et le chauffage ICR ,
trois systèmes de chauffage sont disponibles et peuvent être comparés sur la
même machine en ce qui concerne l' efficacité de chauffage et les effets
synergiques :
         la combinaison chauffage      ICR-NBI donne une    efficacité  de  chauffage
         supérieure à celle que l' on obtient avec les deux chauffages séparés au
         même niveau de puissance ;
          le régime "H", qu' on ne pouvait jusqu' alors obtenir qu' avec le NBI , a
         également été obtenu avec le chauffage combiné NBI-ICR , et même avec le
          seul chauffage ICR ;
          le NBI à un niveau réduit d' énergie des particules a montré que le dépôt
          d' énergie aux bords du plasma conduisait au même temps de confinement
          que le dépôt central ;
 ---pagebreak---                                                                       26 .
-     les ondes LH ont permis d' exciter tout le courant du plasma sans
      transformateur ohmlque    (OH)  et de   démontrer  le  rechargement  du
      transformateur OH ;
-     la stabilisation des oscillations en dents de scie a été obtenue avec
      les ondes LH dans la gamme de faible densité des plasmas chauffés par
      OH et NBI ;
-     la limitation du bêta ( limite de stabilité du plasma MHD) est
      confirmée ;
-     l' injection de glaçons d' hydrogène permet d' augmenter considérable¬
      ment les limites de densité , ce qui conduit à des temps de
      confinement de l' énergie globaux ^ h = 0,16 s (ce qui est exception-
      nellement élevé pour des machines de la dimension d' ASDEX).
11.2.6 . TORTUR (Nieuwegein) . Construite pour l' étude du chauffage
turbulent , cette expérience montre une déposition d' énergie dans un profil
de courant de peau MHD instable , qui subit une relaxation par la suite . La
machine    sera   perfectionnée   en vue   de   l' étude des   phénomènes  de
fluctuations .
11.2.7 . TEXTOR (Jülich) : Le programme porte essentiellement sur
l' interaction plasma /paroi .
-     Le module de limiteur pompé ALT-I , projet de collaboration exécuté
      avec les Etats-Unis dans le cadre de l' AIE , est entré en service
      début de 1984 et s' est révélé être un outil efficace pour influencer
      la couche limite du plasma (possibilité d' extraction de l' hélium
      démontrée ). Un limiteur pompé axisymétrique (ALT-II ) , a été préparé
      ( entreprise conjointe Japon-Etats-Unis-EURATOM) et était prêt à être
      installé vers la fin 1986 .
-     La technique de la carburation " in situ", a été appliquée fin 1984 et
      a réduit fortement les concentrations initiales d' impuretés ( facteur
      5 pour l' oxygène et 25 pour les métaux). Une durée d' impulsion de 4 s
      environ et un temps de confinement de l' énergie de 0,1 s ( régime
      ohmique) ont été obtenus . Cette technique , mise au point pour la
      première fois à Jülich , s' est révélée être si prometteuse qu' on
      l' applique pratiquement sur tous les tokamaks .
-     Un système de chauffage ICR - construit et exploité par une équipe de
      l' Association belge - est appliqué avec succès à TEXTOR au niveau de
      2,3 MW pendant plus d' une seconde . La modification du système HF
      (pour la mise en oeuvre du limiteur ALT-II ) est activement préparée ,
      en même temps que l' adaptation éventuelle du système HF à la gamme de
      4 - 4,5 MW .
 ---pagebreak---                                                                          27 .
-        En coopération avec des laboratoires ayant de l' expérience dans ce
         domaine , la conception de deux injecteurs de neutres ( sur la base du
         concept    JET) , qui  seront  installés  sur  TEXTOR , est  maintenant
         terminée .
11 . 2.8 DITE (Culham) . Avec cette machine , on a pu démontrer le bon
fonctionnement d' un écorceur en hernie et obtenir la base expérimentale
nécessaire à l' évaluation de ce concept en tant que système d' extraction
et de contrôle des impuretés . Elle a fourni la première preuve ( preuve
exclusivement européenne ) de l' excitation du courant de plasma par
injection de faisceaux de neutres et a permis la codification du régime de
fonctionnement des tokamaks ( diagramme de Hugill ) . Elle a également montré
que la limite supérieure de densité conduisant à des disruptions est
également déclenchée par refroidissement dû au rayonnement .
11 . 2.9 CLEO ( Culham). Cette machine a démontré le potentiel du chauffage
ECR , en vue d' améliorer le confinement du plasma , grâce au contrôle du
profil de température du plasma . Avec une puissance de 200 KW à la
fréquence de 60 GHz , la température des électrons a augmenté d' un facteur
8 pour atteindre plus de 2 keV . La limite de densité a augmenté de 70% .
11 . 2 . 10 . DANTE ( Ris^) . Chauffage ECR dans des plasmas de très forte
densité ( conversion en mode double ) et désintégration des glaçons ( glaçons
appropriés aux diagnostics ) ont été étudiés .
11 . 2 . 11 . TCA (Lausanne ). La production de décharges plus propres a conduit
a une plus forte puissance des ondes d' Alfven ( jusqu' à 0,57 MW au moyen du
générateur d' ondes d' Alfven récemment mis en service ). L' importance du
spectre excité pour la détermination des effets de la puissance HF a été
démontrée . Le chauffage effectif de la partie centrale a été démontré .
L' onde d' Alfven cinétique a indiqué un comportement conforme à la théorie .
II . 3 . TOKA^_MOYENS_EN_CONSTRUCTION_OU_EN_COURS DE_SOUMISSION :
II . 3.1 TORE-SUPRA ( Cadarache ). Ce dispositif supraconducteur doit apporter
une contribution à la fois sur le plan physique et sur celui de la
 technologie : il permettra en particulier d' étudier l' interaction
plasma-paroi , de même que le chauffage et l' excitation de courant dans les
 décharges à impulsion longue . Tandis que le regroupement des effectifs de
 ---pagebreak---                                                                             28 .
Fontenay et Grenoble à Cadarache était terminé fin 1986 , l' assemblage de
TORE-SUPRA est entrée dans sa phase active .
Après des essais réussis , toutes les bobines supraconductrices ont été
livrées . Les parties inférieures du circuit magnétique sont installées et
le montage des modules commence . La collaboration active avec différentes
équipes américaines porte sur l' injection de glaçons , sur les limiteurs
pompés et sur les écorceurs ergodiques : la construction a commencé . TORE
SUPRA doit être mis en service en décembre 1987 .
Des prototypes des différents            systèmes  de  chauffage ont été mis à
l' essai :
-        la source d' ions a fourni ( 10 A , 60 kV) pendant 0,2s . L' extrapolation
         aux valeurs nominales ( 40A , 100 kV , 30s ) ne pose pas de problème
         important ;
-        un klystron prototype ( 3,7 GHz , 0,5 MW , 0,03 s ) a été couplé dans
         PETULA à un module de gril multijonctions ( circulateur superflu ) ;
-        des structures de couplage pour le chauffage ICR ( deux types
         d' antennes ) sont conçues de manière à permettre l' utilisation des
         queusots horizontaux pour leur installation .
11 . 3 . 2 . FTU ( Frascati ) . Cette nouvelle machine permettra d' étudier des
performances de plasma de densité et de température élevées . La
construction a commencé en septembre 1984 et toutes les commandes
principales ont été passées . Le choix du chauffage électronique LH pour
FTU a été approuvé et des expériences préliminaires sur FT avec un module
de gril de 8 GHz commence en 1986 : le but de l' expérience est à la fois
physique        ( contrôle de la limite de densité )            et technologique
(démonstration de forte densité de puissance ). Le dispositif FTU devrait
pouvoir être mis en service au début de 1988 .
11 . 3 . 3 . ASDEX-UPGRADE ( Garching ) . Ce dispositif a pour but l' étude des
performances du plasma et de l' interaction plasma-paroi en cas
d' utilisation d' un écorceur poloïdal applicable au réacteur . La
construction est en cours et tous les composants du système tokamak sont
commandés .      Le  fonctionnement devrait   commencer au cours de     la deuxième
moitié de 1988 . Le chauffage additionnel consistant en des systèmes
d' injection de neutres d' hydrogène de 6 MW et de chauffage ICR de 6 MW est
en préparation ( début de mise en service en 1989 ).
 ---pagebreak---                                                                            29 .
11 . 3 . 4 . COMPASS ( Culham) . Le but de ce dispositif est l' étude des bêtas
élevés et de la stabilité MHD . La livraison des grands composants pour ce
dispositif , approuvé en mars 1984 , se déroule de façon satisfaisante .
L' alimentation électrique du champ toroïdal a été livrée et testée de
façon satisfaisante . L' installation des trois gyrotrons du stade 1 ( ECRH
de 0,6 MW) progresse en vue du programme expérimental sur DITE qui précède
la mise en service de COMPASS ( laquelle doit commencer en 1988 ).
11 . 3 . 5 . TCV (Lausanne ). Ce projet de tokamak , approuvé en 1986 , a pour but
la production de plasma avec de grandes élongations (jusqu' à 3) qui
devraient offrir la possibilité d' atteindre des courants de plasma plus
intenses et , partant , des valeurs bêta plus élevées . La mise en service
des dispositifs devrait avoir lieu fin 1989 .
III . CONFIGURATIONS ALTERNATIVES
         Comme nous l' avons déjà dit , l' un des trois principaux objectifs du
programme fusion est d' explorer le potentiel en vue du réacteur de
certaines configurations alternatives : principalement les stellarateurs
et les pinches à champ inversé . Les résultats expérimentaux des
dispositifs en service , ainsi que l' état d' avancement de ceux qui sont en
construction ou à l' état de projet , sont présentés dans ce qui suit .
111 . 1 . STELLARATORS
111 . 1.1 . WENDELSTEIN VII A ( Garching ) . Ce dispositif a été démantelé
récemment après dix années de bon fonctionnement . Le chauffage ECR ( 28 GHz
et , par la suite , 70 GHz , 0,2 MW) a donné ( en coopération avec
l' université de Stuttgart ) les résultats suivants :
         production et chauffage de plasma (T      jusqu' à 2,5 keV) ;
-        confinement néo-classique pour électrons au centre du plasma ;
         génération de champs électriques radiaux , lorsque l' ECR est combiné
         avec la NBI ;
         fonctionnement en mode torsatron , qui a prouvé que les zones de
         confinement stables pouvaient être accrues par un cisaillement
         positif .
 111 . 1.2 . WENDELSTEIN VII-AS (Garching ). La construction par l' industrie
 des principaux composants a été récemment terminée et l' assemblage des
 modules est en bonne voie . La bobine prototype a été testée avec succès et
 ---pagebreak---                                                                             30 .
toutes les bobines sont terminés .            En l' état actuel ,  W VII-AS devrait
pouvoir être mis en service l' été 1987 . Le chauffage ECR de 0,8 MW ( longue
impulsion ) sera disponible dès le départ et les chauffages NBI ( 1,2 MW) et
ICR (3 MW) seront opérationnels quelques mois plus tard .
III . 1.3 . WENDELSTEIN VII-X (à l' étude à Garching) . La construction du
dispositif faisant suite à W VII-AS est envisagée . Il devrait permettre de
conclure si le concept de stellarator avancé est faisable pour les
réacteurs de fusion ( les calculs numériques donnent des valeurs moyennes
de bêta de 5%). En outre , une étude des propriétés au niveau reacteur qui
distinguent le stellarateur du tokamak est en cours (en collaboration avec
Karlsruhe ) .
III . 1.1 .    TJ-II    (Madrid ,  en   cours   de   soumission ).  En  vue  de  la
participation de l' Espagne au programme fusion européen (depuis le 1er
janvier 1986 ), JEN-MADRID s' est concentré sur la construction d' une
expérience de confinement , un Héliac flexible (TJ-II ), complémentaire aux
autres stellarateurs en Europe . Ce projet est actuellement à l' étude au
sein des instances EURATOM .
III . 2 . PINCHES A CHAMP INVERSE
111 . 2.1 . ETA-BETA II (Padoue ) . Les expériences faites avec ce dispositif
servent d' études de soutien pour le prochain projet RFX . Des études de
fluctuations ont été faites sur le confinement du plasma et les phénomènes
de relaxation conduisant à l' inversion du champ toroïdal . Un plasma propre
(Ze^=.V10l) ^ des haute   densité ( 10^m ^) avec ^^10% , T = 0,1 keV et
                   a été réalisé .
111 . 2 . 2 . HBT-X ( Culham) . Les expériences faites avec ce dispositif
montrent que le contrôle de la position d' équilibre du plasma et la
réduction des erreurs de champ donnent des temps de confinement plus
longs . La température des électrons et le temps de confinement augmentent
avec le courant : dans certains cas , la température augmente
proportionnellement au courant à une valeur constante de bêta (/>/ 10 % ) .
111 . 2 . 3 . RFX ( Padoue ). Il s' agira du plus grand Pinch à champ inversé du
monde (R = 2m , a = 0,5 m , courant de plasma jusqu' à 2 MA). Il permettra
 ---pagebreak---                                                                       31 .
d' étudier le confinement et le chauffage du plasma dans des conditions
plus proches du régime thermonucléaire que dans les RFP actuels . Après la
phase de projet détaillé , la construction des bâtiments et des principales
infrastructures a commencé et des appels d' offres ont été lancés pour les
principaux composants de la machine . Culham apporte une contribution
importante à cet effort . La machine devrait être mise en service en 1989 .
11 1.3 . AUTRES MACHINES
Outre ces deux principales configurations alternatives         en Europe , il
existe d' autres dispositifs dont le but principal est l' élargissement de
la base de données sur la physique fondamentale du plasma :
111 . 3.1 . SPICA (Nieuwegein) . Dans ce pinch hélicoïdal , le plasma est
stabilisé à des valeurs élevées de bêta par les courants à force nulle qui
entourent le plasma et par la coque conductrice . Les expériences dans
SPICA I ont démontré la possibilité de créer de tels plasmas à valeurs
bêta élevée ; SPICA II , dont la construction s' est achevée en 1984 , donne
des résultats préliminaires prometteurs (valeurs élevées de bêta avec
grandes élongations ) .
111 . 3 . 2 . EXTRAP ( Stockholm). EXTRAP est une expérience qui fait suite à
celles ( linéaires ou secteur toroïdal) ayant apporté la preuve d' un état
macroscopiquement stable . Ce pinch en Z est stabilisé par un champ
magnétique superposé octupolaire engendré par des conducteurs extérieurs .
Les expériences ont commencé récemment .
111 . 4 . CONFINEMENT INERTIEL
Le programme fusion européen consacre 1 % de ses efforts au maintien de
contacts avec la recherche faite ailleurs et à l' évaluation des progrès
accomplis dans ce domaine . Les deux laboratoires qui y participent sont :
-       Garching , qui met au point un laser à gaz à impulsions courts pulsé
        de haute puissance (2 KJ ) ;
-       Frascati , qui met au point un laser en verre à double faisceau
         (2 X 70 J ).
 ---pagebreak---                                                                         32 .
IV . RECHERCHE DE SOUTIEN ET TRAVAUX DE DEVELOPPEMENT
Outre la planification , la construction et la mise en service des machines
mentionnées dans les sections précédentes , une importante activité au JET
et dans les laboratoires associés est consacrée à :
-       des études de soutien et des développements pour JET , ainsi que pour
        NET ;
-       la mise au point de sous-systèmes nécessaires pour étendre notre
        connaissance des phénomènes qui caractérisent les plasmas et pour
        améliorer les performances de ces derniers .
IV . 1 . SUPPORT A JET ( Contrats Article 14 et accords de tâches )
-       Les deux principaux contrats concernant le NBI ( conclus avec Fontenay
        et Culham) ont été exécutés avec succès et la première application du
        chauffage par faisceaux de neutres à JET a permis de doubler la
        température ionique centrale et de la porter à 6,5 keV .
-       Au cours de cette période 1984-86 , un grand nombre de diagnostics ont
        été mis au point par les Associations , installés sur JET et exploités
        par le personnel détaché par les Associations :
        .     Diffusion Thomson ponctuelle ( Ris^)
        .     Interf éromètre FIR et balayage spatial VUV ( Fontenay )
        .     Analyseur de particules neutres et spectromètre à rayons X
              ( Frascati )
        .     Caméras à rayons X mous ( Garching )
        .     Système rapide de détection de l' émission cyclotronique des
              électrons (Nieuwegein)
        .     Diagnostics neutroniques (Harwell ) et diagnostics spectrosco¬
              piques ( Culham)
        .     Spectromètres neutroniques à temps de vol de 2,4 MeV ( Studsvik )
        .     Sonde à la limite du plasma ( JET , Culham et Garching )
        .     Réseau de bolomètres ( Garching )
-       Des contrats ont été conclus pour la mise au point d' un prototype
        pour la production de glaçons ( Grenoble ) , leur accélération par un
 ---pagebreak---                                                                              33 .
         canon à gaz chauffé par un arc (Ris^) et pour la conception
         d' injecteurs de glaçons pour JET (Garching ) .
-        L' équipe du JET a également chargé les Associations de faire
         différentes études analytiques et numériques sur l' équilibre et le
         transport du plasma , sur le dépôt d' énergie par différents systèmes
         de chauffage et sur l' interaction plasma /paroi .
-        De    nombreux    laboratoires     associés  participent    directement   à
         l' exploitation du JET en détachant du personnel selon le système du
         personnel associé . Le laboratoire de Culham en particulier , adjacent
         au JET , détache une fraction importante de ses chercheurs .
ÎÏ^2JL_AUTRES_DEVEL0PPEMENTS_DANS_LES_LAB0RAT0IRES_ASS0CIES
IV . 2.1 . NBI . Les travaux faits pour l' installation de systèmes NBI sur
certains tokamaks en construction et sur TEXTOR se poursuivent .
IV . 2 . 2 . Gyrotrons . Des études et des travaux concernant les gyrotrons se
poursuivent dans quelques laboratoires et dans l' industrie :
         Un contrat industriel a été conclu par la Commission pour le
         développement d' un gyrotron de 100 GHz , 0,2 MW et 0,1 s . Les tubes
         prototypes sont à l' essai .
-        Un gyroklystron quasi-optique expérimental de 120 GHz est en cours de
         réalisation au sein de l' Association suisse , avec la collaboration de
          l' industrie . Tous les composants sont construits et le système se
          trouve actuellement au stade du montage .
         Etudes physiques sur les gyrotrons à très haute fréquence
          (Karlsruhe ) : tous les composants sont construits et les expériences
          ont commencé .
          Un contrat industriel a été conclu par Garching pour un gyrotron de
          70 GHz . Les tests préliminaires ont été couronnés de succès .
 IV . 2 . 3 . Glaçons . A Risy5 , des glaçons de deutérium ( 3,2 mm de diamètre) ont
atteint des vitesses de presque 2 km/s dans un canon à gaz chauffé par un
 ---pagebreak---                                                                           34 .
arc . Un injecteur multiple - pour glaçons de dimension variable - fondé
sur le principe de la centrifugation a été mis au point à Garching .
IV . 2 . 4 . Diagnostics . Outre les différents diagnostics pour JET , de
nombreux diagnostics (dont quelques-uns sont novateurs ) ont été développés
et mis en place dans les dispositifs par les Associations :
-        Réf lectrométrie ( Fontenay) pour les mesures de densité des électrons
-        Polari-interféromètre à laser HCN (Jülich ) pour mesurer la
         distribution locale du courant
-        Diagnostics nouveaux pour la zone de bord du plasma , telle que
         fluorencence par résonance induite par laser et faisceaux de lithium
         ( Jülich ) .
IV . 2 . 5 . Faisceaux d' ions . Les travaux concernent :
-        les faisceaux H et l' accélération des ions (Amsterdam) : 4 petits
         faisceaux ont été produits ( courant de 3 mA , avec énergie des
         particules de 120 keV ) ;
-        les faisceaux d' ions négatifs (Culham) : on a obtenu 30 mA/cm2 avec
         de bonnes perspectives d' extrapolation à de larges surfaces ;
-        les faisceaux d' ions négatifs ( Stockholm en coopération avec
         Grenoble ) : les expériences ont donné des courants de 150 mA formés
         d' ions H    accélérés à 55 kV .
IV . 2 . 6 . Travaux pour NET
L' état d' avancement dans la conception du NET a permis à l' équipe NET de
définir les tâches détaillées à exécuter en technologie dans les
Institutions Associées . Jusqu' à présent , environ cent tâches ont été
attribuées en ce qui concerne les aimants , la couche fertile , les
matériaux , le tritium , la télémanipulation et la sécurité . Les résultats
de ces travaux ont déjà été ré-injectés dans la conception établissant
ainsi       une interaction    étroite  et fructueuse  entre les laboratoires  et
l' équipe NET . En outre , NET a attribué environ 90 contrats d' étude aux
Associations à la fois dans le domaine de la physique et de l' ingénierie .
Les Associations détachent également du personnel à l' équipe NET , dans le
cadre de l' Accord NET (" NET Agreement ") .
 ---pagebreak---                                                                                35 .
IV . 3 . ETUDES_THEORigUES
Des études analytiques et numériques et des codes de calcul ont été
réalisés dans la plupart des laboratoires :
-       L' équilibre MHD et le transport sont étudiés dans la plupart des
        laboratoires . En particulier , c' est l' activité principale de l' équipe
        de recherche de l' université libre de Bruxelles ;
-       Les instabilités macroscopiques et microscopiques et notamment les
        limites de bêta sont étudiées essentiellement dans les laboratoires
        ayant    les    ordinateurs   indispensables      aux   calculs     numériques
        importants ;
-       Des   codes    de  calcul   sont  mis   au    point   dans    les   principaux
        laboratoires ( et à Lausanne ) pour l' équilibre , le transport , etc
        ( code en 3-D à Garching pour l' étude du concept de stellarator
        avancé ) ;
-       Des études sur le chauffage (propagation des ondes et dépôt
        d' énergie , traçage des rayons ...) et sur l' excitation de courant
        sont   effectuées   essentiellement   dans   les  laboratoires    faisant   des
        expériences de ce genre .
V. TECHNOLOGIE
La bonne mise en oeuvre du programme de fusion a été l' une des principales
réalisations des dernières années . La majeure partie des travaux est axée
sur le NET , mais une partie concerne aussi les applications à long terne
(matériaux à        faible  activation , études    de   sécurité   et   influence   sur
l' environnement ) .
Les domaines couverts sont les aimants , la technologie du tritium , la
couche fertile , les matériaux , la sécurité et l' environnement ; les
travaux sont effectués au sein des Associations ( dans de nombreux cas , par
le détachement de groupes des laboratoires de fission), au CCR et , dans
une mesure moindre , dans l' industrie .
V.l . AIMANTS SUPRACONDUCTEURS
Le programme de développement a été axé sur les principaux besoins du
NET : bobines supraconductrices à champs poloïdal et toroïdal . Le
principal projet entrepris concernait la conception et la réalisation ,
avec le concours de l' indsutrie , de la bobine EURATOM pour le LCTF à Oak
 ---pagebreak---                                                                        36 .
Ridge ( ORNL) , aux Etats-Unis . Cette bobine à champ toroïdal , de 38 tonnes ,
au NbTi , refroidie à l' hélium supercritique , a été testée dans
l' installation de Karlsruhe avant d' être envoyée au LCTF en même temps que
5 autres provenant du Japon ( 1 ), de la Suisse ( 1 ) et des Etats-Unis ( 3 ),
qui doivent toutes être testée conformément à un accord avec l' AlE . Le
programme à l' ORNL a commencé en avril 1986 .
Pour son champ toroïdal , il se peut que le NET ait besoin de
supraconducteurs d' une capacité atteignant ou dépassant 12 Tesla , ce qui
exige le développement de matériaux avancés , tels que NbSn3 et NbAl^, et ,
à cette fin , un consortium de 3 laboratoires associés a construit le
dispositif d' essai d' un champ intense qui fonctionne actuellement à 8 T
( performance à 12 T en 1987 ) .
Le tokamak TORE-SUPRA a fourni une expérience précieuse en matière
d' appréciation du concept global de tokamak expérimental supraconducteur
et permettra l' essai " in situ", dans quelques années , d' une bobine modèle
à champ poloïdal adaptée au NET . La mise au point de cette bobine est en
cours .
Y - 2 - _TEÇHNOLOGIE_DU_TRITIUM
Les travaux sont axés sur la mise au point des composants du système de
tritium du NET et sur les aspects de sécurité de la manipulation du
tritium .
Un important thème d' étude est la purification du plasma extrait du NET .
Le DT qui sera empoisonné par l' hélium et par différentes impuretés
métalliques et gazeuses doit retrouver toute sa pureté . La méthode par
excellence est son passage à travers des membranes au Pd-Ag , qui est
actuellement étudié dans la boucle PALLAS . Les getters sont étudiés comme
solution de rechange et le Ti-Zr est jugé particulièrement efficace . Les
impuretés gazeuses séparées contiennent encore du tritium et doivent donc
subir d' autres procédés de détritiation . Ces procédés font actuellement
l' objet d' études expérimentales ( lit-U , autres lits de métaux chauds ). De
même , on étudie les techniques de décontamination de l' air et des déchets
solides tritiés . Pour manipuler l' eau fortement tritiée , on met au point
deux prototypes d' électrolyseur . Enfin , on a obtenu des spécifications
détaillées       avec   le    concours de    l' industrie pour    des   pompes
turbomoléculaires de grande capacité ( compatibles avec le tritium) et pour
de grandes vannes à fermeture rapide entièrement métalliques ( étude de
faisabilité en cours dans l' industrie ). Une grande partie des expériences
 ---pagebreak---                                                                       37 .
ci-dessus implique l' utilisation de tritium et exige donc des dispositifs
particuliers . Ces dispositifs sont maintenant mis à la disposition du
programme de fusion en France ( Bruyère-le-Châtel, Saclay) et d' autres sont
en construction (KfK et CCR , Ispra) , afin d' augmenter les capacités
expérimentales nécessaires au programme .
V.3 . COUC HE FERTILE
Les études techniques concernant la couche fertile surrégénératrice de
tritium ont révélé deux options : l' une utilisant un surrégénérateur
eutectique liquide autoréfrigérant au lithium-plomb , l' autre utilisant des
composants céramiques solides du lithium avec de l' hélium comme
réfrigérant . Les travaux expérimentaux ont donc été axés sur
l' établissement de la base de données nécessaires en ce qui concerne ces
matériaux .
Pour ce qui est du surrégénérateur eutectique au lithium-plomb , les
données sur la solubilité et la diffusion de l' hydrogène ont été
complétées par de nouvelles mesures . Les essais de compatibilité et de
fragilisation par le métal liquide n' ont pas révélé de fissures ou de
fractures imminentes du matériau du container . Une première expérience a
été acquise en matière de récupération du tritium du métal liquide au
moyen de getters au Ti ou de gaz inertes en ébullition .
En ce qui concerne les composés céramiques du lithium , six laboratoires
européens (partiellement intégrés au sein d' un accord AIE ) collaborent à
l' exécution d' un projet d' importance majeure . Des méthodes de fabrication
d' aluminates et d' ortho- et métasilicates de lithium très purs ont été
établies . Des premières expériences d' irradiation de courte durée du type
capsules aérées produisant des quantités infimes de tritium ( 300-350
Ci / spécimen) ont permis de sélectionner les spécimens au "meilleur
comportement ". Ils vont être soumis à présent à des irradiations de plus
longue durée dans des dispositifs à la fois thermiques et de fission
rapide , l' objectif final étant la preuve de la possibilité de générer du
tritium .
У^._^ТЕК1А11Х
A la suite des études conceptuelles du NET , la portée de ce domaine a
maintenant été élargie de manière à recouvrir aussi les matériaux
 structurels , la protection de la première enceinte , les matériaux isolants
 et optiques et ceux des écorceurs .
 ---pagebreak---                                                                        38 .
Le matériau structurel pour le NET sera de l' acier soit austénitique
( 316 ), soit martensitique ( 1.4914) ; pour les applications à long terme ,
on pourrait choisir des aciers austénitiques au Mn-Cr , des alliages au
vanadium et des aciers de faible activation judicieusement dosés .
Quelques résultats préliminaires importants ont été obtenus sur la tenue
sous irradiation des aciers austénitiques 316 et ce , dans le cadre d' un
exercice international qui a commencé en 1981 et qui porte sur trois
réacteurs de fission en Europe (HFR/Petten , BR-2 /Mol , R2 / Studsvik) et sur
deux aux Etats-Unis ( HFIR et ORR , tous deux à Oak Ridge ) . Des échantillons
d' acier de référence d' Europe , du Japon et des Etats-Unis sont employés .
La plupart des essais de résistance à la traction et à la fatigue après
irradiation sont terminés (doses d' irradiation : 5 dpa et 10 dpa ) . Les
expériences de fluage en pile sont encore dans les réacteurs , accumulant
les doses , et la première expérience de résistance à la fatigue en pile
(BR-2 ) était prête à être effectuée dans le réacteur à la fin de 1986 .
La plupart des alliages structurels précités ont aussi fait l' objet
d' essais mécaniques pendant ou après irradiation par un faisceau de
particules d' accélérateur simulant les dégâts par irradiation de fusion ;
par exemple , mesures de fluage sous torsion sur acier austénitique 316 L ;
étude sur la fatigue oligocyclique et l' interaction f luage-f atigue ; étude
de fluage sous irradiation montrant un fluage identique pour la traction
et la compression , étude de rupture des aciers 316 mettant en évidence une
forte diminution de la résistance aux alentours de 1 000 ppm de
concentration d' hélium , et autres études .
En ce qui concerne les matériaux de protection de la première enceinte ,
après examen d' un grand nombre de matériaux proposés , ceux qui ont été
retenus en fin de compte sont des graphites à grain fin , une certaine
catégorie de SiC et des composés de graphite /SiC .
De même , la recherche sur les isolateurs électriques céramiques appropriés
indique que les plus prometteurs sont ceux à l' alumine , au spinel et à
l' oxyde de magnésium . En outre , des méthodes ont été mises au point pour
mesurer la tangente de perte diélectrique , pendant et après irradiation ,
des matériaux optiques à utiliser dans différentes gammes de fréquence du
chauffage HF du plasma .
V_L5:L_SECURXTE_ET_ENVXR0NNEMENT
Les     travaux  sont  axés  essentiellement   sur  les  causes   et  sur   les
conséquences éventuelles du rejet de tritium gazeux et l' élimination des
déchets tritiés ( solides ).
 ---pagebreak---                                                                      39 .
Des modèles de calcul des termes source radioactifs et de la dispersion
globale du gaz de tritium et du HTO ont été mis au point ( premier essai de
validation en cours ) .
Des modes de défaillance ont été analysés et une évaluation des risques a
été faite pour différents composants du NET . La décontamination des
déchets métalliques tritiés a été étudiée et on a constaté que la fusion
sous vide et le dégazage étaient des plus efficaces .
Une évaluation de l' impact écologique de la fusion a été préparée et sera
communiquée au Parlement et au Conseil . Ce document examine également les
perspectives économiques de la fusion .
VI . NET
L' équipe NET a commencé ses travaux sur la définition du NET en 1983 dans
le but de définir les objectifs , les principaux aspects conceptuels , les
options et la planification du NET et d' identifier la R et D , notamment
dans le domaine de la technologie , nécessaire à la conception du NET .
Cette phase s' est achevée fin 1985 et était suffisamment avancée pour que
l' on puisse passer à la phase pré-conceptuelle ; le programme de R et D
technologique a été lancé dans la plupart des domaines intéressant le NET .
Les objectifs du NET sont la production d' un plasma présentant des
paramètres et des performances intéressant le réacteur et l' étude des
principaux problèmes de faisabilité technique d' un réacteur de fusion .
Ainsi , le NET devrait avoir pour but le contrôle de l' ignition et de la
combustion de longue durée , la démonstration et la fiabilité et de la
possibilité d' entretien du système ainsi que de sa sécurité de
fonctionnement et de son faible impact sur l' environnement . Enfin , le NET
pourra servir à sélectionner les concepts et à essayer des matériaux et
les systèmes d' extraction du tritium et de l' énergie pour le DEMO
 ( réacteur de démonstration). A cette fin , un scénario d' exploitation
flexible à plusieurs étapes ( 13 ans ) a été mis au point . La conception et
les paramètres de la machine ont été choisis après des études
d' optimisation approfondies .
Les lois d' échelle pour les performances du plasma qui sont à la base du
choix des paramètres sont conformes aux résultats expérimentaux actuels
 ---pagebreak---                                                                      40 .
obtenus sur les Tokamaks ; toutefois , compte tenu d' une éventuelle
dégradation de ces performances , on a adopté des marges assez importantes
pour obtenir l' ignition et une combustion de longue durée . Les dimensions
totales seront nettement supérieures à celles du JET ; le courant du
plasma pourra atteindre jusqu' à 15 MA et le rayon principal sera de 5
mètres en comparaison de 3 mètres dans le JET , ce qui reflète aussi le
fait qu' une couche fertile et un écran sont prévus entre la chambre du
plasma et les bobines supraconductrices à champ magnétique toroïdal .
Pendant une impulsion de combustion D-T ( d' une durée de 500 secondes
environ) , une puissance pouvant atteindre 600 MW sera générée par les
réactions de fusion .
Des projets de conception pour les principaux composants de la machine de
base ont été élaborés , afin de donner aux Associations des indications
concernant le développement de ces composants et de confier à l' industrie
des études de faisabilité . Pour les composants en contact avec le plasma ,
dont les conditions de service sont extrêmement difficiles et quelque peu
incertaines , plusieurs conceptions ont été retenues , et la sélection de
solutions de référence exige un complément de travaux et de base de
données . Des tâches ont été définies dans ce sens pour les Associations et
l' industrie .
VII . CONCLUSION
Au moyen du JET , l' expérience la plus importante au monde qui dès le début
a été projetée comme effort en commun de toutes les Associations , des
tokamaks de dimension moyenne et des machines des configurations
alternatives dans les laboratoires associés , l' Europe a atteint ces
dernières années une position de premier rang incontestée dans le monde .
Le programme " fusion" européen participe à tous les modes de collaboration
actuellement en discussion entre les programmes fusion mondiaux . Il est
bien pourvu pour maintenir une telle position de premier plan dans les
années à venir à condition qu' un soutien financier suffisant lui soit
accordé .
 ---pagebreak---                                                                             41
                      B)  PROPOSTA DE REGULAMENTO DO CONSELHO
          que adopta um programa de investigação e formação ( 1987 a 1991 )
                    no domínio da fusão termonuclear controlada
0 CONSELHO DAS COMUNIDADES EUROPEIAS ,
Tendo em conta o Tratado que institui a Comunidade Europeia da Energia Atómica
e , nomeadamente , o seu artigo 72 ,
Tendo em conta a proposta da Comissão 1 , apresentada após consulta do Comité
Científico e Técnico ,
                                                  2
Tendo em conta o parecer do Parlamento Europeu      ,
Tendo em conta o parecer do Comité Económico e Social
Considerando que o problema da energia é comum a todos os Estados-membros ;
que um esforço conjunto para a resolução deste problema será a via mais
adequada para a obtenção de melhores resultados ; que a fusão termonuclear é
uma solução possível para o problema da energia a longo prazo ; que deve ser
coordenada a utilização racional de todas as diferentes fontes de energia ;
que a Comunidade deve , portanto , continuar a assegurar uma consistência de
esforços óptima entre as actividades comunitárias nos diversos sectores da
energia e a investigação energética ;
Considerando que o Conselho adoptou em . ^ o programa-quadro                   de
actividades comunitárias no domínio da investigação e do desenvolvimento
tecnológico ( 1987 a 1991 ), que tem em conta as considerações anteriores ;
    JO ns
    JO ns
    JO ns
    JO ns
 ---pagebreak---                                                                                 42
Considerando que a fusão termonuclear é uma nova fonte potencial de energia
que utiliza um combustível virtualmente inesgotável e universalmente
acessível ; que os reactores de fusão magnética terão incorporados dispositivos
de segurança e parecem garantir um impacto relativamente baixo sobre o
ambiente , tornando -se , assim , a fusão termonuclear um objectivo importante
dentro do programa-quadro ;
                                                         5
Considerando que , na sua Decisão 85/201 /Euratom , o Conselho adoptou um
programa de investigação e formação ( 1985 a 1989 ) no domínio da fusão
termonuclear controlada ; que o artigo 3e da referida decisão prevê que a
Comissão , baseada numa revisão a ser executada durante o segundo ano do
programa ,    deverá apresentar ao Conselho uma proposta de revisão com o
objectivo de substituir em 1987 o programa de 1985 a 1989 por um novo programa
de cinco anos , sendo 1987 , 1988 e 1989 anos comuns a ambos os programas ; que
a Decisão 85 / 201 /Euratom deverá , portanto , ser substituída ;
Considerando que ,     como resultado da substituição da Decisão 85/ 201 /Euratom ,
aproximadamente      171 milhões de ECUs da quantia estimada necessária para o
programa anterior ,     excluindo o JET ( Joint European Torus ),   e aproximadamente
209 milhões de ECUs da quantia estimada necessária para o programa anterior
para o projecto JET não terão sido utilizados ;       que estas quantias podem ser
atribuídas ao novo programa ;      que essa atribuição ,   e ainda o facto de que o
novo programa abrange todo o trabalho executado nos Estados-membros neste
domínio ,   deverão ser factores de peso na determinação das quantias estimadas
necessárias à execução do novo programa ;
Considerando que ,     perante a amplitude do esforço necessário para atingir a
fase de aplicações da fusão termonuclear controlada ,             que poderá trazer
benefícios à Comunidade ,     o trabalho empreendido até aqui neste domínio deverá
continuar numa base conjunta nas várias fases do seu desenvolvimento ;
Considerando que a investigação proposta pela Comissão constitui um meio
adequado de prosseguir tal acção e é , consequentemente , do interesse comum
adoptar um programa plurianual no domínio da fusão termonuclear controlada ,
cuja existência é , além disso , necessária para que a Comunidade possa
participar da cooperação internacional neste domínio ;
5
  JO ne L 83 de 28.3.1985 , p. 25 .
 ---pagebreak---                                                                            43
Considerando que a estratégia sobre a qual se baseia a prossecução do programa
deverá permanecer inalterada , designadamente :
    continuação de um programa de grande envergadura orientado para um reactor
    de demonstração e baseado presentemente no conceito Tokamak ; conclusão da
    primeira fase do programa constituído pelo projecto JET com as suas
    extensões e pela exploração total dos dispositivos existentes ou em vias
    de construção nas Associações ,
    continuação do ante-projecto da segunda fase do programa Tokamak , o Next
    European Torus ( NET ),   e prossecução dos desenvolvimentos tecnológicos
    necessários à sua concepção e construção ,          e dos que se revelem
    necessários , a longo prazo , para o reactor de fusão ,
    investigação ,   dependendo   dos   recursos disponíveis ,  de  sistemas  de
    conf inamento alternativos ,  com especial incidência nas estrições de campo
    invertido e nos stellarators ,    sujeita a uma reavaliação periódica do seu
    potencial de reactor em comparação com o do Tokamak ;
Considerando que esta estratégia deverá ser reavaliada na próxima revisão do
programa ,  com o objectivo de substituir o presente programa ,     no dia 1 de
Janeiro de 1990 , por um novo programa de cinco anos ; por altura dessa revisão ,
seria conveniente decidir -se quando se deverá passar à operação D-T no JET e
iniciar o projecto detalhado do NET ;
Considerando que o programa de investigação do Centro Comum de Investigação
prevê a participação do CCI no domínio do NET e da Tecnologia ;
Considerando que a Suécia e a Suíça estão associadas             às  actividades
comunitárias no domínio da fusão termonuclear controlada ;
Considerando que a Comunidade deverá continuar a encorajar a construção de
determinados equipamentos relacionados com projectos aos quais foi dado um
estatuto prioritário , o apoio ao JET e ao NET por parte das Associações , e
certos desenvolvimentos no domínio da tecnologia de fusão , concedendo uma taxa
preferencial de participação nas despesas com tais projectos ;
 ---pagebreak---                                                                               44
Considerando que se deve consolidar a participação directa da indústria na
execução do programa ,    em especial no que respeita ao NET e à tecnologia da
fusão ;
Considerando ,   além disso , que se deve promover a mobilidade do pessoal entre
as organizações que cooperam na execução do programa ;
ADOPTOU 0 PRESENTE REEGULAMENTO :
                                     Artigo 1 2
É adoptado ,   por um período de cinco anos com início em 1 de Janeiro de 1987 ,
um programa da Comunidade Europeia da Energia Atómica de            investigação e
formação no domínio da fusão termonuclear controlada ,     nos termos definidos no
Anexo .
                                     Artigo 2fi
Os fundos estimados necessários para a execução do programa ,      com exclusão do
JET ,   elevam-se a 533 milhões de ECUs ,    incluindo as despesas relativas a uma
força de trabalho de 105 pessoas .        Os fundos estimados necessários ao JET
durante o período do programa elevam -se a 378 milhões de ECUs ,       incluindo as
despesas relativas a uma força de trabalho de 191 agentes temporários , na
acepção da alínea a ) do artigo 22 das condições de recrutamento de outros
agentes das Comunidades Europeias .
 ---pagebreak---                                                                            45
                                    Artigo 3fi
No decorrer do terceiro ano ,  a Comissão passará à avaliação do programa tendo
em conta os seus objectivos estabelecidos no Anexo . Na sequência desta
avaliação , a Comissão deverá apresentar ao Conselho , em 1989 , uma proposta de
revisão destinada a substituir o programa actual por um programa de cinco
anos , o qual entrará em vigor em 1 de Janeiro de 1990 .
                                    Artigo 43
A Decisão 85/ 201 /Euratom é revogada pela presente decisão ,     com efeitos a
partir de 1 de Janeiro de 1987 .
                                    Artiso 5e
0 presente regulamento entra em vigor em 1 de Janeiro de 1987 .
0 presente regulamento é obrigatório em todos os seus elementos e dirctamente
aplicável em todos os Estados-membros .
Feito em Bruxelas , em
                                                     Pelo Conselho
                                                       0 Presidente
 ---pagebreak---                                                                                46
                                   ANEXO
                       FUSÃO TERMONUCLEAR CONTROLADA
O programa a ser executado abrange :
a)  física dos plasmas no sector em questão ,           particularmente estudos de
    carácter básico relativos ao confinamento com dispositivos adequados e
    aos métodos para a produção e aquecimento do plasma ;
b)  investigação do confinamento ,        em configurações fechadas ,     de plasmas
    de hidrogénio ,   deutério e trítio de densidade e temperatura altamente
    variáveis ;
c)  investigação das interacções          e fenómenos de transporte de matérias
    leves e do desenvolvimento de lasers de alta potência ;
d)  desenvolvimento e aplicação ,         aos dispositivos de confinamento ,      de
    métodos suficientemente potentes para aquecimento do plasma ;
e)  aperfeiçoamento dos métodos de diagnóstico ;
f)  ante-projecto e ,     possivelmente ,      início do projecto de engenharia
    pormenorizado    do   NET  ( Next    European    Torus ),   e   desenvolvimentos
    tecnológicos necessários para        a sua concepção e construção ,     bem como
    os necessários a longo prazo para o reactor de fusão ;
g)  extensão    do  dispositivo     JET    até  obtenção    do  rendimento   máximo ;
    funcionamento e exploração do JET .
0 trabalho referido em a ),    b ),   c ), d ), e ) e f ) será executado através de
associações ou de contratos de duração limitada ,          concebidos para produzir
os  resultados necessários       à   execução    do  programa ,    e  que  tomam  em
consideração o trabalho executado pelo Centro Comum de Investigação ,
nomeadamente em relação ao NET e à tecnologia referida em f ).
 ---pagebreak---                                                                                47
     A execução do projecto JET referido em g ) foi confiada à " Joint European
     Torus ( JET ), Joint Undertaking", criada pela Decisão 78/471 /Euratom 1 .
2 . 0 programa estabelecido no ponto 1 faz parte dum projecto cooperativo a
     longo prazo que abrange todas as actividades empreendidas nos
     Estados-membros no domínio da fusão termonuclear magnética controlada . Foi
     concebido com o objectivo de levar , no momento oportuno , à construção
     conjunta de protótipos para produção industrial e comercialização .
3.   A quantia de 533 milhões de ECUs , estimada necessária para a execução do
     programa com exclusão do JET , destina-se a financiar :
     a)  projectos prioritários a uma taxa uniforme de aproximadamente 45 % ,
         como especificado no ponto 4 ;
     b)  despesas     correntes   das  associações ,   a  uma  taxa     uniforme  de
         aproximadamente 25 % ;
     c)  certos contratos       industriais nos domínios de    "NET/tecnologia da
         fusão " e o desenvolvimento de métodos avançados de aquecimento do
         plasma a uma taxa de 100 % , tal como    definido no ponto 4 ;
     d)  custos administrativos e despesas destinadas a assegurar a mobilidade
         do pessoal ,    de modo a que este possa trabalhar em organizações que
         cooperam na execução do programa e na equipa NET ;
     e)  custos operacionais da equipa NET a uma taxa de aproximadamente 75 % ;
     Os saldos positivos resultantes das contribuições de países terceiros
     associados ( Suécia e Suíça ) no âmbito do programa com exclusão do JET ,
     destinar- se - ão à participação financeira , por parte da Comunidade , nas
     despesas referidas no ponto 3 .
4 . Após consulta do Comité Consultivo do Programa de Fusão , a Comissão poderá
     financiar , a uma taxa uniforme de cerca de 45 % nos termos da alínea a ) do
     ponto 3 ,   projectos que se insiram numa das áreas seguintes :
1
   JO no L 151 de 7.6.1978 , p. 10 .
 ---pagebreak---                                                                                     48
   a)   sistemas Tokamak e apoio ao JET ;
   b)   outras máquinas toroidais ;
   c)   aquecimento e injecção ;
   d)   NET e tecnologia da fusão .
   Se estes projectos pertencerem às áreas c ) e d ),            e se forem executados
   pela Indústria , a Comissão poderá financiá -los a uma taxa de 100 % nos
   termos da alínea c ) do ponto 3 .
   Por seu turno ,     todas as Associações terão o direito de participar nas
   experiências executadas com o equipamento construído nessas condições .
5. 0 total das contribuições dos membros do JET Joint Undertaking necessário
   para financiar os pagamentos do JET durante o período de 1987 a 1991 está
   estimado   em   531  milhões   de  ECUs .  Destinam -se     a   financiar  a  fase  de
   extensão do dispositivo       JET até obtenção do rendimento máximo ,            e seu
   funcionamento e     exploração .    Segundo os estatutos do JET ,           80% deste
   montante ,    equivalente   a   425 milhões de      ECUs ,     são  financiados pelo
   orçamento   comunitário .     Deste   montante ,     19    milhões   de   ECUs   foram
   autorizados pela Comissão antes de 1987 .        Os restantes 406 milhões de ECUs
   serão financiados do seguinte modo :
   . 378 milhões de ECUs da dotação do Programa para o JET ;
   . 28 milhões de ECUs como participação da Suécia e da Suíça pagos através
      do orçamento comunitário .
 ---pagebreak---                                                                                  49
                                    C ) FICHA FINANCEIRA
                         I. PROGRAMA FUSÃO ( com exclusão do JET )
1.    NÚMERO ORÇAMENTAL CORRESPONDENTE : 7310
2.    DESIGNAÇÃO : Fusão Termonuclear - Programa geral
3.    BASE LEGAL : Artigo 72 do Tratado Euratom
                      Decisão 85/201 /Euratom do Conselho < 1 )
                      e Regulamento previsto em 1987 .
4.    DESCRIÇÃO , OBJECTIVOS , JUSTIFICAÇÃO DO PROGRAMA      com inclusão do JET :
4.1 Descricâo
      O     programa  é   concebido    de  forma  a  prosseguir  a  investigação   e  o
      desenvolvimento no domínio da fusão termonuclear controlada e abrange
      todas     as  actividades   realizadas   nos  Estados-membros neste   domínio . A
      Suécia e a Suíça estão associadas ao programa .        Este incide especialmente
      no estudo do confinamento magnético do plasma e no estudo da tecnologia da
      fusão .
4.2 Obiectivos
       ( a ) Os objectivos a curto prazo do programa são :
             - estabelecer a base física e tecnológica necessária para a concepção
               pormenorizada do NET ( Next European Torus ), o grande dispositivo que
               constitui a próxima fase após o JET ,
 (1 )
       JO no L 83 de 25.3.1985 .
 ---pagebreak---                                                                                50
          - iniciar a concepção pormenorizada do NET antes do final do período
            do Programa se o banco de dados necessário já existir nessa altura ,
          - explorar o potencial de reactor de algumas linhas alternativas
            ( principalmente Stellarator , estrição de campo invertido ),
          - realizar um programa mínimo de confinamento por inércia .
    ( b ) 0 objectivo final deste programa é averiguar se a energia pode ser
          produzida a preços competitivos a partir de reacções de fusão nuclear
          entre núcleos de baixo peso atómico e , em caso afirmativo ,     construir
          conjuntamente      protótipos tendo em vista a produção          à escala
          industrial e a comercialização .
4.3 Justificação
    0 problema das fontes de energia a nível mundial e a longo prazo está
    longe de poder ser considerado resolvido .       A fusão termonuclear é uma das
    pouquíssimas fontes capazes de resolver este problema ou ,       pelo menos , de
    dar uma contribuição importante para a sua solução , com grande proveito
    para a Europa . Um reactor de fusão magnética utilizará um combustível que
    é virtualmente inesgotável e universalmente acessível , será dotado de
    características inerentes de segurança e garantirá um baixo índice de
    efeitos sobre o ambiente . São as seguintes as principais razões para a
    realização de investigação e desenvolvimento comunitários neste domínio :
    - o nível de recursos humanos e financeiros exigidos ,             que sugere a
       dificuldade    da   realização  à   escala   nacional  destes   trabalhos  de
       desenvolvimento ;
    - a longa duração do esforço ( que se prolongará pelo próximo século )
       necessário para chegar à construção do reactor ;
    - a existência de         uma  necessidade    colectiva ,  comum    a  todos  os
       Estados-membros ;
    - a realização de um mercado europeu para indústrias europeias no domínio
       de altas tecnologias ;
 ---pagebreak---                                                                                51
    - etn caso de êxito ,     a abertura de um amplo mercado comunitário para o
       reactor europeu ;
    - fornecer um parceiro potencial de dimensão comparável aos outros 3
       programas de fusão mundiais ,        fomentando ,    assim ,  a colaboração
       internacional no domínio da fusão ;
    - a qualidade do Programa Europeu de Fusão , cuja posição primordial é
       reconhecida a nível mundial e ao qual a Suécia e a Suíça estão
       plenamente associadas .
    A fusão    encontra -se ,  portanto ,  dentro dos critérios pertinentes ao
    Programa comunitário de I&D .
5.   INCIDÊNCIAS FINANCEIRAS TOTAIS DO PROGRAMA GERAL PARA 0 PERÍODO DE 1987 A
    1991
5.1 Incidências em matéria de despesas
5.1.1     Custos incorridos :
          - pelo orçamento das Comunidades :             616,0 milhões de ECUs
          - pelas administrações nacionais e
            outros sectores a nivel nacional
            ( estimativa ) :                         1 117.0 milhões de ECUs
                                       Total :       1 733,0 milhões de ECUs
     Os 616 milhões de ECUs incluem 83 milhões de ECUs autorizados antes de
     1987 dentro do programa de 1985-89 para trabalhos a serem executados
     depois de 1986 . A contribuição comunitária para 1987-91 constante da
     proposta de Regulamento do Conselho é , portanto , de 616-83 = 533 milhões
     de ECUs .
 ---pagebreak---                                                                               52
5.1.2      Fraccões e calendários plurianuais
         Em 1976 , o Conselho adoptou , sob proposta da Comissão , o princípio do
         "programa deslizante" juntamente com o programa 1976-1980 . 0 Conselho
         fixa   em   cada decisão   de   programa   o  montante  de  dotações  para
         autorizações atribuído ao programa ,      bem como o montante de dotações
         para autorizações remanescente do programa anterior . A fracção aberta
         para cada programa corresponde às dotações atribuídas menos as
         dotações remanescentes . As fracções agregadas abertas para um dado
         período constituem o conjunto dos créditos de que a Comissão dispõe
         para a execução dos programas durante esse período . Tendo em conta a
         contribuição proposta para o programa geral 1987-1991 , estes créditos
         elevam -se a 1180,0 milhões         de ECUs para o período 1976-1991 ,
         calculados do seguinte modo :
                                                            Fraccâo
         Programa 1976-80 :                                  124,0 milhões de ECUs
         Programa 1979-83 : 190,5 - 44,0
         ( dotações remanescentes do programa
         1976-80 ) :                                         146,5 milhões de ECUs
         Programa 1982-86 : 301,0 - 67,0 ( dotações
         remanescentes do programa 1979-83 ) :               234,0 milhões de ECUs
         Programa 1985-89 : 360,0 ( 1 > - 45,5 ( dota¬
         ções remanescentes do programa 1982-86 ) :          314.5 milhões de ECUs
              Total de fracções abertas para 1976-89 :       819,0 milhões de ECUs
(1 )
     Ver Comunicação da Comissão ao Conselho sobre o Programa Fusão ,     documento
     C0M(85 ) 789 final .
 ---pagebreak---                                                                  53
Programa 1987-91 proposto : 532,0 - 171,0
( dotações remanescentes previtas do programa
1985-89 ) :                                     362.0 milhões de ECUs
                               Total :         1181,0 milhões de ECUs
Os calendários que se seguem dizem respeito ao período de 1976 a 1991 ,
cobrindo os programas anteriores , o programa 1985-1989 em curso e o
programa 1987-1991 proposto :
 ---pagebreak---                                     Quadro . Programa Geral , autorizações (milhões de ECUs ) sem as contribuições de países terceiros ( Suécia e
                                            1976-85     1986
                                                                 I   1986
                                                                 [ Valores   I 1987 1 1988 ( 1989 1 1990 [ 1991
                                                                             «
                                                                                                                              lotai       lotai
                                           Execução   Execução   I transitai
                                                                       a                                                    1976-91      1987-1991
                                                                                      Despesas       estimadas                              (2 )
                                                                     (D
           Programas 1976 / 86             ** 9,0          8,0           2,0             -         -       -        -
                                                                                                                               <• 55,0        -
          Programa em curso 1985 /89         90,8        9*,1                  100,3     60,7     10,0     -        -
                                                                                                                               3-60,0     171 ,0
          Programa proposto 1987 / 91          -           -              -
                                                                                         56,0   100,0    113,0    93,0         562,0      362,0
          Toinl                         J  539,8    \
                                                        102,1            6,1   100,3    116,7   110,0    113,0    93,0      1 181,0       533,0
                                      Pagamentos (milhões de ECUs ) sem as contribuições de pafses terceiros ( Suécia e Suíça )
                                          1976-85      1986
                                                               11986
                                                               I Valorçs       1987 1 1988 | 1989 1 1990          1991       lotai      lotai
                                                      Execução       a                                          eanos
                                                                                                                           1976-91      1987-1991
                                          Execução              transita             Despesas estimadas        posteriores              e  anos
                                                                     (1 )                                                              jos^iore:
        Programas 1976 /86                 389,2         33,2           1,6    11 ,*    20,6     -        -        -
                                                                                                                             *59,0          35,0
        Programa em curso 1985 / 89         10,1         75,6           0,7    78,8     66,0   81,7     21 ,*     25,7       360,0        273,6
        Programa proposto 1987 / 91           -           -             -       -
                                                                                        10,2   * 0,0   115,0    196,8        362,0        362,0
         Tot « L                           399,3       108,8           2,3     93,2     96,8  121,7    136 ,*   222,5      1181,0         670,6
Notas : <1 > As dotações transitadas de 1986 fazem parte do Programa 1985-89 .                                                                      I
        (2 ) Os valo res nesta coluna não incluem quaisquer montantes transitados de 1986 para despesas em 1987 .                                  yj}
                                                                                                                                                    I
 ---pagebreak---                                                                               55
5.2 Modo de cálculo
     a ) Despesas com o pessoal
     Os efectivos propostos são os seguintes :
            Anos        I      A              B     I      C     I    Total
                        I                           I            I
                        I                           I            I
                        1                           I            I
            1987-91     I     73             29     I      3     I     105
                        I                           I            I
                        I                           I            I
     Os cálculos de despesas com o pessoal baseiam -se na despesa real em 1987
     aumentada de 4% por ano para os anos 1989-91 .       As dotações inscritas no
     orçamento para as despesas com o pessoal não têm em conta o facto de o JET
     reembolsar à Comissão as despesas com o pessoal do programa geral afecto
     ao JET .
     As despesas comunitárias relativas aos custos do pessoal são incluídas nas
     alíneas b ) e c ) seguintes .
     b ) Despesas de funcionamento administrativo e técnico e gestão
         Cobrem as despesas com viagens , deslocações em serviço , peritos , e a
         organização de reuniões , bem como a utilização do apoio administrativo
         e técnico . Incluindo o financiamento do Programa de Avaliação , na parte
         que diz respeito à fusão ( 1 > , e o custo do pessoal da Comissão afecto à
         Direcção Fusão em Bruxelas , estas despesas são calculadas em 14 milhões
         de ECUs financiados a 100% pelo orçamento comunitário . Isto representa
         1,4% da contribuição comunitária e 0,6% do custo total da I&D
         comunitária no domínio da Fusão , incluindo o JET .
(1 )
     0 custo do Painel de Avaliação mencionado na Secção VI da Exposição de
     Motivos está actualmente estimado em cerca de £meio milhão de ECUs^.
 ---pagebreak---                                                                                  56
     c)  Despesas contratuais
         i)    Contratos de associação .      Para o período 1987-91 ,     o custo da
              realização   do   programa    fusão   nos   laboratórios  associados   à
              Comunidade é calculado em 1 611 milhões de ECUs , incluindo o apoio
              destes laboratórios ao JET e NET ,      a sua actividade no domínio da
              tecnologia da fusão e as despesas com o pessoal destacado nos
              laboratórios    associados .    A   Comunidade   poderá  participar   no
              financiamento destas despesas nas seguintes percentagens :
              - apoio geral às despesas correntes e ao trabalho de base em
                tecnologia : cerca de 25% ;
              - apoio preferencial a acções prioritárias no domínio da física e
                da tecnologia ,   bem como aos trabalhos relativos ao JET e NET :
                 cerca de 45% ,
              - funcionamento administrativo e técnico da equipa NET :        cerca de
                 75% .
              As despesas autorizadas da Comunidade relativas à participação no
              financiamento das despesas das Associações estão           estimadas em
              429 milhões de ECUs < 1 ) .
         ii ) Contratos industriais .     Prevê -se um aumento do número de contratos
              de desenvolvimento industrial no âmbito do NET e da tecnologia da
              fusão e o desenvolvimento de métodos avançados de aquecimento do
              plasma . Em 1990 e 1991 , data em que o Projecto NET passará à fase
              de projecto de engenharia pormenorizado , terão de ser encomendados
              a empresas industriais protótipos de componentes do dispositivo
              NET . A Comunidade poderá financiar estes contratos a 100% , estando
              à disposição para esse efeito 74 milhões de ECUs .
(1 )
     Aos 429 milhões de ECUs devem adicionar -se 83 milhões dee ECUs autorizados
     antes de 1987 para o período de 1987 a 1989 .
 ---pagebreak---                                                                             57
       iii ) As despesas relativas à mobilidade do pessoal não pertencente à
             Comissão são calculadas em 6 milhões de ECUs que serão financiados
             a 100*4 pelo orçamento comunitário . É necessário prever um montante
             de 8 milhões de ECUs para financiar numa percentagem de cerca de
             42% o pessoal da Comissão destacado junto das Associações .
       iv )   É atribuído um montante de 2 milhões de ECUs para bolsas de
              investigação .
5.3 Dotações para autorizações não utilizadas remanescentes do programa de
    1985-89 :
    - Contribuição para o programa de 1985-89              360,0 milhões de ECUs
    - Menos : dotações autorizadas em 1985 e
      1986 ,  dotações transitadas de 1986 :             - 189.0 milhões de ECUs
    - Dotações não utilizadas , disponíveis
      para 1987-89 :                                       171,0 milhões de ECUs
5.4 Incidências em matéria de receitas
    - Impostos comunitários sobre os salários do pessoal da Comissão
    - Contribuição deste pessoal para o regime de pensão .
6.  FINANCIAMENTO DO PROGRAMA
    - Dotações inscritas nos orçamentos das Comunidades Europeias para os anos
      de 1976 e 1987 .
    - Dotações    a    inscrever  em   orçamentos   futuros   ( 1988  a  1991    e
      posteriormente ) .
 ---pagebreak---                                                                             58
7. REGIME DE CONTROLO A APLICAR
   Controlo cientifico :        - Comités de gestão constituídos por contratos
                                  de associação celebrados com os laboratórios
                                  nacionais .
                                - Comité    Consultivo    do     Programa   Fusão
                                  constituído   pela   Decisão   do  Conselho  de
                                  16.12.1980 .
   Controlo administrativo
   e financeiro :               - Comités de gestão .
                                - DG do Controlo Financeiro no que se refere à
                                  execução do    orçamento e    à regularidade e
                                  conformidade das despesas      e a Divisão de
                                  Contratos    assistida    por   sociedades   de
                                  auditoria da confiança da Comissão ( DG XII ).
 ---pagebreak---                                                                               59
                                 II ) PROJECTO JET
1.  NÚMERO ORÇAMENTAL CORRESPONDENTE : 7311 .
2.  DESIGNAÇÃO : Participação no JET , Joint Unertaking .
3.  BASE LEGAL : Artigos 45a a 512 do Tratado Euratom e
                  artigo 92 dos Estatutos do JET ,
                  Decisões do Conselho 78/ 470 /Euratom de 30.5.1978
                  ( JO n2 L 151 de 7 de Junho de 1978 , p. 8 ), 30/31 8/Euratom
                  de 13.3.1980 , 81 /380 /Euratom de 19.5.1981 ,
                  82/350/Euratom , 85/201 /Euratom e Regulamento do Conselho
                  previsto em 1987 .
4.  DESCRIÇÃO , OBJECTIVOS E JUSTIFICAÇÃO DO PROJECTO :
4.1 Descrição
    Construção ,   operação e exploração ,      como parte do programa fusão da
    Comunidade e em benefício dos participantes neste programa ,      de uma grande
    máquina tórica do tipo Tokamak e das suas instalações anexas ( Joint
    European Torus - JET ), de forma a ampliar a gama de parâmetros aplicáveis
    às experiências de fusão termonuclear controlada até condições próximas
    das exigidas num reactor termonuclear .
 ---pagebreak---                                                                                   60
4.2 Obiectivos
      Obter e    estudar um plasma de dimensões e         em condições próximas das
exigidas num reactor termonuclear .       Este objectivo exige 4 áreas de actividade
      principais :
      i)    a  evolução   do  comportamento    do plasma   quando   os  parâmetros se
            aproximam dos necessários num reactor ;
      ii )  a interaccão plasma-parede nestas condições ;
      iii ) o estudo do aquecimento do plasma ;
      iv )  o  estudo  da  produção  e   do  confinamento   de  partículas   alfa  e do
            consequente aquecimento do plasma .
4.3 Justificação
      A    execução   do   Projecto    JET    constitui   uma    etapa   essencial   no
      desenvolvimento do programa fusão da Comunidade .          No que se refere ao
      objectivo final deste programa e à sua justificação ,          remete -se para a
      Parte I , ponto 4.3 , da presente ficha financeira .
5.    INCIDÊNCIAS FINANCEIRAS TOTAIS DO JET PARA 0 PERÍODO DE 1987 A 1991
5 . 1 Incidências no Programa - Quadro de 1987-91
Para o período       do programa de 1987 a 1991 ,      o JET necessita dos seguintes
financiamentos :
     Contribuição para o Programa de 1987-91                 378.2 milhões de ECUs
     Fundos remanescentes do Programa Fusão 1985-89          209.2 milhões de ECUs
     Nova contribuição necessária para 1987-91                169,0 milhões de ECUs
 ---pagebreak---                                                                            61
Estes valores não incluem a participação da Suécia e da Suíça .
5.2 Modo de cálculo
Na sua reunião de Março de 1987 ,        o Conselho JET aprovou um Plano de
Desenvolvimento de Projecto e uma Estimativa dos Custos do Projecto cobrindo
a restante duração do projecto até 1992 . 0 financiamento associado do JET para
o período de 1987 a 1991 foi calculado em :
                  Autorizações                490,6 milhões de ECUs
                  Pagamentos                  542,5 milhões de ECUs
                  Contribuições dos Membros   531,3 milhões de ECUs
Estas estimativas têm em conta uma taxa de inflação constante de 4% ao ano a
partir dos índices de inflação médios do JET para o ano de 1986 .        80V. das
Contribuições dos membros ( 425,0 milhões e ECUs ) deverão ser financiados pela
Comunidade .  Uma vez que 19,2 milhões de ECUs foram autorizados antes de 1987
para o período de 1987-91 ,   as autorizações estimadas para esse período são ,
portanto , de 405,8 milhões de ECUs .
Estes 405,8 milhões de ECUs serão financiados do seguinte modo :    27,6 milhões
de ECUs previstos como participação da Suécia e da Suíça no JET pagos pelo
orçamento comunitário , restando 378,2 milhões de ECUs a serem financiados
directamente pela Comunidade dentro da sua contribuição para o Programa de
1987-91 . 0 método de cálculo para as participações da Suécia e da Suíça
encontra -se descrito na Parte III do presente relatório financeiro .
0 cálculo é apresentado no quadro junto e resumido a seguir :
 1987-91 Contribuição para o Programa                 378,2 milhões de ECUs
Participação da Suécia e da Suíça                      27,6 milhões de ECUs
Montantes autorizados antes de 1987
para o período 1987-91                                 19.2 milhões de ECUs
 ---pagebreak---                                                                          - 62
Contribuição de 80% dos Membros do JET
para o período 1987-91                                425,0 milhões de ECUs
Contribuições da Organização-Hospedeira ( 10% )       106,3 milhões de ECUs
e de Membros do JET que têm contratos de
Associação com a EURATOM ( 10% )                      _
Contribuições dos Membros para o JET para o
período 1987-91                                       531,3 milhões de ECUs
5.3 Incidências em matéria de receitas
Impostos comunitários sobre os salários dos agentes temporários .
6.    FINANCIAMENTO DO PROJECTO :
Dotações inscritas nos orçamentos das Comunidades Europeias de 1976 a 1987 .
Dotações a inscrever em orçamentos futuros ( 1987 a 1991 ).
7 . REGIME DE CONTROLO A APLICAR
( A ) Controlo científico : Conselho do JET
                            Comité Consultivo do Programa Fusão
( B ) Controlo Administrativo
      e Financeiro :        Conselho do JET
                            Tribunal de Contas .
 ---pagebreak---                                                                                63
                                      Notas ao Quadro
( 1 ) Todos os valores da parte superior do quadro correspondem ao Plano de
       Desenvolvimento do Projecto e Estimativa dos Custos aprovados pelo
        Conselho do JET em Março de 1987 .
( 2 ) As Contribuições dos Membros para o período 1987-91 foram calculadas a
        partir do perfil dos pagamentos estimados através da subtracção de
        estimativas de receitas diversas , principalmente juros bancários .
( 3 ) A contribuição JET no Programa Fusão de 1985-89 ,                incluindo as
        participações da Suécia e da Suíça , totalizaram 330,0 milhões de ECUs . As
        participações suecas e suíças foram estimadas em 23,9 milhões de ECUs ,
        sendo os restantes 306,1 milhões de ECUs a participação directa da
        Comunidade .
( 4 ) As dotações transitadas de 1986 dizem respeito ao programa 1985-89 .     Estas
        dotações para pagamentos transitadas pelo JET já foram financiadas pelas
        contribuições dos Membros em 1986 .
 ( 5 ) No final de 1986 ,    as contribuições dos Membros ao JET totalizaram 633,8
        milhões de ECUs dos quais 80% , o equivalente a 507,1 milhões de ECUs ,
        foram financiados pela Comunidade . Dado que 526,3 milhões de ECUs tinham
        sido autorizados até essa data , a quantia autorizada relativamente ao
        período posterior a 1986 era , portanto , de 19,2 milhões de ECUs .
 ( 6 ) Do total de 531,3 milhões de ECUs de contribuições dos Membros para o
        período de 1987-91 , 80% , equivalente a 425,0 milhões de ECUs , serão
        financiados pela Comunidade .       Dado que 19,2 milhões de ECUs foram
        autorizados antes de 1987 para aquele período , as autorizações estimadas
        para esse período elevam -se a 405,8 milhões de ECUs .
  ( 7 ) Os valores nesta coluna não incluem quaisquer montantes transitados de
         1986 para despesas em 1987 .
 ---pagebreak---              Quadro : Perfis financeiros tanto do JET Joint Undertaking como da Participação Comunitária no JET
   Milhões de ECUs prevendo-se      1976-85       1986      1986(4 )  198 7 1988   1989   1990    1991    Total         Total
   uma taxa de inflação conti ¬    Execução      Execução   Valores                                       1976-91      1987-91
                                                                             Despesas estimadas
   nua de 4% ao ano                                       a transitar                                                    (7)
  ORÇAMENTOS JET ( 1 )
  Autorizações                     600,5         100,2           30,4  33,7 125,1  106.8   39,9    80,1    1221 ,7       490,6
  Pagamentos                       542.3          95,3           12,3 104,4 108,5  118,1  115,4    96,1    1192,4        542,5   ,
  Contribuições dos Mentiros ( 2 ) 548,5 · ;15 )
                                                  35,3            -
                                                                       90,3 106,4  116,6  113,9    94,6    1165,1        531 , 3W
  PARTICIPAÇÃO COMUNITARIA
  Autorizações ( excl . CH+S )
  . Programas      1976-1986       393,3                                                                    393,3
  . Programa       1985-1989         23,9         73,0              -
                                                                       75.1   78,7   55.4     -      -  ·   306,l ( 3i   209,2
  . Programa       1987-1991           -            -               -    -     -
                                                                                     12,9  85,3   70,8      169,0        169,0
  Total   ( excluindo CH+S )       417,2          73,0              -
                                                                       75,1   78.7   68,3  85 , 3 70,8     868,4        378,2
  Suécia e Suíça                     31,1          5,0              -
                                                                        5,4    5,7    5,8   5,8     4,9      63.7         27,6
  Total ( incluindo CH+S )         448 , 3 ( b )  78,0 ^*           -
                                                                      80,5    84,4   74,1  91,1   75,7      932,1       405,8 ^
  Pagamentos ( excluindo CH+S )
                                   393,3                                 -       -     -     -       -
                                                                                                            393,3            -
  . Programas 1976-1986
                                                    -               -
  . Programa 1985-1989               14.3        63,2             2,8 75,1    78,7   72,0    -       -
                                                                                                           306,1        225,8
  . Programa 1987-1991                                                                                     169,0        169,0
                                       -            -              -     -      -
                                                                                     12,9 85,3    70,8
                                                     • .
  Total ( excluindo CH+S )         407,6         63,2             2.8 75,1   78,7    84,9 85,3    70,8     868,4        394,8
>
  Suécia e Suíça                    31,1           3,0                  5,4    5,7    5,8   5,8    4,9       63,7         27,6
  Total   ( incluindo CH+S )                                                 84,4    90,7 91,1    75,7     932,1        422,4
                                   438,7         68,2             2,8 80,5
 ---pagebreak---                                                                              65
III . Contribuições de países terceiros associados ao programa fusão
1.    Programa geral
1.1   Período de 1976-1986
As contribuições recebidas são estimadas em               42 milhões de ECUs
Menos : despesas comunitárias para a execução
     dos acordos de cooperação , estimadas em :         - 25 milhões de ECUs
Saldo positivo disponível para o programa geral
de fusão , estimado em :                                  17 milhões de ECUs
0 montante de 17 milhões de ECUs foi utilizado para manter ao nível de 25% o
apoio geral às Associações com a Comunidade .
1.2 Período de 1987-1991
A participação financeira da Suécia e da Suíça no Programa Geral será
calculada , como anteriormente , com base em pagamentos comunitários ao Programa
Geral e proporcionalmente aos seus Produtos Internos Brutos em relação ao
Produto Interno Bruto da Comunidade .
Dado que os contratos de associação negociados actualmente com a Suécia e com
a Suíça terminarão em 31.12.1989 , não é possível estimar     as despesas nestes
países até ao final de 1991 .        Prevê -se que ambos os    países participem
fortemente no programa de tecnologia da fusão , em expansão .  Prevê -se portanto ,
que o saldo positivo diminua ou mesmo desapareça . Caso se     verifique qualquer
saldo positivo , a Comissão propõe a sua utilização para      o financiamento de
despesas nas Associações com a Comunidade .
Com a adesão de Espanha à Comunidade Europeia em 1.1.1986 ,    a sua contribuição
para o Programa Fusão enquanto País Terceiro associado deixou de existir a
partir dessa data .
 ---pagebreak---                                                                            66
2.  JET
2.1 Período 1976-86
A participação da Suécia e da Suíça no JET para este período está estimada em
36,1 milhões de ECUs .
2.2 Período 1987-91
Supondo que :
    as dotações para pagamentos constantes do calendário plurianual ( ver Parte
    I , 5.1.2 ) para 1987-91 serão inscritas nos orçamentos para estes anos ,
    o conjunto dos Produtos Internos Brutos da Suécia e da Suíça serão ,      em
    média , equivalentes a 7% ao da Comunidade ,
    a Suécia e a Suíça se manterão plenamente associadas ao Programa Fusão
    para o período 1987-91 ,
as contribuições da Suécia e da Suíça podem ser estimadas em 27,6 milhões de
ECUs .
 ---pagebreak---                                                                               - 67 -
                       D ) PARECER DO COMITÉ CIENTÍFICO E TÉCNICO
            Sobre o Programa de Investigação e Formação ( 1987-1991 ) no
                        domínio da Fusão Termonuclear Controlada
Durante a reunião de 12 de Maio de 1986 , o CCT examinou o projecto de
Orientação do Programa-quadro das Actividades Comunitárias de I&D ( 1987-1991 ).
0 CCT examinou especialmente as propostas relativas à fusão termonuclear
controlada e encarregou um pequeno grupo de trabalho de formular um parecer de
carácter geral a esse respeito ,            enquanto espera pela discussão mais
pormenorizada prevista para A de Julho de 1986 por ocasião do exame pelo CCT :
-   do projecto      de proposta      de um  programa de    investigação quinquenal
    ( 1987-1991 ) no domínio da fusão termonuclear controlada ( doc . XII / 475 ).
-   e do projecto de proposta de uma alteração dos estatutos da Empresa Comum
    JET com vista a prolongar a existência dessa Empresa até 31 de Dezembro de
    1992 ( doc . XII / 498 ) .
Os pareceres emitidos pelo CCT para esses dois projectos em A de Julho são a
seguir referidos .
A fusão termonuclear controlada pode constituir ,        a longo prazo ,   uma fonte
preciosa para o aprovisionamento energético da Comunidade .           No entanto , e
apesar dos progressos importantes já realizados , serão ainda necessários pelo
menos trinta anos para atingir o estádio do reactor de demonstração .              Um
esforço dispendioso de tão prolongada duração só é aceitável se as actividades
de investigação sobre a fusão na Comunidade continuarem totalmente integradas
num programa bem coordenado . Se na execução de um tal programa houver um
grande      cuidado no que respeita aos aspectos económicos e se se evitar
duplicações inúteis ,          pode-se esperar levar a fusão até ao estádio
pré - industrial com uma despesa que , apesar da duração bastante mais longa das
actividades de investigação , não ultrapassaria o esforço financeiro concedido
no caso da cisão .
A física , incluindo a tecnologia que lhe está associada , ocupa o primeiro
 lugar nas actividades de investigação sobre a fusão . 0 JET é , neste domínio , a-
 instalação mais eficiente , cujo sucesso contribuiu significativamente para
 ---pagebreak---                                                                                 68
                       (*)
tornar a Comunidade        no líder incontestável a nível mundial .      A construção
da máquina foi realizada dentro dos prazos e do orçamento fixados e a fase
inicial de funcionamento ,    com aquecimento por resistências apenas ,      registou
melhores resultados do que se previa .
No  entanto ,    na  fase  seguinte iniciada   em 1985 ,     a utilização de tipos
adicionais de aquecimento permitiu ,      é certo ,    o aumento da temperatura do
plasma ,  mas não foi possível evitar a redução do tempo de confinamento já
observada com outras máquinas . Para ultrapassar esta dificuldade e para dar ao
plasma características que justifiquem o funcionamento com trítio ,                são
propostos alguns     equipamentos adicionais bem como       o adiamento do fim do
funcionamento do JET de 31       de Maio de 1990 para 31      de Dezembro de 1992 ,
mantendo -se a despesa anual ao nível de 1986 .        0 CCT sublinha a urgência da
decisão sobre a continuação da Empresa Comum ,        da qual depende ,   a partir de
agora , o bom desenvolvimento do programa do JET .
0 CCT emitiu um parecer favorável sobre as propostas apresentadas para o JET
tanto no que respeita à continuação da Empresa Comum como à manutenção do seu
orçamento .   É claro que não é absolutamente certo que os vários equipamentos
adicionais propostos sejam eficazes . No entanto , o CCT considera que um atraso
na sua entrada em serviço poderia ser muito prejudicial para o programa em
geral e provocar um aumento significativo das despesas ,          devido ao elevado
custo do funcionamento de base do JET .
A duração de 12 anos fixada inicialmente para a Empresa JET implicou um
calendário muito apertado .    A continuação proposta por 2 anos e 7 meses impõe ,
novamente ,   restrições severas , mas o CCT considera que é importante sublinhar
o carácter exemplar de uma limitação estrita do tempo de duração da Empresa
Comum em comparação com todas as outras grandes instalações internacionais de
investigação fundamental ou aplicada .
Os programas de física realizados nas Associações são indispensáveis como
apoio ao JET para determinados estudos que não podem ser efectuados no JET e
para a exploração de outras configurações para além do TOKAMAK .               Várias
máquinas de dimensão média estão em vias de serem terminadas .          Algumas delas
apresentam     características   únicas  no   mundo .     0  CCT  considera    que   o
     A Suécia e a Suíça aderiram aoo programa comunitário respectivamente em
     1976 e 1978 .
 ---pagebreak---                                                                            69
financiamento proposto para esta rubrica é muito razoável e está bem adaptado
aos programas já iniciados . É conveniente notar que é neste domínio que as
tentações de dispersão e de duplicação são maiores e que perante tal situação
importa não ceder . Em especial , o funcionamento de dispositivos de dimensão
média deve ser submetido a uma programação tão rigorosa como a do JET .
Um programa metódico sobre a tecnologia da fusão no âmbito comunitário só foi
lançado em 1982 . 0 seu objectivo é a aquisição de outros conhecimentos , além
daqueles do domínio da física , necessários para avaliar a viabilidade de
vários conceitos de reactores de fusão .   Este programa pôde arrancar com meios
relativamente limitados ,   apoiando -se em competências e em meios de ensaio
criados para a aplicação da energia de cisão . De momento , a tarefa mais
urgente é a aquisição de conhecimentos técnicos necessários para o projecto
NET , que é definido como a única etapa intermédia entre o JET e um reactor de
demonstração . Em 1990 , na altura da revisão do programa 1987-1991 , espera -se
dispor de suficientes dados físicos e técnicos para poder tomar a decisão de
iniciar o projecto pormenorizado do NET e o desenvolvimento associado de
componentes protótipos . 0 CCT considera que actualmente não é conveniente
prejudicar uma tal decisão que , no momento oportuno , deve ser objecto de uma
proposta da Comissão ao Conselho .      Assim ,  o CCT apresenta a proposta de
reservar para as rubricas NET e Tecnologia a quantia total de 91+166 milhões
de ECUs que não prejudica a decisão de iniciar em 1990 o projecto
pormenorizado do NET e que assegura o financiamento da equipa NET para a
totalidade do programa ( cf . Anexo I , quadro 1 , coluna da esquerda ). Tal
corresponde a um orçamento total para a fusão de 1.059 milhões de ECUs , ao
qual o CCT é favorável , e que retoma a proposta apresentada pela Comissão para
o programa 1987-1991 .
A estes trabalhos vêm juntar-se as activiades do CCI consagradas à fusão . 0
CCT lamenta que , por motivos formais , os pormenores dessas actividades sejam
objecto de uma discussão e de um parecer separados do CCT . 0 CCT insiste para
que as actividades do CCI no domínio da fusão sejam julgadas com os mesmos
critérios que as actividades análogas do programa a custos repartidos .
 ---pagebreak---                                                                                 70
                                       PARECER
                   do Comité Consultivo o Programa Fusão ( CCPF )
      relativo ao projecto de proposta de programa quinquenal ( 1987-1991 )
           de fusão termonuclear controlada , adoptado na sua reunião de
                                 19 de Junho de 1986
Após discussão do projecto de proposta de programa ao longo de três sessões
consecutivas ,   o CCPF aprova o conteúdo científico e técnico da proposta ,       que
considera perfeitammente      coerente  com os   objectivos a     longo prazo   e   as
modalidades de     realização do    Programa Fusão    tal   como   foram previamente
definidas pelo Conselho de Ministros .
0 programa inclui três componentes principais :       o JET , o trabalho de física e
de engenharia dos plasmas nas Associações e NET/Tecnologia .          0 CCPF apoia a
recomendação no sentido de aumentar a duração da Empresa Comum JET até 31 de
Dezembro de 1992 a fim de tirar proveito dos êxitos alcançados na evolução do
projecto .
Com base na análise pormenorizaa de custos feita pela Comissão e parceiros a
ela associados ,    o CCPF considera que o pacote financeiro proposto é adequado
ao conteúdo científico e técnico do programa proposto .
0 CCPF apoia a ideia básica subjacente à proposta de programa , cujo principal
objectivo é estabelecer a base física e tecnológica para a próxima fase .         Isto
significa que , aquando da próxima revisão do programa , poderá ser feita a
proposta de dar início ao projecto pormenorizado de engenharia do NET . Para
uma decisão desta importância , o CCPF recomenda que a Comissão procure , quando
tal for oportuno , o conselho de um painel independente .
Na linha do parecer expresso em Dezembro de 1985 ,        o CCPF reconhece o êxito
demonstrado pelo Programa europeu de Fusão ,     totalmente integrado ,    que faz da
Europa um     parceiro   importante  em qualquer    esquema de     ampla colaboração
internacional no domínio da fusão ,     e exprime novamente o seu receio de que os
objectivos do Programa de Fusão não venham a ser atingidos se o nível dos
fundos postos à sua disposição for consideravelmente reduzido em comparação
com a proposta . Caso tal se verificasse , o programa teria de ser completamente
reavaliado .
 ---pagebreak---                                                                          71
Considerando que a fusão já adquiriu um vasto conteúdo em matéria de alta
tecnologia e gerou resultados que vieram a beneficiar outros ramos da ciência
e da indústria europeia , o CCPF apoia a proposta da Comissão no sentido de
ampliar a participação da indústria . Esta participação terá de aumentar
substancialmente quando o NET entrar na fase de projecto de engenharia .
A mobilidade do pessoal científico entre os vários laboratórios de fusão
atingiu um nível significativo e é de especial importância para os países que
não dispõem dos seus próprios programas de fusão . Por esse motivo , o CCPF
apoia o esquema de mobilidade , bem como o programa de bolsas que fazem parte
da proposta .
 ---pagebreak---                                                                            72
COMISSÃO DAS COMUNIDADES EUROPEIAS
              Proposta de Decisão do Conselho que aprova alterações aos
              estatutos da Joint European Torus ( JET ), Joint Undertaking
                     ( apresentada pela Comissão/
 ---pagebreak---                                                                 73 .
                         A ) EXPOSE DES MOTIFS
1. Le Conseil a créé l' Entreprise Commune JET pour une durée de 12 ans
   à partir du 1er juin 1978 jusqu' au 31 mai 1990 . Les objectifs de
   l' entreprise sont décrits comme suit dans les statuts :
   "Construire , faire fonctionner et exploiter   une grande machine , un
   dispositif torique du type Tokamak , pour       étendre la gamme des
   paramètres applicable aux expériences de       fusion thermonucléaire
   contrôlée jusqu' à des conditions proches de  celles requises dans un
   réacteur thermonucléaire ".
2. Le succès du JET est indispensable pour l' avant-projet et pour la
   construction de la machine prochaine NET (Next European Torus ) et ,
   partant , au programme " fusion" européen dans son ensemble .
3. Le JET poursuit quatre objectifs scientifiques qui figurent dans le
   rapport EUR-JET-R5, "Le projet JET - Proposition d 'Avant-projet ",
   1976 , auquel il est fait explicitement référence dans les Statuts
   du JET de 1978 . Ces objectifs restent inchangés :
   a)    étudier la manière dont se comportent le confinement et les
         propriétés du plasma lorsque les dimensions et les paramètres
         se rapprochent de ceux requis pour un réacteur ;
   b)    examiner et contrôler l' interaction plasma-paroi et l' influx
         d' impuretés dans ces conditions ;
   c)    faire la démonstration de techniques de chauffage efficaces ,
         capables de produire des températures élevées ;
   d)    étudier la production et le confinement des particules alpha
         et le chauffage du plasma qui en résulte .
 ---pagebreak---                                                              74 .
4. Pour atteindre ces objectifs , le projet se déroule par phases
   successives :
         Phase 0 : Construction de la Machine
         La machine a été construite en cinq ans comme prévu , entre
         1978 et 1983 .
         Phase 1 : Plase de Chauffage Ohmique
         Les principaux objectifs de cette phase , qui est maintenant
         terminée , étaient la mise au point de la machine et de ses
         principaux systèmes , ainsi que la production d' un plasma
         propre d' hydrogène se prêtant à des études de chauffage
         additionnel au cours des phases ultérieures .
   -     Phase 2 : Etudes de Chauffage Additionnel         et     de Pleine
         Optimisation
         Au cours de cette phase , qui a commencé en 1985 comme prévu ,
         un chauffage additionnel de plus en plus important sera
         installé sur la machine . Les principaux objectifs de cette
         phase seront la réalisation des performances maximales de la
         machine et d' atteindre les paramètres du plasma nécessaires
         pour passer à la phase finale du programme .
         Phase 3 : Phase du Tritium
         Si la phase 2 est couronnée de succès , la phase du tritium
         pourra commencer . Cette phase , qui exige deux années , sera
         consacrée à l' étude de la production de particules alpha dans
         des plasmas de deutérium et de tritium . L' objectif ultime est
         de parvenir à un niveau significatif de chauffage par les
         particules alpha .
5. L' Etat actuel du JET
   La machine a été construite avec les crédits et dans les délais
   prévus . La phase du chauffage ohmique qui a commencé avec le
   premier plasma en juin 1983 a été terminée avec succès au cours de
   la deuxième moitié de 1984 , comme prévu . Tous les systèmes
   mis en service ont fonctionné suivant les spécifications établies
   et les résultats physiques ont donné entière satisfaction . En fait ,
   un courant contrôlé de plasma de 5 millions d' ampères (MA) a été
   obtenu , en comparaison de la valeur projetée de 4,8 MA . Avec le
   seul chauffage ohmique , le JET a atteint des températures de plasma
   de presque 40 millions de degrés C° et des temps de confinement
   d' environ 0,9 seconde .
 ---pagebreak---                                                                  75 .
   En 1985 , le programme de chauffage additionnel a commencé avec
   l' application réussie du chauffage HF suivie , en 1986 , de
   l' introduction du chauffage par injection de neutres . En novembre
    1986 , une puissance totale de 18 MW était couplée au plasma , au
   moyen des deux méthodes de chauffage additionnel et des
   températures maximales de 145 millions de degrés C° ont été
   atteintes . Dans la configuration habituelle du limiteur matériel ,
   on obtient avec le chauffage additionnel une dégradation du temps
   de confinement par rapport à celui obtenu du chauffage ohmique . Des
   premières expériences avec la configuration de limiteur magnétique
    (point X) ont donné , fin 1986 , des résultats encourageants qui
   laissent entrevoir une possibilité d' éviter une telle "dégradation"
    (mode H ) .
6. Projets futurs
   Comme on a l' intention de porter la puissance de chauffage totale à
   40-45 MW , il est extrêmement important de trouver des moyens
   d' éviter le phénomène de "dégradation du confinement " qui a été
   observé jusqu' à présent lorsqu' on appliquait le chauffage
   additionnel . Des études théoriques ont suggéré pendant un certain
   temps - suggestion qui est maintenant confirmée par les expériences
   du JET et ailleurs - que des moyens pouvaient être mis au point
   pour éviter cette dégradation . En fait , une série de nouvelles
   mesures expérimentales se dessinent , mesures qui devraient
   permettre au JET de tirer le meilleur profit des possibilités de
   performance de la machine . Ces développements couvrent les quatre
   sujets suivants :
   ( i)    accroissement de la densité centrale du plasma par injection
           de glaçons ;
   ( ii ) extraction du plasma et contrôle de la densité aux bords ;
   ( iii ) meilleur contrôle de l' interaction plasma /paroi grâce à la
           modification de la configuration magnétique ( points X ) ;
   ( iv ) contrôle du profil du courant dans le plasma .
   Le but de ces mesures est de produire une configuration de plasma
   stable avec des densités et des températures plus élevées pour un
   temps de confinement suffisant . Pour cela , des équipements
   supplémentaires sont requis dont le coût a été estimé à 70 MioECU
   au plus aux prix de 1986 . L' augmentation nette des frais
   d' investissements , compte tenu d' une réduction de quelque 25 MioECU
   du coût de la phase d' extension aux performances élargies , est de
   45 MioECU environ , ce qui représente une augmentation de moins de
    10 % des frais d' investissements totaux du projet .
 ---pagebreak---                                                                 76 .
Ces développements pourraient avoir . lieu sans accroître le taux
actuel de dépenses du JET ( entre 100 et 105 MioECU par an aux prix
de 1986 ).
Ces équipements supplémentaires doivent être opérationnels avant
que l' on puisse aborder la phase finale du programme JET , celle du
tritium . Leur avant-projet , fabrication et installation exigent du
temps et pourraient donc retarder le démarrage de la phase du
tritium par rapport au calendrier initial . Pour réduire à un
minimum l' extension du programme JET et garder son élan , la mise en
oeuvre de ces mesures ne doit pas être retardée . Le démarrage
rapide de ces travaux n' a de sens que dans le contexte d' une
prolongation de l' Entreprise Commune , pour permettre l' exploitation
de ces équipements supplémentaires . C' est pour cette raison que le
Conseil du JET , lors de sa réunion d' octobre 1985 , a conclu que
l' exploitation du JET devrait se poursuivre jusqu' à la fin de 1992 ,
afin que le NET et le programme "Fusion" dans son ensemble puissent
tirer pleinement parti du potentiel du JET . La Commission en a
informé   le  Conseil   des  Ministres   dans  sa   Communication    sur  le
programme " Fusion " ( document ( 85 ) 789 final du 23 décembre 1985 ) en
décembre 1985 . Le Conseil du JET a décidé à l' unanimité , lors de sa
réunion de mars 1986 , de prendre les mesures formelles nécessaires
pour prolonger l' Entreprise Commune de deux ans et sept mois , du 31
mai 1990 au 31 décembre 1992 et de modifier en conséquence
l' article 19 des Statuts du JET . La Commission propose que le
Conseil    des Ministres ,   conformément    à  l' article  50 du     traité
EURATOM , approuve cette modification des Statuts du JET .
 ---pagebreak---                                                                                  - 77 -
                                            B ) PROPOSTA
                                                  de
                                     DECISÃO DO CONSELHO
            que aprova alteraçôes aos estatutos da Joint European Torus
                               ( JET ), Joint Undertaking .
0 CONSELHO DAS COMUNIDADES EUROPEIAS ,
Tendo em conta o Tratado que institui a Comunidade Europeia da Energia Atómica
e, nomeadamente, o seu artigo 50a,
Tendo em conta a proposta da Comissão,
Considerando que ,      a fim de dar execução ao projecto JET,                 o Conselho
                          -_ _     i g 14 i c   . . _ (1
                                                      Ni )
                                                         /
constituiu , pela Decisão 78 / 471 / Euratom               , a Joint European Torus ( JET ),
Joint Undertaking , e adoptou os seus estatutos, alterados mais tarde pelas
Decisões 79/720/ Euratom ^ e 83/310/Euratom ^
Considerando que, para atingir os objectivos do projecto JET tal como são
definidos na Decisão 78 / 471 / Euratom, é necessário mais equipamento, cuja
manufactura, operação e exploração não pode ser realizada dentro do período de
duração da Joint Undertaking tal como está actualmente definida nos estatutos
da JET;
Considerando que o Conselho JET aprovou um prolongamento da Joint Undertaking
até 31 de Dezembro de 1992 e a alteração correspondente aos estatutos da JET,
DECIDE :
 (1 )
      JO na L 151 de 7.6.1978 , p . 10
 (?)
      JO n a L 213 de 21.8.1979, p . 9 .
 (3) JO na L 164   de 23.6.1983 , p . 35 .
 ---pagebreak---                                    Artigo 1 a
São aprovadas as alterações aos estatutos da "Joint European Torus ( JET ),
Joint Undertaking", anexas à presente decisão .
                                   Artigo 2a
A presente decisão entra em vigor no dia seguinte ao da sua publicação no
Jornal Oficial das Comunidades Europeias .
Feito em                                          Pelo Conselho
                                                  0  Presidente
 ---pagebreak---                                                                        79 -
                                       ANEXO
O na 1 do artigo 19a dos estatutos da Joint European Torus ( JET ), Joint
Undertaking passa a ter a seguinte redacção :
" 19.1 . A Joint Undertaking é constituída até Dezembro de 1992 ."
 ---pagebreak---                                                                         - 80 -
                              C ) FICHA FINANCEIRA
0 custo total da JET e as contribuições financeiras do orçamento comunitário
para a JET durante todo o período proposto de duração da Joint Undertaking
são estabelecidos na ficha financeira anexa à Proposta de Regulamento do
Conselho que adopta um programa de investigação e formação 1987-1991 no
domínio da fusão termonuclear controlada .     Esta ficha limita -se aos custos
suplementares que resultam da introdução proposta de equipamento adicional e
do prolongamento da Joint Undertaking . As despesas suplementares são
calculadas a preços de 1986 do seguinte modo :
. Custos de capital do equipamento adicional :               70 milhões de ECUs
. Prolongamento da operação da JET durante 2 anos
  e 7 meses :                                              190 milhões de ECUs
. Menos : Redução nos custos da fase de alargamento até
  ao pleno funcionamento                                   - 25 milhões de ECUs
. Custos suplementares líquidos                            235 milhoes de ECUs
0 volume dos custos de capital com equipamento adicional diminuirá entre 1987
a 1990 , bem como os custos restantes do alargamento até ao pleno funcionamento
e as despesas de operação da JET . A despesa com o prolongamento da operação da
JET diminuirá entre 1990 e 1992 .    Nos termos do artigo 9a dos estatutos JET ,
80% dos custos suplementares ( 188 milhões de ECUs ) deverão ser financiados
pelo orçamento comunitário (na 7311 ). A distribuião orçamental anual é
apresentada na ficha financeira da proposta para o Programa Fusão 1987-1991 .
 ---pagebreak---                     COMISSÃO DAS COMUNIDADES EUROPEIAS
Relatório sobre " Impacte Ambiental e Perspectivas Económicas da Fusao "
                Preparado pelos Serviços da Comissão e
           apoiado pelo Comité Consultivo do Programa Fusão
 ---pagebreak---                                                                           82
      Environmental Impact and Economic Prospects of Nuclear Fusion
Following a request from both Parliament and Council , the Commission has
asked a group of European experts to establish a technical report on the
" Environmental Impact and Economic Prospects of Nuclear Fusion".
The Commission is pleased to forward this technical report , together
with a less technical summary on the state of the art in this matter
that has been endorsed by the Consultative Committee for the Fusion
Programme .
The Commission is conscious that the results of this work and the views
expressed represent the present stage of knowledge in an evolving field .
Indeed ,   as  the   development   of   nuclear   fusion  moves    from   the
demonstration of the scientific principles to the demonstration of the
technological    feasability ,  research   on   safety , environmental    and
economic aspects of fusion will grow in the future . This will permit to
refine in due course the views expressed at this stage .
The Commission is also aware that decisions of major importance will
have to be taken in a few years time in the field of fusion , such as :
launching   the engineering design of NET and       initiating  the   tritium
operation of JET . Before presenting such proposals , possibly in the
frame of the next programme revision , the Commission will undertake an
in depth evaluation of the fusion programme , including the environmental
and economic aspects .
 ---pagebreak---                                                                                  - 83 -
              IMPACTE AMBIENTAL E PERSPECTIVAS ECONÓMICAS DA FUSÁO
Exposição elaborada pelos Serviços da Comissão e apoiada pelo Comité Consultivo
para o Programa de Fusão
1 . INTRODUÇÃO
    0 objectivo da investigação e do desenvolvimento da fusão europeia é a con¬
    cepção de uma central de energia capaz de satisfazer um determinado número
    de critérios de aceitação social , tais como :
    - utilizar combustiveis existentes em abundância e acessiveis à Comunidade
      Europeia ,
    - ser quimicamente limpa , na medida em que não produza dióxido de carbono
      ou substâncias tóxicas ,
    - exercer sobre o ambiente uma carga radiológica pequena quando comparada
      com as condições naturais ,
    - o seu potencial de acidente previsivel excluir calamidades capazes de pro ¬
      duzir grandes alterações â normalidade da vida das populações que habitam
      nas regiões limitrofes ,
    - ser tecnicamente fiável ,
    - ser economicamente aceitável .
    A energia de fusão tem potencialidades para se tornar numa das novas fontes
    de energia mais importantes . Não corresponderá automaticamente a todos estes
    critérios , mas é possivel encontrar alternativas de concepção da fusão por
    confinamento magnético capazes de corresponder a cada um dos critérios refe ¬
    ridos . Está ainda distante uma concepção consistente que satisfaça todos es ¬
    tes pontos , mas foram alcançados grandes progressos e está a ser desenvolvi ¬
    do um esforço persistente para a integração num projecto coerente de todos
    os aspectos desejáveis do ponto de vista do ambiente , segurança e economia .
    0 Programa Europeu de Fusão , que se ocupa de sistemas de confinamento magné ¬
    tico , prevê a passagem por três fases distintas antes da construção de cen¬
    trais comerciais de energia de fusão : a demonstração de exequibi lidade cien¬
    tifica , tecnológica e , eventua Imaente , económica . Actualmente , em conjunto
    com o J ET , os Tokamaks  médios e os seus equivalentes estrangeiros , estamos
 ---pagebreak---                                                                             - 84 -
    ainda essencialmente na fase cientifica . 0 Next European Torus ( NET ), agora
    na fase de pré-projecto , está de momento concebido como um dispositivo que
    deverá confirmar a exequibi lidade cientifica da fusão numa primeira fase ,
    e tratar do problema da exequibi lidade tecnológica numa segunda fase . Se o
    NET for bem sucedido , dever -se -á construir um reactor de demonstração ( DEMO )
    antes de ser realizada a comercialização da energia de fusão , o que não se
    prevê que aconteça antes de passarem muitos anos do próximo século .
    Por conseguinte , qualquer exposição actualizada acerca do impacte ambiental
    da fusão ( comercial ) terá de se fundamentar de preferência nos princípios
    da fusão magnética e em projectos de concepção do que em pormenores técnicos
    de projectos propostos de reactores . É muito cedo para se ser antecipadamente
    muito especifico acerca do custo da energia de fusão no próximo século .
    Durante 1986 e a pedido da Comissão , peritos europeus elaboraram um relatório
    técnico sobre o impacte ambiental e as perspectivas económicas da fusão ( Ref . 1 ).
    A partir deste último relatório e de outras fontes que actualmente constituem
    o nosso melhor conhecimento do assunto , obtiveram -se argumentos qualitativos
    que se apresentam nos pontos que se seguem .
    Podem encontrar -se outras avaliações pormenorizadas na lista de referências
    técnicas seleccionadas que actualizarão o leitor interessado relativamente
    a recentes estudos especializados .
2 . UM REACTOR DE FUSÃO CONCEPTUAL
    Na última década foram feitos vários projectos de reactores de fusão concep ¬
    tuais . Todos se baseiam no actual conhecimento da fisica dos plasmas de alta
    temperatura associado à tecnologia disponível actualmente e em desenvolvimen ¬
    tos que seja licito esperar para um futuro próximo .
    Num reactor de fusão , a energia será gerada através da conversão do deutério
    e do tritio em hélio . Ao contrário do deutério , o tritio não é proveniente
    do exterior , mas gerado no próprio reactor a partir do litio em camada fértil .
    Será pois o litio que provirá do exterior : os combustiveis primários da fusão
    deutério-trit io são o deutério e o litio .
 ---pagebreak---                                                                             - 85 -
    Grande parte da energia de fusão gerada aparecerá como neutrões rápidos que
    se tornarão mais Lentos numa camada fértil envolvente constituída por um
    composto de litio , o que levará a camada fértil a atingir temperaturas capa ¬
    zes de produzir vapor . Os neutrões fornecerão não só a fonte de calor para
    a geração de electricidade do modo convencional , como converterão parte do
    litio em tritio . Os neutrões também tornarão radiactiva a estrutura interna
    do reactor . Os niveis de radioactividade e de decadência ( meia-vida ) depen¬
    derão dos materiais escolhidos : em principio , ambos podem ser tornados
    baixos .
3 . A ABUNDÂNCIA DE COMBUSTÍVEIS DE FUSÃO
    A quantidade de combustível primário consumido para produzir um milhão de
    qui lowatt hora de electricidade numa central de fusão é de cerca de 35 gra ¬
    mas de litio convertido em tritio e de 10 gramas de deutério comparando , por
    exemplo , com as 240 toneladas de petróleo ou 360 toneladas de carvão necessá -
    rais numa central a combustível fóssil . Em compensação , para o dominio do mui ¬
    to mais complexo processo de fusão nuclear , o consumo directo de combustível
    torna -se perfeitamente secundário .
    Tanto o litio como o deutério se encontram com abundância nas águas superfi ¬
    ciais e o litio também se encontra presente em grandes quantidades em minérios ;
    apesar de não existirem dados precisos relativamente ao conjunto da Comunidade ,
    prospecções de litio em alguns paises da Comunidade indicam que os fornecimen¬
    tos internos serão suficientes e que não limitarão , de forma alguma , a utiliza ¬
    ção da energia de fusão na Europa .
4 . AUSÊNCIA DE POLUENTES QUÍMICOS
    0 produto de reacção resultante da fusão do deutério e do tritio é o hélio ,
    um gás nobre , quimicamente inactivo . Entre os processos já conhecidos ou em
    desenvolvimento para o ciclo de combustível de fusão não se encontra nenhum
    que implique emissões quimicamente tóxicas ou poluentes . Em especial , não
    se gera dióxido de carbono , nem óxidos de azoto ou de enxofre .
5 . PERIGO RADIOACTIVO REDUZIDO
    A única substância radiactiva inerente ao ciclo de combustível dos reactores
    de fusão actualmente em vista é o tritio . Os combustiveis primários , o deutério
 ---pagebreak---                                                                             - 86
   e o litio , não são radioactivos e o produto da reacção de fusão é o hélio
   não radioactivo .
   0 tritio é um isótipo radioactivo do hidrogénio . Tem uma meia-vida radioac -
   tiva de 12,3 anos e decai emitindo raios beta     ( electrões ). 0 tritio está
   sempre presente em quantidades muito pequenas ,  provenientes de fontes naturais ,
   nas camadas superiores da atmosfera . 0 tritio   gasoso oxida -se no ar e no solo ,
   formando água tritiada ( HTO ) e sob esta forma  é mais rapidamente absorvido
   pelos tecidos humanos . Contudo , a água tritiada não se concentra no corpo ,
   sendo excretada com uma meia-vida biológica de cerca de dez dias . Felizmente ,
   a água tritiada no ambiente dispersa -se e dilui -se no ecossistema muito mais
   rapidamente que os produtos de cisão e os aetinldios . Por exemplo , a meia-vida
   da perda de água tritiada das camadas superiores do solo mede -se em dias , ao
   passo que ós produtos de cisão e os actinidios podem contaminar a terra e os
   edificios durante periodos muito longos . Não existe prova ou mecanismo conhe ¬
   cido da concentração de tritio na cadeia alimentar .
   Em funcionamento normal , o tritio numa central de energia de fusão está con¬
   finado a um circuito interno que inclui a alimentação , a exaustão e a purifi ¬
   cação de combustivel , bem como á recuperação de tritio no próprio local a pai–
   tir da camada fértil de regeneração . A experiência adquirida durante a opera ¬
   ção dos reactores de cisão canadianos CANDU com concentrações comparáveis de
   tritio no refrigerante , indica que , com a actual tecnologia , é possivel manter
   as perdas para a atmosfera a niveis muito baixos . A decadência rápida do tritio
   exclui qualquer concentração acumulada de radioacti vidade de tritio a longo
   prazo .
   A radioacti vidade é induzida na estrutura do reactor pelos neutrões produzidos
   pelas reacções de fusão , mas a quantidade e a natureza dessa radioactividade
   depende do tipo de material estrutural escolhido *. A radioactividade induzida
   pelos neutrões é em grande parte imobilizada na estrutura do reactor . A peque ¬
   na fracçãoque será introduzida no refrigerante principal por meio de processos
   de corrosão , está confinada a um circuito interno fechado .
   Serão produzidos residuos radioactivos de diferentes categorias ( baixo , médio e
   alto niveis ). Os residuos com maior actividade são essencialmente uma consequência
* Por conseguinte , é possivel que o desenvolvimento com êxito de novos materiais
  de baixa activação permita uma redução substancial da radioactividade estrutu ¬
  ral , comparando , por exemplo , com o aço comercial .
 ---pagebreak---                                                                              - 87 -
    necessária da substituição de peças desgastadas dos reactores . Este residuo
    compõe -se de elementos da estrutura activada e , por este motivo , poder-se - iam
    obter grandes vantagens a partir de materiais de baixa actividade que poderiam
    ser ainda reciclados . Também existirão alguns residuos tritiados que , de acor ¬
    do com estudos recentes ( Ref . 3 ), podem ser rejeitados sem efeitos apreciáveis
    no ambiente . Não existem residuos alfa associados à fusão , tais como os acti -
    nidios de vida longa produzidos na cisão .
    Foram feitas estimativas acerca da quantidade de materiais radioactivos , tanto
    tritão como a estrutura activada , que seriam mobilizados e libertados no am ¬
    biente em possiveis situações de acidente que envolvessem igualmente uma bre ¬
    cha no contentor ; Ainda que todo o tritio libertado fosse sob a forma de água
    tritiada , parece estar ao alcance do desenvolvimento da fusão restringir o
    impacte no exterior das instalações do reactor a tal nivel que não seriam ne¬
    cessárias medidas de evacuação . Isto implica que , mesmo na possibilidade do
    mais grave acidente concebível , não se verificaria qualquer perturbação grave
    nas áreas residenciais junto da central .
6 . SEGURANÇA POTENCIAL PASSIVA
    A fusão magnética tem importantes caracteristicas implícitas de segurança que ,
    se forem convenientemente exploradas num projecto , podem ter como resultado
    uma segurança passiva alargada , se não total , do reactor . A mais importante
    destas caracteristicas de segurança consiste no facto de , seja qual for a
    fatha ou erro num reactor de fusão , não ser possivel em caso algum um " run-
    -away" ( fora de controlo ) nuclear . Além disso , a quantidade de combustivel
    presente a qualquer momento no reactor é a suficiente apenas para algumas de¬
    zenas de segundos de operação , e a interrupção do fluxo de combustivel , ou
    qualquer variação do sistema de confinamento magnético em resultado de falha
    da central , terá como consequência a rápida extinção da reacçãode fusão .
 ---pagebreak---                                                                             - 88 -
    As caracterlsticas de grande importância que contribuem para a segurança passi
    va do reactor são as seguintes :
        o calor residual relativamente baixo ( menos de 2% da energia de funciona ¬
        mento , de acordo com o material da estrutura do reactor ) de modo que , mes ¬
        mo na situação pouco provável de falha total de todos os sistemas de arre ¬
        fecimento , não ocorrerá por várias horas a fusão da estrutura , podendo mes
        mo ser totalmente evitada por uma concepção adequada ;
    -   a imobilidade da maior parte dos inventários radioactivos , os quais são
        constituídos por materiais estruturais não voláveis ;
        o baixo   potencial de perigo biológico ( radiotoxicidade ) dos radio-isótopos
        presentes que , relativamente ao aço , é cerca de 100 vezes menor que para
        os produtos de cisão e actinidios , com a perspectiva de vir a ser mais re ¬
        duzido por meio da escolha adequada de materiais estruturais ;
    - o tratamento do combustível de tritio no próprio local , o que suprime os ris
      cos associados ao transporte de tritio ( excepto , evidentemente , para o inven
      tário de tritio necessário para pôr um novo reactor em funcionamento pela
      primeira vez ).
7 . PERSPECTIVAS ECONÓMICAS DA FUSÃO
    0 desenvolvimento da energia de fusão comercial é um objectivo a longo prazo .
    A altura exacta e o alcance da exploração comercial dependerão do seu custo
    enquanto fonte de energia . Nesta fase , qualquer tentativa para prever com pre ¬
    cisão o custo da energia gerada por fusão , talvez à distância de duas gerações
    tem inevitavelmente de ser qualitativa . Os custos futuros de outros processos
    de produção de energia estão igualmente sujeitos a incertezas . Assim sendo , é
    impossível afirmar com segurança se o fusão será economicamente competitiva
    como fonte de energia na primeira metade do próximo século .
    É evidente que têm sido efectuados estudos acerca das perspectivas económicas
    da fusão ( Ref . 1 , por exemplo ). Estes sugerem niveis de custo para a produ¬
    ção de electricidade na mesma gama dos para as tecnologias de energia actuais .
    Tais niveis de custo parecem viáveis e realizáveis desde que os esforços a
    longo prazo para melhorar e simplificar as tecnologias de fusão obtenham êxito
 ---pagebreak---                                                                                       - 89 -
    Além disso , podem esperai–se vantagens ijnportantes em conjunto com o desenvol ¬
    vimento continuado da fusão , tais como as que já surgiram e que hoje podem ser
    demonstradas em áreas paralelas de alta tecnologia .
    Além disso , as perspectivas para a fusão e as comparações em termos económicos
    com outros métodos de produção de energia necessitam ser consideradas num con¬
    texto mais alargado quando os custos ligados â segurança , ao problema da auto -
    - suficiência e do impacte ambiental forem também incluidos . A fusão tem grandes
    vantagens em termos de segurança e de ambiente , e tais vantagens podem tornar -
    -se factores importantes em favor da introdução da fusão como uma nova fonte
    importante de energia para o mundo .
8 . REFERÊNCIAS
    1 . 0 impacte ambiental e perspectivas da fusão nuclear
                                               ( EUR FU BRU / XII 828 /86 )
    Outras    Referências
     2.    Environmental Aspects of Fusion Reactors
           CASINI , G. , PONTI , C -, ROCCO , P.             ( EUR- 10728 -EN, 1986 )
     3.    The Implications for Health and              the  Environment of the
           Disposal of Tritiated Wastes                      ( EUR 10617 EN , 1986 )
     4.    Fusion Reactors - Safety and Environmental Impact
           HANCOX , R. , REDPATH , W.         (Nucl . Energy 24 ( 1985 ), p. 263 )
      5.   Preliminary Findings of a U.S. National Committee on
           Environmental , Safety and Economic Aspects of Magnetic Fusion
           Energy
           H0LDREN , J.P.
            ( Paper presented at the IAEA Technical Committee Meeting on
           Fusion Reactor Safety , Culham , 3-7 November 1986 ).
      6.    Fusion Safety Status Report             ( IAEA - Tec . Doc . 388 , 1986 )
 ---pagebreak---                                                                     90 .
        DECLARATION CONCERNANT LA COMPETITIVITE ET L' EMPLOI
I. Objet de la proposition de programme
        Le programme proposé tend à poursuivre la recherche et le
        développement dans le domaine de la fusion thermonucléaire
        contrôlée et couvre toutes les activités des Etats Membres
        dans   ce   domaine . Le    but   final  de  ce  programme   est   de
        déterminer si de l' énergie peut être produite à un prix
        compétitif à partir de réactions de fusion entre noyaux
        légers , et dans ce cas , de construire en commun des prototypes
        pour     leur   production     et   commercialisation    à   échelle
        industrielle .
        Les raisons principales pour poursuivre la recherche et le
        développement dans ce domaine sur une base communautaire sont
        parmi d' autres les suivantes :
        .     l' ampleur des ressources tant humaines que financières
              nécessaires , qui suggère qu' un tel développement ne
              pourrait que très difficilement être accompli sur une
              base nationale ;
        .     la longue durée de l' effort ( s' étendant largement dans le
              siècle prochain)        nécessaire pour aboutir à            la
              construction du réacteur ;
        .     la réalisation d' un marché européen pour les industries
              européennes dans les domaines de hautes technologies et ,
              en cas de succès , l' ouverture d' un grand marché
              communautaire pour le réacteur européen .
        Si la proposition de programme ne poursuivait pas son cours ,
        il en résulterait des dommages irréversibles , dont le plus
        sévère concernerait JET . En fait , en parallèle avec la
        présente proposition de programme , est également soumise une
        proposition pour la prolongation du projet JET jusqu' à fin
        1992 . Une telle prolongation est cohérente avec l' installation
        et l' exploitation d' équipements supplémentaires sur JET , de
        manière à en assurer       le succès ultérieur .   L' absence d' une
        décision pour le programme fusion remettrait en question la
 ---pagebreak---                                                                      91 .
            date de mise en oeuvre de ces équipements et par conséquent
            rendrait impraticable la conclusion du projet à la date
            proposée : cette conclusion serait donc repoussée après 1992 ,
            ce qui entraînerait des coûts supplémentaires considérables .
II .  Avantages pour l' entreprise
            La proposition a des implications pour l' industrie européenne
            dans le domaine des hautes technologies , avec des retombées
            ( en particulier dans les domaines de la technologie des
            aimants superconducteurs , de la robotique , et des systèmes
           micro-onde de haute puissance ), au bénéfice d' autres branches
            de la science et de l' industrie .
      -     La proposition a également des implications pour les PME . Le
            rôle de l' industrie devrait augmenter lorsque le " European
           Next Step " (NET) entrera dans sa phase de projet . En
            particulier l' expérience de JET a montré que de nouvelles PME
            travaillant principalement dans le domaine de la fusion ont
            été créées ou ont connu un développement considérable suite à
            la nécessité de satisfaire      les demandes des laboratoires de
            fusion .
III . Implications du programme pour l' entreprise
            Pour la mise en oeuvre du programme , JET et les institutions
            associées au programme fusion communautaire lancent des appels
            d' offre européens pour leurs équipements et services , en
            particulier dans les domaines des hautes technologies . Les PME
            techniquement compétentes sont invitées à participer à chaque
            appel d' offre , quand c' est nécessaire .
IV .  Inconvénients possibles pour l' entreprise
      AUCUN
 ---pagebreak---                                                                      92 .
V.    Dispositions particulières en rapport avec les PME
      Il n' y a aucune disposition de cet ordre . La présente proposition
      est susceptible de stimuler les PME , comme indiqué plus haut .
VI .  Effets attendus
      -     Comme    indiqué  ci-dessus ,  les  effets   auxquels   on    peut
            s' attendre sont une stimulation dans les domaines des hautes
            technologies de la compétitivité de l' industrie européenne par
            rapport aux autres industries dans le monde .
      -     La proposition n' a pas d' effet négatif 'sur la situation de
            l' emploi dans la Communauté : au contraire , elle aide à
            accroître le savoir-faire nécessaire pour développer cette
            nouvelle source potentielle d' énergie . A long terme ,
            l' ouverture d' un grand marché européen pour le réacteur
            européen aurait un effet positif sur l' emploi .
VII . Consultations des organismes représentatifs concernés
      Les Etats Membres sont consultés par l' intermédiaire du Comité
      Consultatif pour le Programme Fusion , dont l' avis ( proposition
      1986 ) et les "vues " ( proposition révisée 1987 ) sont favorables , et
      par l' intermédiaire du Comité Scientifique et Technique , dont
      l' avis est aussi favorable . Les avis du Parlement Européen et du
      Comité Economique et Social seront aussi demandés .
 ---pagebreak---                                                   EURFU BRU/XI 1-828/86
/I
  /j-J-
     iUritNrek
       rraaiinrh
    ENVIRONMENTAL IMPACT
                      and
        ECONOMIC PROSPECTS
                        OT
                 NUCLEAR FUSION
                      ANNEXE
  BRUSSELS ,
  NOVEMBER 1986
                    C Commission ofthe European Communities
                      Directorate General XII - Fusion Programme
                      Brussels
 ---pagebreak---                      CONTENTS
                                          Page
Explanation                            ( i)-(ii )
Executive Summary                          1
Environmental Impact of Nuclear Fusion     15
Economic Prospects of Nuclear Fusion -     52
A 1986 Viewpoint
 ---pagebreak---                                                                         (i)
Explanation :
1)   By a Resolution adopted on 17 January 1985 , the Council embodied
     the Opinion of the European Parliament on a Proposal ( COM(84 ) 271
     final) from the Commission of the European Communities to the
     Council :
            " For a Council Decision adopting a research and training
            programme ( 1985-1989 ) in the field of thermonuclear Fusion"
     The European Parliament , in its aforesaid Opinion :
     (Art . 4 )   Calls again on the Commission to launch , in the next few
                  years , a public discussion on nuclear fusion and on the
                  indispensability and impact thereof ;
     (Art . 5 )   Instructs its f the E.P 's ) Committee on Energy , Research
                  and Technology , as the committee responsible , to hold a
                  wide-ranging hearing , at the time of the next programme
                  review , on the prospects for and hazards of controlled
                  nuclear fusion ;
2)   In response to the requests of the E.P. mentioned above and in view
     of the impending programme revision in 1987 the Consultative
     Committee for the Fusion Programme advised the Commission :
     " to start , without delay , the necessary actions to prepare on a
     strictly European basis , a response to the European Parliament
     concerning questions raised on the Environmental , Safety and
     Economic Aspects of Fusion" ( Extract from Minutes of CCFP 23 of 30
     Sept . 1985 ).
     Subsequently the Commission asked two groups of experts to carry
     out , during 1986 , a study on the present state of knowledge
     concerning the subjects in question .
     One group studied the Environmental aspects the other the Economic
     prospects .
3)   The work of the two Expert Groups was supervised by a Working Group
     composed of leading fusion scientists coming from the European
     fusion laboratories , from JET , from NET and from the Joint Research
     Centre .
 ---pagebreak--- The members of a Working Group were as follows :
                Messrs : BRAAMS    ( FOM , Rijhuizen )
                         BRUNELLI  ( ENEA , Frascati )
                         CASINI    ( JRC , Ispra )
                         GIBSON    ( JET )
                         GRIEGER   ( IPP , Garching )
                         HENNI ES  (KfK , Karlsruhe )
                         PEASE     ( UKAEA , Culham )
                         PREVOT    ( CEA , Cadarache )
                         TOSCHI    ( NET , Garching )
The Group met four times during the year in order to advise the
experts on the issues raised in their reports .
The final outcome is the Report which follows and which consists of
three parts , an Executive Summary prepared by the Services of the
Commission and two Technical sections prepared by the Expert Groups
concerned .
 ---pagebreak---    ENVIRONMENTAL IMPACT AND ECONOMIC PROSPECTS OF FUSION
                    AND EXECUTIVE SUMMARY
CONTENTS
1.   Introduction                                    2
2.   The Route Towards a Fusion Reactor              3
3.   A Conceptual Fusion Reactor                     4
4.   Environmental Impact During Norma ] Operation   7
5.   Environmental Impact due to Accidents           9
6.   Safety Aspects                                  9
7.   The Economie Prospects                          11
8.   Conclusions                                     13
 ---pagebreak---                                                                   2.
      ENVIRONMENTAL IMPACT AND ECONOMIC PROSPECTS OF FUSION :
                       AN EXECUTIVE SUMMARY
INTRODUCTION
The aim of European fusion research and development is to produce a
design of a power plant that satisfies a number of social
acceptance criteria such as :
     it is economically acceptable
     it is technically reliable
-    it is chemically clean , in that it produces no carbon
     dioxide or toxic emissions
-    its radiological burden to the environment , either from the
     plant or from waste products , in normal conditions is small
     compared to the natural background
     its credible accident potential excludes calamities disrupting
     normal life in the community outside the reactor site boundary
-    it relies on fuels and construction materials that are
     abundant and accessible to all countries of Europe .
Fusion energy , when available , will not automatically fulfil all
the above criteria . It is , in fact , possible to conceive of
applications that violate one or more of these . However , this
report will show that design options for magnetic confinement
fusion are being put forward to meet each one of them . This is not
to say that a consistent design along these lines is in hand .
Although great progress has been achieved that brings us close to
fusion conditions , it remains a formidable challenge to the science
and technology of our time to integrate all desirable
environmental , safety and economic features into a coherent design .
All this applies to the deuterium-tritium fusion system . There is a
long-term prospective that this may eventually be superseded by
so-called advanced fuels ,    but the case is made that deuterium-
tritium fusion is a worthy goal to pursue on its own merits .
Clearly , our acceptance criteria must be further refined and
quantified before they reach the level of precision that will
ultimately be required when decisions to enter the commercial stage
of fusion power are to be made . In this context , a report such as
 ---pagebreak---                                                                       3.
   this can serve a multiple purpose . First , to remind workers in the
   field of the stringent standards society is likely to apply to the
   outcome of their work and to focus their attention on all questions
   raised in this context .
   Secondly , to reassure both the responsible authorities and the
   general public that the efforts devoted to the subject are striving
   for the highest standards , and that encouraging progress is being
   made towards providing society with a supply capable of filling a
   sizeable , indeed the major , portion of its long-term energy needs
   in the best possible way . Finally , the report is likely to provoke
   reactions that contribute to a better understanding of the promise
   held by fusion and of the constraints to be imposed on this
   emerging technology if and when it comes to widespread application .
   This report summarises , with a minimum of technical detail , two
   technical reports by teams of specialists drawn from several
   European research institues : "Environmental Impact of Nuclear
   Fusion" and "The Economic Prospects of Nuclear Fusion : A 1986
   Viewpoint".
2. THE ROUTE TOWARDS A EUROPEAN FUSION REACTOR
   The European fusion programme , which concentrates on magnetic
   confinement systems , envisages three distinct steps to be taken
   before commercial fusion power stations can be built .
   The first is to establish the scientific feasibility of the process
   and this is the main thrust of the present programme with the JET
   Joint Undertaking at Culham , UK , as the principal experimental
   apparatus and with complementary studies in the national
   laboratories . The next step , NET (Next European Torus ), will be to
   establish the technological and engineering feasibility . The NET
   design team has already been established at Garching , Federal
   Republic of Germany , and is currently in the pre-design phase of
   the Project . The construction of NET will depend on the main
   experimental results of JET (Joint European Torus) and other fusion
   experiments . After the successful operation of NET , a demonstration
   reactor - DEMO - will be required to establish the design features
   that will determine the economic feasibility of a fusion reactor .
 ---pagebreak---                                                                      4.
   The timescale for such a programme is long but if all stages
   proceed to plan a commercial fusion power station could be in
   operation in the first half of the next century , a time when ,
   according to current predictions , new sources of pollution-free
   energy will be required to supplement nuclear fission and other
   energy sources . In addition , the dwindling supplies of the fossil
   fuels , coal , gas and oil will be needed increasingly for other
   industrial purposes .
   JET , one of the world 's leading fusion experiments of the tokamak
   class , aims at achieving conditions approaching those required in a
   reactor . To do this , the fuel , which is a mixture of deuterium and
   tritium ( the heavy isotopes of hydrogen ) gas , must be heated to
   temperatures in excess of 100 million degrees Celsius and held in
   isolation from container walls by magnetic fields . These fields
   provide the necessary thermal insulation to prevent excessive
   cooling of the hot ionised gas known as plasma . The plasma in JET
   is contained in a large ring-shaped vacuum vessel called a torus .
   If the plasma physics revealed in the JET experiments is favourable
   then   the power which would be       released from  fusion reactions
   occurring in the JET plasma could be several tens of megawatts for
   a few seconds . NET , an experimental test reactor producing a
   thermal fusion power of about 600 MW , is being designed to
   demonstrate sustained reactions , (which themselves should continue
   to keep the plasma hot ) , and to provide the necessary technological
   data for designing a demonstration reactor ( DEMO ) with a net
   electrical output of several hundred megawatts .
3. A CONCEPTUAL FUSION REACTOR
   A number of conceptual fusion reactor designs have been made over
   the last decade . They are based on the present knowledge of the
   physics of high temperature plasmas together with the technology
   currently available    or of developments that can reasonably be
   expected in the near    future . Based on plausible extrapolations to
   the reactor level , a  reactor of net electric power of 1200 MW has
   been defined for the   purpose of the attached technical reports and
   been used in the environmental and economic comparisons .
   The simplest view of a fusion reactor is a unit into which the
   basic fuels - deuterium and lithium - are fed and the output is
 ---pagebreak---                                                                      5.
    electricity with helium as the principal waste product .
    Lithium is required to produce tritium (a radioactive form of
    hydrogen) which will be subsequently "burnt " with deuterium to
    produce power from fusion reactions . Deuterium from water and the
    light metal lithium from the earth 's crust are both plentiful and
    geographically well distributed . Less than one tonne of these fuels
    would be consumed in a 1200 MW fusion power station per year . Most
    of the fusion power generated will appear as high speed particles
    called neutrons i which will be slowed down in a surrounding blanket
    made of a compound of lithium causing the blanket to heat up to
    temperatures suitable for raising steam . The neutrons not only
    provide the heat source for generating electricity in the
    conventional way ,    but  also convert  some of  the  lithium  into
    tritium . The neutrons also cause the reactor internal structure to
    become radioactive . The level of radioactivity and the decay rate
    (half-life) will depend on the structural materials chosen ; both
    could in principle be made low .
3.1 Radioactivity in a Fusion Reactor
    The only radioactive substance inherent to the fuel cycle of the
    currently-envisaged fusion reactor is tritium . In addition ,
    radioactivity is induced in the structure of the reactor by the
    neutrons arising from the fusion reactions . These two sources of
    radioactivity have been considered in assessing the safety and
    environmental aspects of fusion reactors in the following sections .
3.2 Tritium
    Tritium is a radioactive Isotope of hydrogen . It has a radioactive
    half-life of 12.3 years and decays by emitting beta-radiation
    ( electrons ) . Tritium is present in very small quantities at all
    times from natural sources in the upper atmosphere . Man-made
    tritium , mainly from thermonuclear weapons testing programmes , far
    exceeds the natural background levels of tritium . Gaseous tritium
    oxidises in air and in the soil to form tritiated water (HTO ) and
    in this form it is more readily absorbed by human tissue . However ,
    tritiated water does not concentrate in the body but is excreted
    with a biological half-life of about ten days . Fortunately ,
 ---pagebreak---                                                                        6.
      tritiated water in the environment disperses and dilutes in the
      ecosystem much faster than fission products and actinides . For
      example , the half life of the loss of tritiated water from the
      upper layers of the soil is measured in days , whereas fission
      products and actinides can contaminate land and buildings for very
      long periods . There is no evidence or known mechanism for the
      concentration of tritium in the food chain .
3.3 . Tritium Inventories
      The amount of tritium in the plasma of the reactor at any given
      time is very small - less than 1 g . The total tritium inventory for
      a 1200 MW plant will be about 3 kg of which about one third will be
      kept in a number of separated bunkered store rooms until required .
      The stored tritium need not be in the gaseous form but may be kept
      in a solid stable form such as a metallic tritide . There will also
      be tritium trapped in the lithium blanket surrounding the reactor
      and in the processing plant ; the quantity of tritium therein will
      depend upon the reactor design ranging from a few hundred grams to
      about 2 kg . The bulk of the tritium in a reactor - in store and in
      the blanket - is effectively immobilised and has a very low chance
      of escaping into the environment . Present knowledge , however ,
      indicates that the quantity of tritium that could be released in
      any conceivable accident could be reduced to about 200 g and this
      value  has   therefore  been  assumed   in   the assessment  of   the
      environmental consequences of the worst conceivable accident .
3.4   Radioactivity of the Internal Structure of the Reactor
      The neutrons resulting from the fusion reactions will make the
      structural materials of the reactor radioactive , but the level and
      longevity of the radioactivity depends essentially on the chemical
      composition of the elements used in the manufacture . The components
      closest to the plasma - particularly the torus wall and the blanket
      structure - will be subject to the most intense neutron bombardment
      and if made , for example , from conventional stainless steel will
      become the major fraction ( over 90% ) of the radioactive inventory
      of the plant . Although the total radioactive inventory of a fusion
      reactor at the time of shut down using conventional stainless
      steels for the torus wall and other internal structures will be
 ---pagebreak---                                                                                 7.
      almost comparable to that of a fission plant of similar power the
      biological    hazards    ( radiotoxicity )     associated     with     steel
      activation products are significantly lower ( about one hundred
      times lower ) than those of fission products and actinides .
      Furthermore , the bulk of the activation products are trapped in the
      solid  structural material    of   the  reactor and    cannot   as  such be
      dispersed into the atmosphere .
      In making any safety and environmental assessments of fusion
      reactors , it is necessary to consider potential hazards specific to
      fusion that could arise especially from the radioactive tritium and
      from the activated reactor structure .      Studies have therefore been
      made on the environmental impact during normal operation ,               the
      radioactive waste generated during the life of the reactor , and the
      environmental impact due to the worst possible accidents . These are
      reported in depth in the accompanying reports together with the
      assessement of the economics of a fusion reactor . A summary of each
      of these aspects is given in the following sections .
4.    ENVIRONMENTAL IMPACT DURING NORMAL OPERATION
4.1   Routine Emissions
      The only gaseous part of the radioactive inventory of the
      currently-envisaged    fusion      reactor    will    be    the    tritium .
      Multiple-containment systems will be used with the steel-lined ,
      air-tight reactor building being the final barrier against the
      release of tritium into the environment . The largest internal loss
      of tritium during normal operation may occur via the coolant lines .
      This is because tritium can permeate into the cooling channels of
      the blanket . Operating experience gained from Canadian CANDU
      fission reactors , with comparable tritium concentrations in the
      coolant , indicates that , with existing technology , losses to the
      atmosphere can be kept to very low levels . On the basis of this
      experience , the total tritium released daily from a 1200 MW reactor
                                                          *
      is expected to be less than 1 / 100 g (3.7 TBq )       which would result
      in maximum dose to the most exposed individual of the public local
                                       *
      to the plant of about 10      Sv    (1 mrem) per year . This is well
 Bq = Becquerel ; TBq = 1,000,000,000,000 Bq ;
  Sv = Sievert ; mSv * Milli-Sievert ;     Sv = Micro Sievert
 ---pagebreak---                                                                       8.
    below the limit Imposed by current regulations for fission reactors
    ( 50-300    Sv or 5-30 mrem per year ) and would , for this most
    exposed person , Increase the dose burden above that due to average
    natural background radiation by about 1% - much less than the
    variations in background radiation from place to place .
    The most likely release of activation products during normal
    operation is from the leakage of corrosion products from the
    primary cooling circuits or from a loss of cooling water during
    maintenance . Based on fission reactor experience , at most this
    would amount to a relatively small amount per year and the
    consequences to any member of the public would be negligible .
4.2 Radioactive Wastes
    The principal radioactive components of a fusion reactor will be
    the torus wall and the blanket structure , both of which will have
    become activated by the fusion neutrons . If conventional steels are
    employed , it is likely that these components will be replaced about
    four times during the life of the reactor . Low level wastes will
    also arise from various processing systems around the reactor .
    Experimental facilities , such as JET , use conventional types of
    stainless steel for the construction of the torus ; these steels are
    not ideal materials for a fusion reactor . The fusion technology
    programme is therefore investigating new materials , in which the
    alloying elements that become radioactive with long half-lives are
    replaced by elements with only short-lived radioactivity . These
    materials could reduce the radioactive inventory of the structure
    by a factor between 10 and 100 , the decay rates would be faster and
    recycling of many of these selected materials could be considered
    after about 100 years . The storage problems for such wastes would
    not only be for much shorter duration than waste from fission
    reactors (where the long-lived actinides are inherent to the
    process ) but would also be much easier to handle . The fusion waste
    would be in solid form and , having a large surface area , active
    cooling would not be necessary and furthermore deep geological
    disposal would not be required .
 ---pagebreak---                                                                      9.
   In general , it is concluded that the radioactive wastes from the
   fusion process will be considerably easier to store and dispose of
   than the wastes from fission reactors .
5. ENVIRONMENTAL IMPACT DUE TO ACCIDENTS
   Studies are being made of accident scenarios resulting from major
   technical failures of the reactor or plant . If such a severe
   accident caused the reactor building to be breached ( although this
   seems impossible ) then the radioactive release into the environment
   would be mainly tritium and some activated structural materials .
   No mechanism has been identified that could mobilise more than a
   few grams of radioactive particles from the reactor structural
   materials .
   The maximum quantity of tritium contained inside several different
   buildings of the fusion plant is considered to be about 3 kg . No
   sequence of events leading to the release of all this tritium could
   be found and the most severe accident identified would lead to the
   release into the environment of not more than 200 g of tritium. If
   this 200 g of tritium in the most hazardous form (HTO ) were
   released from the building roof ( rather than from a high chimney
   stack ) under adverse weather conditions it would cause a maximum
   dose of 60-80 mSv (6 to 8 rems ) at a distance of 1 km from the
   plant . In such an incident , the levels of radiation would not cause
   direct harm to any member of the public or lead to the evacuation
   of the public outside the power station boundary fence .
   It is concluded , therefore , that releases of tritium - the most
   hazardous material in a fusion reactor - and radioactive internal
   structural materials will cause no immediate harm to an individual
   or cause disruption to the normal life of the community outside the
   power station boundary fence during normal operation , during
   maintenance operations or even following a major accident or plant
   failure .
6. SAFETY ASPECTS
   Fusion reactors will be complex nuclear installations but
   nevertheless appear to have a number of intrinsic safety features .
 ---pagebreak---                                                                  10 .
The most important safety aspect is that whatever fails or goes
wrong with a fusion reactor , it cannot in any circumstance lead to
an uncontrolled , self-started and self-sustained nuclear runaway .
Moreover , the amount of fuel in the reactor core at any given time
is only sufficient for a few tens of seconds of operation and the
interruption of the flow of fuel , or a variation in the magnetic
confinement system because of a failure of the plant , will lead to
the instantaneous quenching of the plasma and the fusion reaction
will cease .
In the event of the shut-down of the reactor , cooling systems must
continue to operate to cope with the afterheat in the torus wall
and the blanket structure . In a fusion reactor , the afterheat will
be relatively low ( up to 2% of the operating power depending on the
structural materials of the reactor ) . Even in the unlikely
situation of the total failure of all the cooling systems , the low
level of afterheat and the large volume and surface area of the
structures are such that melting of the structures would not occur
for several hours or even may be avoided altogether by appropriate
design .
Safety for any nuclear reactor is of the utmost importance . A
fusion reactor will have a number of specific safety features built
in . The tritium plant will be built with multiple-containment
systems and the bulk of the tritium will be stored in a solid
immobile form and in separate bunkers away from the reactor to
minimise leakage to the environment . The tritium reprocessing will ,
in general , be carried out on site as an integral part of the
plant . There may be some transportation of tritium in immobilised
form outside the plant to start up new reactors . The reactor
building itself will be designed such that under all conceivable
internal accident conditions the building would not be breached .
Virtually all the radioactive inventory of a fusion reactor is
non-volatile structural materials and there are prospects that
long-lived radioactive materials can be avoided . The biological
hazard potential of the radio-isotopes from fusion reactors is low .
Even in the worst conceivable accident scenario ,    there seems no
circumstance resulting in immediate harm to an individual beyond
the site boundary or the evacuation of the public .
 ---pagebreak---                                                                        11 .
   It is concluded therefore that fusion reactors will provide a safe ,
   environmentally-acceptable future source of energy .
7. THE ECONOMIC PROSPECTS
   For fusion power to be established as a commercial source of
   energy , it is necessary for it to be economically competitive , to
   satisfy existing safety requirements and to be acceptable to the
   public . Just as it is not easy to predict the price of oil next
   year , to predict some fifty years ahead whether an as-yet unproven
   system will be competitive is difficult and uncertain , and by
   necessity , will be based on a number of assumptions . The emphasis
   of the current research programme has been directed to making the
   fusion process work in large-scale experimental apparatus . In
   parallel with these studies of the physics of plasma , several
   conceptual design studies of fusion reactors have been carried out
   to identify the general trends for future technological
   developments . The majority of these studies have concentrated on
   tokamak reactors (reflecting the emphasis of the fusion research
   programme ) although some alternative systems have been included .
   These studies have produced preliminary estimates of both the
   construction cost of a fusion plant and the cost of generating
   electricity . As part of the NET study , for example , cost methods
   suitable   for a f irst -of -a -kind tokamak fusion reactor have been
   evolved . From these , it appears that if a prototype commercial
   reactor of 1200 MW electrical output ( sent out ) were built solely
   based on the present knowledge of plasma physics and technology ,
   the generating cost of electricity would be 2-3 times that
   generated by today 's thermal fission and coal stations . This is , of
   course , taking a very pessimistic case for fusion and comparing it
   with a well-established reactor design . Series production is
   expected to reduce this gap significantly or even close it . It
   should be noted that the present generating cost of electricity
   from a fast breeder reactor ( also first of its kind ) is twice that
   from conventional thermal fission reactors . As the development of
   fusion power proceeds , it is reasonable to expect considerable
   improvement and simplifications in both the technology and the
   physics of plasmas which will lead to a reduction in the generating
   costs . For example , the cost of the superconducting magnets
   required for a fusion reactor are very high due principally to the
   present very limited market for superconducting materials but their
 ---pagebreak---                                                                    12 .
cost is expected to drop as their applications increase . Also , the
costs of the blanket and cooling systems , and the reactor building
itself , are likely to fall in series production as operational
experience leads to simpler designs . A dramatic cost reduction
could also be made with improved plasma operations . If the beta
value - a measure of the efficiency of the magnetic field in
confining plasma - were increased by a factor of 3 from its
presently achieved values , then the generating cost of electricity
would be reduced by about 30% without taking account of increasing
power advantage so gained .
There are many examples where the economics of high technology
systems have been drastically improved from the f irst -of-a- kind
version . Therefore , the demonstration of scientific and technical
feasibility must be followed by physics and engineering
improvements together with simplifications of the overall system to
arrive at an economically-competitive power plant .
In contrast to the extensive literature containing fusion reactor
design studies with detailed cost estimates , there have been
several    publications  which   argue that   fusion  will  never  be
economic .  The main criticisms are that   fusion devices have a  low
power density , a long payback time and are too complex . It can be
seen that the use of power-density -based comparisons is not
reasonable by examining fission reactors themselves where typical
                                                     -3
power densities are between 15 and 0.4 MW(th) /m , whereas the
construction and generation cost differences are within a factor of
two . The energy payback time is made by comparing the total energy
expended in all processes involved in the manufacture , construction
and operation of the plant compared with the total energy generated
during the working life of the reactor . For a fusion reactor , the
energy expended on the construction of the reactor is about twice
that for an equivalent fission plant , but when the energy of
manufacturing and processing of the fuel is taken into acount , then
the energy expended on fusion is si^nif icantly less than that for
the equivalent fission system . With regard to complexity , this
cannot yet be quantified , but by an analogy with aircraft , for
example , the increased complexity has not lead to a decrease in
reliability .                  '
In summary , therefore , the information presented by the critics of
 ---pagebreak---                                                                       13 .
   fusion is often highly selective , and the conclusions are not
   supported by the detailed studies . It is true that the low power
   density of many present designs leads to high capital costs , but
   the estimated cost of electricity from fusion power stations is not
   much greater than forecast costs from existing or other alternative
   energy sources .
   Several studies have attempted to calculate the generating cost of
   electricity from fusion in the mid twenty-first century and to
   compare this with the expected cost of electricity generated by
   coal , thermal fission , and solar photovoltaic cells . Despite fusion
   power having a high capital cost , the overall generating cost of
   electricity from a fusion power station is within the wide range of
   costs expected from existing or other alternative energy sources .
   Fusion can therefore not be dismissed purely on economic grounds .
   Indeed , it is reasonable to expect that nuclear fusion will emerge
   as one of the competing systems for the large-scale production of
   electricity in the middle of the twenty-first century .
8. CONCLUSIONS
   The two appended reports have evaluated the environmental , economic
   and safety aspects of fusion in considerable detail . They show that
   if the scientific feasibility can be demonstrated , then even
   without significant development , fusion would provide a safe power
   source with a very small environmental impact on the public during
   normal operation or even following a major reactor accident . There
   are also good prospects that the cost of fusion power , assuming
   reasonable technical developments and some improvements in the
   confinement of high temperature plasma , will be within the range
   expected from other large-scale energy sources in the middle of the
   next century . In addition , there are other potentially beneficial
   aspects of fusion power . These include the security of fuel
   availability - deuterium and lithium are spread widely - and the
   low price of fuel . As the tritium cycle is integral with the power
   plant , the fuel supply will not depend on external reprocessing
   systems . The handling and storage of the radioactive structure of a
   fusion reactor will create no new problems but the possibility of
   avoiding the need for long-term storage of radioactive waste by
 ---pagebreak---                                                                    14 .
developing suitable low activation materials is likely to be a
major advantage from a public acceptance viewpoint in many
countries . In addition , there would be no significant atmosphere
pollution from a fusion reactor , as is also the case with fission .
There is a range of possible long-term developments which would
result in an even more attractive reactor system . The reports
concentrated on the deuterium-tritium fusion system , but in the
longer term , other reactions involving deuterium alone , or
deuterium and helium-3 , could be considered . The benefit for such
reactions would be a considerably smaller radioactive inventory and
a very substantial simplification of the reactor , since the need
for breeding tritium would be eliminated . These reactions , however ,
require more    sringent  plasma  conditions  than   those yet  to be
established for the deuterium-tritium reaction .
The first concern must therefore be to build on the very good
progress   made  on   demonstrating  the  scientific   feasibility of
deuterium-tritium fusion and to establish the foundation required
to enable the NET programme to proceed . If NET and later DEMO
proceed satisfactorily and at the envisaged timescale , then a first
commercial fusion power station could be in operation towards the
middle of the next century . The high standard of living enjoyed by
industrialised countries owes much to the availability of cheap
energy for both domestic and industrial purposes . New sources of
energy will be needed as reserves of some fossil fuels are
diminished . The vast and well-distributed reserves of fuel and the
inherent safety of fusion reactors , together with the envisaged
environmental advantages and economic competitiveness make fusion a
desirable objective as a major source of safe energy for future
generations .
 ---pagebreak---                                                                             15 .
                           ENVIRONMENTAL IMPACT OF NUCLEAR FUSION
W    Gulden       The NET Team , Max-Planck Institut fur Plasmaphysik ,
                  D-8046 Garching bei München , FRG .
H. Klippel        Energy Research Foundation , NL-1755 ZG Petten ,
                  The Netherlands
P.   Rocco        Joint Research Centre , I 21027 Ispra ( Varese ), Italy .
J . L. Rouyer     IPSN / DPT / STEP , CEN de Saclay , B P. No . 2 ,
                  F -- 91 190 Gi f - sur - Yvette , France
G.   Kessler      Kernforschungszentrum Karlsruhe , INR , D-7500 Karlsruhe 1 , FRG .
                                               CONTENTS
                                                                    Page
0 . SUMMARY                                                          17
1 .  INTRODUCTION                                                   21
2 . FEATURES OF A TYPICAL FUSION POWER PLANT                        22
3 - ENVIRONMENTAL IMPACT OF A FUSION POWER PLANT                    29
4 . DEVELOPMENT POTENTIAL                                           46
5 . CONCLUSIONS                                                     47
6 . REFERENCES                                                      48
7 . GLOSSARY                                                        50
 ---pagebreak---                                                                                    16 .
ACKNOWLKDGEMKNTG
        The authors are very grateful for the comments and suggestions of
Drs. C.M. Braams ( FOM ), B. Brunelli ( ENEA ), G. Casini ( JRC , Ispra ), J. Darvas
( CEC ),   A. Gibson ( JET ),  G.  Grieger  ( IPP ). R. Hancox ( UKAEA ), H.H. Hennies
( KfK ) , A. Malein ( CEC ), D. Palumbo ( CEC ), R.S. Pease ( UKAEA ), F. Prevot ( CEA ),
J. Raeder ( NET ) and R. Toschi ( NET ).
 ---pagebreak---                                                                                        17 .
0.     SUMMARY
0.1     Inherent safety features
           A fusion power plant can be designed for inherent safety such that
effects of all credible accidental circumstances on the environment will be
kept small by generic safety features : neither the externally supplied fuels
( deuterium and lithium ) nor the ultimate fusion reaction products ( helium ) are
radioactive or         toxic ,  there   is   a   small   fuel   inventory   in the plasma ,  an
uncontrolled ,      self-started and self sustained nuclear runaway is impossible ,
the power density in the first waLl and blanket structure is relatively low ,
afterheat at shutdown is moderate ,              the bulk of radioactive material is non ¬
volatile      structural     material ,    and    the  radio - isotopes  have   low  biological
hazard potential .
0.2     Basis for assessment of environmental impact
         Based on plausible extrapolation from todays physics and technology to
reactor level ,        a FCTR ( First Commercial-sized Tokamak Reactor ) was defined .
This     FCTR    ( 1200   MWe )  is   used     as   a  basis    for  the   assessment   of  the
environmental impact of Tokamak reactors .
0.3     Environmental impact during normal operation
         The levels of radioactive effluents in normal operation will match the
regulations in Europe and elsewhere and hence these effluents will not be a
hazard to the public . It is worth noting that the technical potential exists
for further reducing the emission to virtually insignificant levels .
Re 1 e ase of radioactivity during nor mal ope r a tlon
           The principal sources of airborne radioactive effluents will be the
release of tritium from buildings , the corroded activation products that leak
through coolant loops ( forming aerosols ), the activation of the cover gas or
air inside the reactor building and gases released in auxiliary buildings
during radioactive waste management operations .                Assuming adequate containment
measures ,      the     annual    atmospheric       releases    from   normal   operation   and
maintenance procedures could be limited to about 2 g (= 7^0 TBq = 20000 Ci ) of
tritium and 18.5 GBq ( 0.5 Ci ) of activation products .
 ---pagebreak---                                                                                18 .
          Aquatic radioactive releases will be mainly due to losses during
maintenance of water cooling systems and from processing of operational waste .
Annual effluents consist of about 0.15 g (= 55.5 TBq = 1500 Ci ) of tritium and
185 GBq (5 Ci ) of activation products .
       The release values given have been obtained with moderate extrapolation
of present technological capabilities and can be considered as reasonably
conservative .
Radiation doses due to the release o f radioactivity during normal operation
      The above described radioactive release of tritium amounts to a total of
a few TBq / d ( about 800 TBq/ a ) from the fusion plant . This release will result
in a maximum dose of the order of 0.015 mSv / a ( 1.5 mrem / a ) to the most exposed
individual of the public ( stationed permanently downwind at the boundary of
the plant , eating food and drinking water gained at this place ). This is well
below the limit imposed by regulations ( 0.05 to 0.3 mSv / a = 5 to 30 mrem/ a )
and is about 1 % of the average dose burden by natural background irradiation .
Environmental impact of non-radioactive e ffluent s
        Fusion plants do not emit CO^, nitrous oxide , or any other biotoxic
chemicals .   The generation of waste heat is the same as in any other type of
steam raising plant .
0.4  Environmental impact due to accidents
       The analysis of accident scenarios following major technical failures
leads to the conclusion that the radioactive effluents ( mainly tritium ) in
such cases would have a very low impact on the lives and the health of the
surrounding population .
Release of radioactivity under accidental conditions
     The most severe hypothetical accident would lead only to a release to the
environment of about 200 g of tritium .
         Essentially no mechanism was found that could mobilize significant
fractions of structural materials .           The worst    hypothetical  release    of
radioactive particles is a few grams .
 ---pagebreak---                                                                                 19 .
Radiation doses djae _to j^eT^ease of rad i oact i vity under accidenta l cond itionr,
        The hypothetical release of 200 g tritium in the most hazardous form of
HTO from the building roof , although building breaching appears not to be
possible , would cause a maximum dose of 0.06 to 0.08 Sv (6 to 8 rem ) at 1 km
distance , under worst weather conditions and dry deposition .          These values are
within the limit of 0.05 to 0.15 Sv (5 to 15 rem )            accepted by the licensing
authorities for abnormal events of low probability .
0.5   Waste
            The   radioactive  waste  generated     by   fusion  power   plants  will   be
quantitatively comparable to fission reactors , but qualitatively it will be
much less of a potential hazard .
       It is likely that the high level waste from FCTR , mainly first wall ( AISI
316 ) disposals , can be handled like spent fission fuel elements . The amount
of   first   wall   waste is  of  the same   order    but  the  hazards  are  much   lower
compared to spent fission fuel .          Structural materials from spent breeder
blanket segments will have a high volume for disposal if the segments are
replaced frequently ,     but there   is a good potential for material re-use or
easier management when alternative structural materials have been developed .
      The quantity and disposal strategy of low level waste generated annually
from normal operation of FCTR are comparable to that of fission reactors ,
providing that care is bestowed on detritiation and tritium immobilisation .
0.6   Low activation materials
        The presently used austenitic and martensitic steels do not meet fusion
wastes    long term requirements .      Low activation materials under development
could avoid the needs of long term isolation and deep geological disposal .
Even recycling and re-use might be possible after some decades .
0.7    Direct radiation , magnetic fields , radiofrequency radiation
         No difficulties are expected in conforming to existing guidelines for
long term exposure to magnetic fields ,         radiofrequency radiation and direct
radiation ( e.g . by neutrons ).
 ---pagebreak---                                                                            20 .
0.8  Impact o n the public , short and l ong t erm aspects
          All environmental aspects of fusion are presently good ;      the main
advantages to be emphasized are the low risks induced by severe accidents and
the non existence of important long term (> 100 a ) potential hazards .
0.9  Development potent ial
       The good situation for fusion can even be improved by developing the
potentials for further limiting the wastes and the tritium inventory .
 ---pagebreak---                                                                              21 .
1 .  INTRODUCTION
        The final goal of developing fusion power plants is the production of
electric energy in a safe and economic manner and with little short and long
term impact on the environment .
       Present designs which can only be based on todays physics and technology
have to be considered as a first step only .       This holds for both the type of
reactor and the materials used .      However ,  even based on todays technology ,
fusion power plant designs indicate         compared to e.g. coal , oil , fission
power plants - advantages with respect to environmental impact :
    Once    the  ignition  conditions  are  reached ,   the   fuel  is continuously
    introduced    in  the plasma  chamber  at   the   rate  needed  to  sustain   the
    reaction .  When the fuel flow is interrrupted , the reaction stops .
    An uncontrolled , self started and self-sustained nuclear power runaway is
    impossible as a change of operating conditions will lead to instability of
    the plasma and subsequently end the burn process .
    The fuel content in the plasma is small ( about 1 gram ).
    In general all operations on fuel cycle are within the plant itself .
    No emission of CO^, S02 or N0x -
    Development potentials still exist for fusion in the near future , e.g. by
    the use of low activation materials .
         The material presented in the following chapters pertains to tokamak
reactors based on todays technology .       It mainly emerged from the European
Fusion Programme whose focus is the design and construction of NET ( Next
European Torus ). This fusion device will be an experimental reactor with a
thermal power of about 600 MW and has to provide the major part of the
knowledge necessary for designing a demonstration reactor ( DEMO ).
      A " First Commercial-sized Tokamak Reactor " ( FCTR ) has been defined as the
basis for the results and comparisons contained in the following chapters .
This has been done by using plausible extrapolations from todays conceptual
designs to the reactor level ( about 1200 MW e ).
 ---pagebreak---                                                                                             22 .
2.  F !•: A TUH KS OF A TYPICAL FUSI ON POWER J ' LAN T
2.1   Definition of a tokamak power plant
          Extrapolation from present conceptual experimental tokamak devices such
as  INTOR       /1 / and NET    /2/   to fusion power        plants can be performed with
different degrees of conservatism .            Table 1 displays some typical parameters .
       The INTOR and NET parameters reflect a prudent interpretation of present
day physics and technology .          FCTR / 3 / ( First Commercial-sized Tokamak Reactor )
is  a      reasonable     extrapolation   of    todays   conceptual    design    parameters  to
reactor level .        STARFIRE / 4 / - a US conceptual reactor design - contains many
advanced assumptions and design characteristics .
TABLE 1 : Typical fusion device parameters
                                             INTOR        NET-DN     FCTR         STARFIRE
Fusion power ( MW )                          585          600         3590        3510
Electrical power ( net , MW )                    0        0           1200        1200
Toroidal field on plasma axis ( T )          5.5          5.0           5.7        5.8
Plasma current ( MA )                        8.0         10.8         18.0        10.1
                                  2
Neutron wall loading ( MW/ m )                1.3         1 .0          1 .8       3.6
                         2
First wall area (m )                         352          480         1600         780
The following assessment of the radioactive inventory and environmental impact
of Tokamak reactor designs - as will be discussed in the subsequent sections -
will make reference mainly to FCTR because it is considered to be the most
representative        reactor   concept   in    Europe   in   terms of    todays   physics and
technological capabilities .
 ---pagebreak---                                                                                23 .
2.2 Inhérent Safety
         A FCTR will have some generic safety features which suggest that the
effects on the environment will be small .    These are :
    - an uncontrolled , self-started and self-sustained nuclear power runaway
        is impossible ,
    - low fuel inventory in the plasma chamber ,
    - relatively low power density in first wall and blanket structure ,
    - moderate afterheat at shut-down ( up to 2% of operating power in the
                                                                       i
       first wall and blanket structure ) diluted on a large surface .
    - the bulk of the radioactive material is non-volatile structural
      material ,
    - relatively low biological hazard potential of the radio-isotopes .
       In addition it seems to be possible to design a containment such that it
will not lose integrity under all conceivable internal and external accident
conditions .
2.3   Multiple containment concept
       The most volatile part of the radioactive inventory of FCTR is tritium .
Therefore the safe containment of tritium inside the fusion plant for both
normal    operation   and  accidental conditions  will    become mandatory .    This
requires a multiple-containment concept ( in general triple ), to minimize the
release of tritium to the environment .
2.H   Radioactive inventories
2.4.1 Tritium inventory
General remarks
        For the first application ( D-T cycle ) fusion reactors , tritium will be
used as fuel , the D-T reaction products being stable He4 nuclei and high
energy neutrons .      The tritium inventory in the plasma chamber will be very
small (1 gram ).    The total tritium inventory in a plant , however , will be some
 ---pagebreak---                                                                                   24 .
kilograms , distributed in the storage , the process systems and the reactor
structures .    The bulk of the tritium will be stored in a solid immobile form
and in separate bunkers away from the reactor .
       Tritium is of moderate radiotoxicity , with a half life of 12.3 years .          It
emits (5-radiation with a maximum energy of 18 keV .              The radiotoxicity of
tritium strongly depends on its chemical form : gaseous tritium ( T2 , HT ) is
about 25000 times less dangerous compared to the oxide ( HTO ). Gaseous tritium
partly combines with oxygen in the air to HTO or is being oxidized to HTO by
bacteria in the soil .         In HTO form it is more readily absorbed by human
tissue .    However , tritiated water does not concentrate in the body but is
excreted    with   a   half  life  of   about   ten   days .   Tritiated water    in   the
environment disperses through the ecosystem much faster than fission products
and actinides .     For example , the half life of the loss of tritiated water from
the upper layers of the soil is measured in days / 5 /, whereas fission products
and actinides can contaminate land and buildings for very long periods .            There
is no evidence or known mechanism for its concentration in the food chain .
        Tritium was at all times present in the world atmosphere , the natural
inventory of today ( equilibrium concentration ) is in the range of 7 to 1 4 kg ,
primarily produced by the interaction of cosmic rays and nitrogen nuclei .
         Man made tritium reaching the atmosphere by far exceeds this natural
inventory .    Data on tritium production and release are scarce .        As an example
up to 1974 the maximum annual release from the Savannah river plant was
evaluated to be about 70 g / 6 /. Thermonuclear weapon testing in the atmosphere
is responsible for about 90$ of the present worlds atmospheric inventory of
tritium .     For   example the integrated releases over all years of weapons
testing up to 1978 summed up to about 700 kg , leading to a maximum inventory
in the atmosphere of about 310-450 kg in 1963 , declining to 120-170 kg in 1980
/ 6/ .
Tritium systems inventories
         The evaluation of the tritium inventory in fusion reactors is strongly
dependent on design choices and on details of reactor systems design .           Lack of
information    on   tritium   behaviour   in  materials    is an  additional  source    of
uncertainty .     The main uncertainty arises from design alternatives in plasma
feed    and  exhaust ,    isotopic  separation ,    breeding   blanket , fuel   storage .
 ---pagebreak---                                                                                  25 .
However progress has been achieved in recent years during the definition of
experimental reactors like NET and INTOR , and the tritium inventory figures
have tended to decrease . It can also be stated that the design data of the
tritium cycle in an experimental reactor can be transferred to Tokamak power
reactors .    In fact , since fusion physics does not allow small dimensions and
zero power in a representative experimental device , there will be no
significant uprating in design data from experimental to power reactors . The
present data applicable to FCTR are about 3 kg .
Mobilizable tritium inventories
       The definition of mobilizable inventory is somewhat arbitrary without a
thorough accident analysis .        It can be stated ,      however ,  that tritium in
process systems such as plasma chamber evacuation , plasma exhaust impurity
processing ,   solid breeder tritium recovery ,       plasma fuel delivery ,      coolant
loops , has higher probabilites of releases to the environment than tritium
perriieated in structural materials or stored in stable form .
       Tritium mobilizable inventories quoted for INTOR / 8 / are 500 - 1600 g ,
with   maximum    localized  inventories   of   150    -  900   g,  the   higher   values
pertaining to solid ,     the lower values to liquid breeder options .             Design
guidelines proposed for NET / 9 / would seek to maintain localised tritium
inventories    which   could  be released    under   accidental    conditions   into   the
surrounding     containment  to  below   150   g.   It   is   expected   that  the    main
mobilizable inventories of FCTR will be not much larger than those of NET ; a
careful    estimate   for  FCTR  leads  to a value of about           200g .   Operating
experience with an engineering test reactor will permit the tritium handling
of FCTR to be optimized with respect to mobilizable inventories ( if this turns
out to be an important design objective ).
2.4.2    Neutron induced radioactivity
General remarks
        In fusion reactors neutrons formed in the fusion process will activate
the surrounding structures .      The plasma facing components such as the first
wall will be subjected to extreme conditions of the fusion environment .                At
the same time , they will build up the major fraction of the neutron induced
radioactivity in the plant .
 ---pagebreak---                                                                                   26 .
           It is very likely that the austenitic stainless steel AISI 316 or a
comparably well established martensitic steel will have to be the selected
material for experimental reactors such as NET .            These steels , however , being
optimized to meet requirements for use in fission power plants are not an
optimal choice for fusion ( due to their relatively high activation
potentials ). To meet fusion requirements further developments could lead to
the use of austenitic and martensitic steels with constituents chosen in order
to have improved strength and a lower level' of induced activation .                In the
long term the use of low activation alloys can be seen as an important R+D
( research and development ) objective .
Activation inventories
       The total radioactive inventory of FCTR at shut-down , with the parameters
indicated in Table 1 , and AISI 316 as structural material can be evaluated to
be 333,000,000 TBq (9 GCi ) of activated products            after about 5 years of full
                                  2
power operation ( 10 MWa/ m ) / 7 /.            About i| 3%   of this radioactivity is
concentrated in the first wall , with a maximum               value of 9-6 TBq/ cm ( 260
       3
Ci / cm ), 47$ in the blanket structures , 8$ in the         breeder material , and 2$ in
the inner shield .       The specific radioactivity of the breeder material is of
                              3 ,   _. .  3
the order of 148 GBq/ cm        (4 Ci / cm ) in the case of the 17Li83Pb eutectic , and
is mainly due to neutron interaction on lead .
         The neutron induced radioactivity of FCTR decreases after shut-down of
the plant to about 30$ within one year .                The residual radioactivity of
structural materials after 10 years and 100 years is 2.5$ and 0.02$ ,
respectively . The contribution of the 17Li83Pb breeder becomes relevant ( more
                                                                                4
than 10$ of the total ) only after very long decay times ( more than 10 years ).
        However , as mentioned previously , it is more realistic to assume that in
the future improved structural materials other than AISI 316 will be used for
fusion . power     reactors .     The    following  structural    materials  with   a  low
potential for neutron activation are already under development :
- Austenitic stainless steels modified to replace Ni with Mn and Mo with W
    and / or V. The steel AMCR-33 is an example of this family , since it does
    not contain Co and Mo , and Ni is reduced to 0.1$ .               With this material
    instead of AISI 316 significant reduction in radioactivity inventory can be
    expected for long decay times ( better than a factor of 10 after 100 years ,
    see fig . 3 ).
 ---pagebreak---                                                                                 27 .
- Ferritic-martensitic steel in which Mo and Nb are replaced by W , V and Ta .
    The advantages will be comparable to those of AMCR-33 .
- V15Cr5Ti : The radioactive inventory will be about one order of magnitude
    lower compared to AMCR-33 and also the radioactive decay rate will be
    faster ( see Fig . 3 ).
2.5    Indices of radiological hazards
        Various indices of radiological hazards exist to quantify the danger to
the    public   posed    by unanticipated        releases  of  radionuclides   into  the
environment .
2.5.1    Activity
       The most widely available but also the least informative measure for the
hazard is the activity defined in Becquerels (= desintegrations per second ) or
in Curies . Using this measure , a fusion plant employing steel ( AISI 316 ) as
structural material will be comparable to a fission plant of similar power
because the radioactive inventory            is about the same . The use of vanadium
alloys ( e.g. V15Cr5Ti ) reduces the activity by about one order of magnitude .
2.5.2    Biological hazard potential
         The potential biological consequences of steel activation products is
considerably lower than that of fission products and actinides .             To quantify
this effect , a more meaningful index , the biological hazard potential ( BHP ) is
used .     It  takes   into account      the   differences  in  such hazard-determining
properties as half-life , decay mode and energy , radioactive progeny of the
radionuclides , and lifetime in the body tissues .
       The BHP is defined as the activity ( A ) divided by the maximum permissible
concentration ( MPC ) of a radionuclide , summed for all radionuclides present :
                   BHP = I(A 1 /MPC i.)>
 ( The MPC is the concentration of a radionuclide in air or water that would
produce the maximum permissible dose if a person were breathing continuously
the contaminated air or drinking the contaminated water.).
 ---pagebreak---                                                                      28 .
       Using the such defined BHP for comparison , results in hazards about 2
orders of magnitute smaller in the fusion case ( AISI 316 ), than in fission .
This  difference  increases  with decay time and   the scenario is even more
favourable to fusion if vanadium alloys or other low activation materials are
used as structural materials .
 ---pagebreak---                                                                                  29 .
3.     ENVIRONMENTAL IMPACT OF A FUSION POWER PLANT
3.1 Radioactive releases
3.1.1 General remarks
           In the following sections the potential environmental impact of FCTR is
outlined , for both normal operation and accidental situations .           The background
information on which this report is based is given in references / 7 / and / 10 /
to / 1 3 / and the literature quoted therein .      It represents the state of present
day knowledge .         As FCTR is still in the preconceptual stage this assessment
can only be very general .
            Tritium is the most volatile part of the radioactive inventory .           To
minimise       its release to the environment , a multiple-containment concept is
used .        The  inner   primary  containment  consists   of   the  tritium containing
equipment .       This all-metal equipment is installed in a secondary containment
( e.g. glove boxes , jacketed tubing ) which is as small as possible in volume to
allow       continuous     extraction   of  tritium   from   the   enclosed   containment
atmosphere .       The tertiary containment acting as a last barrier against tritium
release into the environment constitutes the reactor building ( with steel
liner inside ), the tritium facility building or other air-tight buildings , see
f ig . 1 .    The atmosphere of these buildings may also have to be detritiated by
an emergency clean-up system in abnormal and accident situations .
           The availability and performance of atmospheric clean-up systems are of
vital       importance    for   the  effectiveness  of   both   secondary   and  tertiary
containments .        In addition , the reactor building is slightly underpressurized
to prevent outward leakage from the containments .
3.1.2 Radioactive releases during normal operation and maintenance
          Most routine releases of radioactive products will originate from liquid
waste processing systems and from ventilation systems of various buildings
where radioactivity may become airborne .           The liquid and gaseous effluents
( consisting of tritium and . gaseous corrosion products )              are continuously
monitored and are released into the environment under controlled conditions .
 ---pagebreak---                                                                                    concrète containment
     heat exchanger                                                                steel liner
                                                                                   cryostat
     primary coolant                                                               vacuum vessel
     loop                                                                          breeding blanket
     secondary                                                                     first wall
    coolant
                                                                                   plasma
    loop
       |
                            / 1^          ///// ”
                                                               *ue^'n9 " ~| /
                                                               fuelling
                                                                            '^//S77ZZ//////////Y;                      /
        \                                t///- –^ \\                        / l tritium
                                                                                    trltlum         <                  /
                                                                                                                       /
                                                                                                      . I.1 -
                                                                                    recovery
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                                                                                                      purification . '
                            /        \\   //      \V      J)                 /           . - / isotopeisotope |||
                            /
                            /
                                      \\-S/        \ W/                      /      tritium
                                                                                    tritium
                                                                             / Ustoraae
                                                                                  Lstoraae -M / l 1
                                                                                                    / separation^
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                                                                                                         K             *
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                            /
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                           A                           n bump pump       rrc-^–v                                       /
                                                                                                                       /
         _a_                                                                H_u_                                       Z
          turbine building        reactor building                                  tritium System building
Fig . 1 : Schematic view of the multiple containment concept of a fusion power plant
 ---pagebreak--- Tritium
           The major sources of tritium release during normal operation and
maintenance are :
   - leakage and permeation from the plasma chamber and fuel handling system ;
   - leakage from first wall and blanket coolant lines , leakage from steam
     generators ;
   - leakage and permeation from tritium processing system .
      To quantify tritium releases it is common to use both mass units ( g ) and
activity units ( Bq or Ci ), the correlation being the specific activity of
about 370 TBq / g or 10000 Ci / g .
       All critical tritium-containing components are located in the tritium
facility building or the reactor building .           Estimates of the atmospheric and
aquatic releases of tritium from the FCTR are given in tables 2 and 3 , taken
from 77 / .
TABLE 2 - Annual atmospheric emissions of a fusion reactor ( FCTR )
                           Operation            Maintenance             Totals
                          TBq          ( Ci )     TBq     ( Ci )
Tritium
Coolant system            185       ( 5000 )     56      ( 1500 )      about
Torus                       0.4         ( 10 )  185      ( 5000 )      450 TBq ( 12000 Ci ) as HTO
Diagnostics                                      37      ( 1000 )    + 330 TBq ( 9000 Ci ) as HT
Process system              4         ( 100 )
(+ waste preparation )                          117      ( 3000 )  = 780 TBq ( 21000 Ci )
Tritium recovery           11         ( 300 )
Reactor hall                                     185      ( 5000 )
                          200          ( 5410 )  580    ( 15500 )
Activation products*^                                                   18 GBq ( 0.5 Ci )
Cover gas                negligible ( with hold-up tank )
+)
    Data for AISI 316
 ---pagebreak---                                                                                       32 .
TABLE 3 “ Annual aquatic emissions of a fusion reactor ( FCTR )
                                   Operation and Maintenance
                                     TBq              ( Ci )
Tritium*^                           55.5            ( 1500 )
Activation products**^               0.185               (5)
                                        .
+)
       Mainly due to losses during maintenance of coolant systems ,
       but also including streams from waste processing .
++)
       Assuming resuspension of corrosion products in the coolant .
         The largest internal loss of tritium during normal operation is expected
to occur from the water coolant lines .          It originates from tritium permeation
into the primary coolant system ( few g / d ) and by permeation and leakage
through the heat exchangers into the secondary coolant circuit .
          The operating experience of existing CANDU HWR ( heavy water reactor )
plants      with  comparable    tritium    concentrations    in the   coolant including
improved tritium containment measures , provides a good basis for the estimate
of tritium leakage from the coolant circuit of FCTR .             Tritium concentration
in the coolant can be maintained at a very low level of order of 37 GBq/ 1 (1
Ci / 1 ) by employing permeation barriers and present technology of detritiation
systems .       Taking  into   account    present   developments   for  CANDU reactors ,
unrecovered water leakage from the primary coolant into the reactor hall are
expected to be less than 10 1 / d , / 1 4 / , resulting in a tritium loss of about
185 TBq/ a ( 5000 Ci / a ).    The atmospheric tritium release from the secondary
coolant loop can be maintained at a small fraction of that from the primary
coolant circuit .
          There exist many more uncertainties on tritium inventory and tritium
recovery from solid breeder materials than for liquid breeder materials .            It
was estimated that the tritium loss from the tritium recovery system is less
than 11.7 TBq / a ( 300 Ci / a ), for both concepts .
 ---pagebreak---                                                                                     33 .
      The routine tritium loss from the fuel handling system and other tritium
processing systems in the tritium facility building is expected to be in the
order of 3*7 TBq/a ( 100 Ci / a ) if efficient multi-containment and detritiation
systems are provided .
      The dominant contribution of the tritium loss to the reactor building of
about 555 TBq / a ( 15000 Ci / a ) comes from maintenance operations on plasma
chamber , from blanket replacements , and from coolant system maintenance .              If
necessary much of the tritium released during maintenance could be removed by
the emergency clean up system or by temporary secondary enclosures around
critical areas with detritiation of the enclosed atmosphere .
        As shown in table 2 the total annual atmospheric tritium emission will
be about 777 TBq ( 21000 Ci ), of which about 60% is in the form of HT0 and 1)0%
as HT .
        The aquatic emissions will be about 55.5 TBq ( 1500 Ci ), mainly due to
losses during maintenance of coolant systems , but also including streams from
waste processing .
        These tritium releases from the FCTR of a few TBq / d ( about 800 TBq / a )
might be acceptable .      This implies a leak tightness of the tritium system of
1  ppm / d   of  the  gaseous   as  well  as  the   liquid  circuits .     The  required
containment appears to be within reach and large scale demonstration of these
capabilities is in progress / 1 5 / .
Activation products
         Assuming water cooling the dominant sources of activation products as
discharged during normal operation are the corrosion products leaking from
the primary coolant circuits .
        Much of the corrosion products are deposited on the inner surfaces of
the primary coolant pipes and the primary side of the steam generator .               The
water treatment system controls the concentration level of dissolved material
in the coolant , being in the range of 1 to *1 GBq/m^ ( 0.03 to 0.11 Ci /m^).
           Approximately 18.5 GBq / a ( 0.5 Ci / a ) of activated products will be
released     from  the  coolant   circuit  at  a  leak  rate  of  10   1/d .   The  main
 ---pagebreak---                                                                               34 .
radionuclides are Fe55 , Fe59 , Mn54 , Mn56 , Cr51 , Co58 , C06O .    The discharge is
assumed to be into the reactor building atmosphere by all-vapour leakage ,
although some of the losses to the aquatic system should also be considered .
The    atmospheric     release    could  be  significantly     reduced   by  efficient
filtering .
         The deposition of the corrosion products on internal surfaces causes
radiation      levels   which  are   of particular   concern   during  inspection   and
maintenance operations .
        Coolant water lost during maintenance will have an enhanced level of
activation products due to resuspension of the crud normally adhering to the
pipe walls (a factor of 100 has been reported ).             This leads to estimated
aqueous releases of 0.185 TBq/ a (5 Ci / a ) of             corrosion products from
maintenance operations .
Building cover ga3
      The activation of the air atmosphere in the reactor building , mainly due
to neutrons leaking from the shielding ,          results   in the build-up of some
radionuclides such as Ar4l and C14 which is formed mainly by the reaction
14         14
   N(n,p )    C.   The use of C0 „ as cover gas would reduce the production of this
                              6 *
nuclide by a factor of 10 .
3.1-3    Potential releases of radioactivity in accidental conditions
General
       Because fusion reactor designs are still at their conceptual stage , any
attempt to quantify non-routine releases of radioactivity is difficult at the
moment .
           For some identified cases maximum possible consequences have been
estimated .       As  fusion  safety studies and reactor designs develop ,         more
credible accidents will        be able   to be   identified ,   not just the maximum
consequences of accidents .
 ---pagebreak---                                                                                35 .
       The definition of potential sequences of accidental events does not
necessarily mean that such accidents will occur frequently or even at all .
Many design features are likely to be envisaged to minimise the probability
of  accidents    and  to  reduce   or  even   exclude   the   consequences  to   the
environment .   Moreover fusion reactors are expected to have a low potential
for accidents which may affect the general public , due mainly to the generic
safety features .
          Two major mechanisms are required       for   an accidental   release of
radioactivity to the environment :     both the volatilizing and mobilizing of
potentially hazardous material and the rupture of the containment .              The
building containment is designed to prevent most materials from reaching the
environment ,  therefore non-routine losses from components normally do not
result in releases which endanger the public .
Possible accidentai tritium releases
      Estimates have been made for INTOR and for other conceptual designs of
the upper limit and the area of tritium loss which can arise from a number of
identified potential accidents / 7 /. These figures are also applicable to a
power  reactor   like FCTR  since a significant      increase   in the mobilizable
inventory is not expected .   They allow the evaluation of the possible tritium
release to the environment and their dose rate to the public .
      In the most severe cases ( rupture of coolant pipes , failure of part of
the tritium processing system , failure of cryopump ) up to 200 g of tritium
can be released into the reactor building .        Tritium may also be lost from
rupture of components inside the tritium recovery and isotopic separation
system ( order of 100 g ), but this loss is within the secondary containment .
Taking into account tritium removal by the detritiating system of the
secondary containment a subsequent tritium release of 0.1 g/ h into the
process hall might be expected .
      Quick detection and effective performance of the emergency atmospheric
clean-up system in the reactor building or process building should be capable
of reducing the personal exposure and the re Lease outside the building to
about 100 GBq / d (a few Ci / d ).    However , for the worst case analysis of
environmental impact no retention mechanism will be accounted for .            As a
reference case for this report a maximum accidental release of 200 g tritium
 ---pagebreak---                                                                                     36 .
to  the   environment   was   defined .    This  source   term   is the basis  for the
determination of the radiation exposures of individuals in the surrounding of
the plant .
Potential release of activation products
       The accidental release of activation products is the most difficult to
assess .    The most mobilizable parts of the plant 's radioactive inventory are
the  fluids    e.g.  the  primary coolant system .        The radioactive structural
material for which melting and vaporization is required for mobilization and
release to the environment has the lowest level of mobilizability .           There is
even hope that , due to inherently safe design , melting of structural material
may be effectively excluded .
      The following most relevant potential mechanisms to mobilize activation
products have been identified :
   - plasma disturbances ;
   - coolant system failures ;
   - magnet failure ;
   - cryogénie depressurization ;
   - hydrogen explosion ;
   - fire ;
   - auxiliary system failure and external hazards .
       The most probable release of activation products in case of accidents
are those related to structural heat-up of first wall and blanket , namely
plasma disruptions and blanket coolant failures .
        The most pessimistic assumption resulting from a plasma disruption is
the release of some grams of ablated first wall material through a broken
vacuum vessel into the reactor hall .          However , most of the eroded material
from   the    first wall    may   be   redeposited   inside    the  plasma chamber  or
elsewhere .
         The main concern in a cooling failure is related to the decay heat
following shut down of the reactor . It has to be expected that in case the
cooling failure is not detected , the plasma burn will automatically be
terminated due to the        ingress of volatilized material subsequent to the
 ---pagebreak---                                                                                  37 .
temperature rise of the first wall . Depending on the design of the blanket
and cooling system different scenarios of coolant system accidents can
follow .     In the most pessimistic case of cooling loss the afterheat
production causes melting and degradation of the structure and consequently
release of activation products only after some hours .             This would appear
sufficient time for      intervention .   Moreover ,  with passive cooling design
solutions and proper material selection , melting of the structure appears to
be inherently avoidable .
        Coolant tube breaks would lead to the release of radioactive corrosion
products ( and tritium ) present in the coolant , and possibly to the generation
of mobile     material  subsequent   to  the  temperature    rise or    break of   the
structure or by chemical reactions .
       The only important accident initiators which could lead to damage of the
magnet and / or other reactor components are arcing across current leads or the
rupture of a single conductor .     Simultaneous rupture of a complete winding at
two different locations has been postulated for the severest event .              The
probability of this event however is extremely low because the prerequisite
leading to such an accident is scarcely imaginable from the physics point of
view .   If such a hypothetical accident is assumed , the broken section could
be accelerated leading to some damage on reactor components such as coolant
lines or tritium processing lines .        The building containment however will
withstand this hypothetical accident as it is designed to withstand even
worse    events    like  airplane   crashes   and    explosions .      Therefore  the
consequences of arcing would be      mainly in terms of economics due to reactor
downtime and costly repair .
       The same holds for an accidental release of He being used as coolant for
the superconducting magnets .     First calculations indicate that the building
containment can be designed to withstand the pressure loads resulting from
evaporation of the total He inventory .
       It is difficult to exclude , as in all complex systems , a fire accident .
However , care is already being taken to choose materials , wherever possible ,
so that    this  event will  be minimized .   This   is  the  case  for the breeders
where materials such as liquid LiPb and Li-ceramics are now preferred to
lithium metal because of their low chemical reactivity with air and water .
 ---pagebreak---                                                                               38 .
       In case of external events ( earthquakes , missiles , aircraft , sabotage )
the tritium which may be involved will at most be that which is contained in
one of the tertiary containments ,        i.e. the reactor building or the tritium
process building ( containing about 100 g of tritium divided between separate
isolated rooms ).        It is a likely assumption that in case of accidental
release    of  activated     material   in   the  reactor building   deposition    and
adsorption effects will strongly reduce the emissions to the environment .
3.2   Radiological effects to the environment
       The dose to an individual ( measured in rem or Sv = Sievert ) at defined
distance from the plant , obtained during a defined time of exposure is the
most meaningful hazard index .         However , to perform dose calculations many
assumptions must be made , leading to greatly varying results .
3.2.1   Dose criteria for normal operation and abnormal events
       Dose criteria are given in the CEC directive 80 / 836 which is in close
agreement with ICRP recommendations / 1 6 / .        The basic recommended maximum
allowable annual dose limits for whole body radiation are :
- 50 mSv/ a (5 rem / a )    for the most exposed working group , and
- 5 mSv/ a ( 0.5 rem / a ) for the Most Exposed Individual ( MEI ) of the public .
   These limits are intended for conditions where the source of radiation is
   subject to control and therefore do not apply to doses from accidental
   releases .
Exposure limits used as design guidelines follow the As Low As Reasonably
Achievable ( ALARA ) principle and are more restrictive .      The following values
are frequently used :
- for normal operation 1 to 2 mSv/ a ( 0.1 to 0.2 rem / a ) as average dose and 5
   to 10 mSv/ a ( 0.5 to 1 rem/ a ) as maximum dose for the most exposed working
   group ; 0.1 mSv/ a ( 10 mrem/a ) as average ( with range of 0.05 to 0.3 mSv/ a (5
   to 30 mrem )) for the MEI
 ---pagebreak---                                                                               39
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                            Whole body dose D [Sv]
Fig 2 : CEGB Safety Criteria for Accidental Releases and Exposures to the
        Public /1 7 / -
 ---pagebreak---                                                                                     40 .
- for* abnormal events doses in excess of the regulatory limits are accepted .
   These values are 50 to 150 mSv (5 to 15 rem ) for events with a probability
                      -7
   of less than 10        per year (= hypothetical accidents ); 0.3 to 5 mSv ( 30 to
                                                     -4 . . ..-2
    500 mrem ) for events of low probability ( 10       to 10     per year ) and 0.05 to
                                                                               -2 .      -1
   0.3 mSv (5 to 30 mrem ) for events of moderate probability ( 10                to 10
    per year ). The values refer to the MEI , values for workers are a factor of
    10 higher .
           As   an example    fig . 2 shows the CEGB design safety criteria for
accidental releases and exposures to the public / 17 /.               It correlates the
total    permissible     frequency   per  reactoryear     with   the   whole   body   dose
equivalent .      A value of 100 mSv ( 10 rem ) is considered as lower limit at
which consideration should be given to the countermeasure of evacuation .
        As tritiated water ( HT0 ) is more readily absorbed by human tissue and
therefore more hazardous than gaseous HT , the permissible concentration of
HT0 in air is much smaller ( factor 25000 ) than that of HT . If tritiated gas
is released into the environment it will subsequently convert to HT0 ( order
of 1$ per day ).      In making estimates for the radiation dose it is therefore
common use      but  conservative ,   to assume that all       the atmospheric tritium
release to the environment is in the form of tritiated water . Tritium in the
aqueous effluent is already in the form of HTO .                The whole life ( 50 a )
committed dose equivalent from intake of tritiated water ( inhalation or
ingestion ) is taken according to ICRP 30 / 1 8 / to be 17 Sv/TBq ( 64 rem/ Ci ).
3.2.2    Radiation doses from routine émissions
         The annual routine atmospheric emission of treated gaseous effluents
from a FCTR is likely to contain about 777 TBq ( 21000 Ci ) of HTO , 18.5 GBq
( 0.5 Ci ) of activation products ( namely Fe , Mn , Co ) and negligible quantities
of Cl 4 and Ar4l . This discharge is expected to be through a 100 m stack to
achieve ci high degree of dilution in the atmosphere . The routine aqueous
emission of radioactive products of 55.5 TBq /a ( 1500 Ci / a ) as HTO , and
0.185 TBq/a (5 Ci / a ) as activation products occurs via the cooling tower
blowdown flow and to an offsite river with a high degree of dilution .
        External doses to exposed individuals result from gamma radiation from
plumes , exposure to contaminated ground surfaces , immersion in contaminated
air    and   submersion    in  contaminated  water .     Internal    doses   result   from
 ---pagebreak--- inhalation of air , ingestion of contaminated food and water .             It is assumed
in the dose calculations that individuals are exposed 10056 of the time to the
contaminated air and ground surface , and that all food consumed is from the
locality . Maximum conservative annual doses calculated for the MEI living at
about 1 km from the stack , is about 0.015 mSv/ a ( 1.5 mrem /a ). ( 0.0065 mSv/ a
( 0.65 mrem/a ) from atmospheric HTO , 0.004 mSv/ a ( 0.4 mrem/ a ) from atmospheric
activation products , and 0.004 mSv / a ( 0.4 mrem/ a ) from aqueous release ).
This is about 1 % of the average dose burden by natural background
irradiation , being 1 to 2 mSv/ a ( 100 to 200 mrem/ a ).
      The collective dose of the local population living in the area within 50
km radius from the plant ( 2.4x10°6 persons at a density of 300 persons/kni 2 ) is
calculated to be about 0.3 man Sv / a ( 30 man rem / a ) , about equally from HTO
and activation products . The average whole body dose for the general local
                   -4
public is then 10     mSv/ a ( 0.01 mrem/ a ).
       For a fusion powered world economy with 2000 fusion reactors all over
the world , each routinely releasing the above activity of tritium , the global
                                          -3
average dose to man would be below 10          mSv/ a ( 0.1 mrem / a ).
3.2.3   Radiation doses from accidental releases
Tritium
       The possible accidental releases from a FCTR to the surroundings are
still uncertain but are hypothesized with moderate conservative assumptions .
As the reference source term for a hypothetical accident a release of 200 g
tritium as HTO in a 30 min discharge from a stack of 100 m is assumed . The
dose pathways are skin absorption and inhalation .                  The outcome is much
dependent on wind velocity distribution and distinction between dry and wet
deposition ( rain reduces the skin and           inhalation dose rate ).       For worst
weather conditions ( Pasquill type B ) the maximum dose as calculated for MEI
is 2.4 mSv ( 0.24 rem ), at 700 m from the stack . For other weather conditions
the maximum dose will be 0.5 to 0.7 mSv ( 0.05 to 0.07 rem ) at distances of 5
to 15 km .
       A hypothetical release of 200 g tritium as HTO from the building roof
( release height 20 m ) would cause ( at 1 km distance , under worst weather
 ---pagebreak---                                                                          42 .
conditions and dry deposition ), a maximum dose of 60 to 80 mSv (6 to 8 rem ),
which would not disrupt society in the immediate surrounding .     These values
are within the limits of 50 to 150 mSv (5 to 15 rem ) accepted by the
licensing authorities for abnormal events of low probability .
       Similar results were recently obtainod from worst-case analyses for the
US conceptual design MARS ( Mirror Advanced Reactor Study ) / 1 9 /.    Assuming
ground level release of 50 g tritium ( HT0 ), which is defined to be the total
vulnerable inventory in MARS , results in a maximum off-site dose of less than
0.04 Sv (4 rem ).      Even if an additional 200 g of HTO were released , the
maximum off-site dose would still be less than 0.25 Sv ( 25 rem ), the present
NRC limit for emergency releases .
          The above mentioned values assuming worst case conditions could be
compared with measured and evaluated doses of a real accidental release of
about 50 g of tritium gas from a Savannah River Plant exhaust stack ( 60 m ) to
the atmosphere over a period of about four minutes / 20 /.      Measurements of
tritium offplant indicated that less than 1 ? of the tritium was in oxide
form , and the remaining 99% in the much less radiotoxic gaseous form .        A
maximum potential dose to a person ( from inhalation and skin absorption ) at
the puff centerline on the plant boundary was calculated to be 0.0014 mSv
( 0.14 mrem ), less than 1$ of the annual dose received from natural
radioactivity .
Activated structural material
          The evaluation of the quantity of accidentally " mobilised " erosion
products leads to a few cubic centimeters of activated first wall material
which may be released to the environment .    The corresponding dose rate , even
in the case of the less suitable material AISI 316 , will be much less than
the dose rate due to the release of 200g tritium which may occur in the same
sequence of accident events .
3-3 .   Waste
        Two categories of radioactive waste will be produced in a fusion power
plant :
     low and medium level waste arising from the processing systems ( i.e. fuel
    cycle and coolant     purification systems ) and  from decontamination and
    maintenance operations ;
 ---pagebreak---                                                                                              43 .
                                                             3      ...     _3
-• high     level   waste    ( more    than    3 ■ '! TBq /m     =  100     Ci /m )   derived     from
    disassembly and      periodic      replacement         of    parts of      the   inner  nuclear
    structure ( mainly first wall and blanket segments ).
          The wet and dry low and medium level wastes ( containing tritium and
activation products ) are of the same nature and have a somewhat higher volume
( 900m with an activity of 44.4 TBq = 1200 Ci ) than the waste streams from a
fission      power   plant ,    but   the    contaminants      have   shorter       half-lives     and
therefore become inactive much sooner .                  The waste management and disposal
strategies as developed for fission reactor plants may be applied , providing
that     sufficient    tritium      recovery / removal       and   tritium      immobilization      is
applied to these wastes .            After waste treatment and packaging near-surface
burial is permitted .
         Handling and treatment of dismantled blanket segments may involve more
complex procedures because of their volume , weight and activation level .                          If
AISI-316 is used as structural material , in the short term the management is
comparable with that for - spent fuel elements of a LWR ( light water reactor ).
After an initial cool down period tritium ., breeder material and some other
valuable      elements with      low specific activity may be separated for later
reprocessing and re-use .            The remaining highly active structures will be
compacted ,     fragmented ,   detritiated and conditioned for intermediate storage
/2 1 / .    After   the   decay     heat    becomes     negligible      ( and   depending on       the
composition of the materials             involved it takes from a few years to many
decades ) the waste can be classified , recovered for recycling or transported
to final repository .
          Assuming AISI-316 as structural material ( large experience exists on
this material due to its use in fission reactor plants ) the first wall and
parts of the blanket structural wastes will need a deep geological deposit .
AISI-316 however is not well suited for fusion uses .                        Therefore for fusion
power plants other structural materials will be developed .                          As an example
fig . 3 shows the neutron induced activity for these advanced materials , as
compared to AISI-316 , as a function of time . According to present rules for
waste disposal , the AMCR type of steels ( austenite , without Co and Mo ,
reduced Ni content ) could be deposited at the surface ( Surface Land Burial )
after a time of 30 to 100 years .                     For V-Cr refractory materials ( e.g.
V15Cr5Ti ) the picture is even more optimistic . In these cases , however , the
question of impurities arises ,             which could make a significant contribution
to long-term activity .
 ---pagebreak---                      T                _| ~T     J     K      S       l      l       l
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    Fig 3 : Neutron induced activity of FCTR first wall
 ---pagebreak---                                                                                45 .
      In conclusion , with a suitable research and development effort , one can
expect that the wastes from fusion should not require deep geological
disposal but simpler near-surface land burial would be sufficient .             Non-
structural materials such as solid breeder materials ( e.g. lithium oxide ) may
be recycled after a few days .     LiPb , however , will not satisfy the recycling
conditions due to the high residual activity of the Pb impurities .
3.4  Other sources of hazard
       Potential additional hazards for the workers inside the plant and the
men near    the site   are  of  various   kinds .   However , no  difficulties   are
expected in conforming to existing guidelines .
        Sources of direct radiation originate from holes in the shield ( e.g
penetrations for diagnostics ), leakage of neutrons through the shield and
permeated tritium ,    from the activation of the building atmosphere and from
maintenance , repair and replacement operations .      No detailed estimates exist
of such occupational doses , but designs can be realized to keep them below
permissible levels .    The external radiation at the site boundary can be made
as low as desired by appropriate shielding design .
      Exposure to high magnetic fields will not be of concern .         There is no
evidence that long exposure to the expected fields of 0.05 Tesla in the
reactor hall constitute an occupational hazard .          It is not likely to be
difficult to make the design guidelines of FCTR conform to presently existing
laboratory rules concerning long term exposure to magnetic fields .        The same
can be said for the exposure to radio frequency radiation from the proposed
RF heating systems and from the plasma .
      Although the fuel cycle is an integral part of the plant , transport of
some tritium quantities outside the plant are foreseen ( e.g. to start-up new
reactors ). The present regulations concerning tritium transport and shipping
are so stringent     that  tritium release    from the   transport flasks to the
ambient is practically nil in both normal and abnormal conditions .
 ---pagebreak--- M.   DEVELOPMENT POTENTIAL
        Work is under way to further reduce the already small environmental
impact    of  fusion   as   derived    from  todays     technologies .    Considerable
development potentials exist in the following areas :
- limitation of waste quantities by improving life time of first wall and
   blanket components ,
- reduction of activation by choice of modified steels containing less nickel
   and molybdenum ,
- reduction    of   activation    by  choice   of   new   structural   ujaterials   ( low
   activation materials ),
- decrease    of  tritium   inventory   in  the   plant   by  appropriate   choices    of
   materials and processes ,
- reprocessing of blanket materials .
       In the long term other fusion reactions than D-T like D-D or D-He3 are
much   more  attractive   from   the  radioactivity    hazard   point  of view .      The
reactor would also be substantially simplified because there would be no need
for a breeding blanket .     Even if the feasibility of these cycles is far from
being proved , these features represent a stimulating challenge for the long
term issue of fusion .
 ---pagebreak---                                                                            47 .
5.   CONCLUSIONS
        Fusion as an energy source is based on nuclear reactions and therefore
the main hazard to the public is due to the presence of radioactivity .         The
sources of radioactivity are tritium and the neutron - induced transmutations
of the plasma surrounding structure .
       Magnetic fusion reactors appear to have very important intrinsic safety
features , such as :
- the    impossibility   of  an  uncontrolled ,   self-started  and self-sustained
   nuclear power runaway ,
- the absence of long-lived volatile radioactive materials ,
- the relatively low power density in the first wall and blanket structure
   during operation ,
- the moderate afterheat at shutdown ,
- the closing of the tritium cycle on reactor site .
       The levels of radioactive effluents in normal operation will match the
regulations in Europe and elsewhere and hence these effluents will not be a
hazard to the public .    It is worth noting that the technical potential exists
for further reducing the emission to virtually insignificant levels .           The
radioactive    waste   generated  by   fusion   reactors   will  be quantitatively
comparable to fission reactors , but qualitatively it will be much less of a
potential hazard .
         The analysis of volatile inventories released after major technical
failures leads to the conclusion that the radioactive effluents ( mainly
tritium ) in such cases would have a very low impact on the lives and the
health of the surrounding population .        Therefore , in no case would fusion
cause a major disruption of normal life in the community outside the reactor
site .
 ---pagebreak---                                                                                       48 .
6 . REFERENCES
/1 /   INTOR Phase Two A ,          Part II - Panel Proceedings Series , IAEA , Vienna ,
       1986 .
/ 2/   NET Status Report .         NET report 51 , EU - FU/XII - 80/81 /5 1 , December 1985 .
/3/    W.R. Spears ; DEMO and FCTR Parameters , NET Report Nr . 41 ,
       EUR -FU/XII - 361/85/41 , August 1985 .
/4/    STARFIRE - A Commercial Tokamak Fusion Power Plant Study .                 Argonne
       National Laboratory Report , ANL/FPP - 80 - 1 , September 1980 .
/ 5/   I.R. Brearley ; The Hazard to Man of Accidental Releases of Tritium .                  SRD
       R 331 , March 1985 , SRD-UKAEA .
/ 6/   F. Luykx , G. Fraser ; The Environmental Tritium Inventory .                    European
       Seminar on the risks from tritium exposure , MOL , 22-24 November , 1982 ,
       EUR 9065 EN .
/!/    G. Casini , C. Ponti , P. Rocco ; Environmental Aspects of Fusion Reactors ,
       1985 . Technical Note I . 04 . B1 . 85 . 1 56 . JRC , Ispra , December 1985 .
/8/    INTOR Phase Two A , Part II .           Critical Issues , Vol . II , EURFUBRU / XII -
       1 33 / 85 / EDV10 , April 1985 , Brussels .
/9/    P. Dinner , M. Chazalon , M. Iseli ; Tritium Handling on NET : Requirements’,
       Design Approaches and Development Issues . 14th SOFT , Avignon 1986 .
/ 10 / J.B. Cannon ;         Background Information and Technical Basis for Assessment
       of Environmental Implications of Magnetic Fusion Energy .
       Department of Energy Report , DOE / ER-0170 , August 1983-
/ II / R .   Hancox ,     W.   Redpath , Fusion    Reactors    -   Safety    and  Environmental
       Impact .      CLM-P750 , May 1985 , Culham Laboratory .
/1 2 / Proceedings IAEA Technical Committee Meeting on Environmental and Safety
       Aspects of Fusion .         Held 17-21 October , 1983 , Ispra , to be published .
 ---pagebreak---                                                                                     49 .
/ 1 3 / M.S. Kazimi ; Safety Aspects of Fusion , Review paper .
        Nuclear Fusion 24 ( 1984 ) 11 , p. 1461-1483 .
/ 1 4 / T.S. Drolet , K.Y. Wong , P.J. Dinner ; Canadian Experience with Tritium -
        the Basis of a new Fusion Project .           Nuclear Technology / Fusion Vol . 5 ,
        January 1984 .
/ 15/ J.L. Anderson ; The Status of Tritium Technology Development for Magnetic
        Fusion Energy . Nuclear Technology / Fusion 4^ ( 1983 ) 2 , 75-82 .
/ 1 6 / Recommendation of the             International Commission         on  Radiological
        Protection , CRP Publication 26 , Pergamon Press , 1977 .
/ 1 7 / Safety Assessment Principles for Nuclear Power Reactors . Nil .
        April 1979 .
/ 1 8 / International Commission on Radiological Protection ( ICRP ) Publication
        30 , Supplement to Part 1 , Annals of the ICRP 3 ( 1-4 ), Pergamon , Oxford .
/ 1 9 / S.A.     Fetter ;    Radiological   Hazards  of   Fusion   Reactors :   Models   and
        Comparison .      University of California , Berkley , PH.D. 1985 .
/ 20 / W.L. Marter ; Environmental Effects of a Tritium Gas Release from the
        Savannah River Plant on May 2 , 1974 .           DP-1369 , UC – 11 , Savannah River
        Laboratory , November 1974 .
/ 2 1 / K. Broden , A. Hultgren , G. Olsson , H. Djerassi , P. Giroux , P. Guetat ,
        J-L Rouyer ;      Fusion Waste Management - Safety and Environment Studies
        1983-84 - European Fusion Technology Programme , NET Report EUR-FU / XII -
        361 / 85 / 35 , 1985 .
 ---pagebreak---                                                                   50 .
T.  GLÔSSARY
Units
Sv    sievert            ( équivalent dose )
rem                            "             (1 rem * 0.01 Sv )
Bq    becquerel          ( activity )
Ci    curie              "               (1 Ci = 3-7x1 0 1 0 Bq )
W     watt               ( power )
eV    electronvolt       ( energy )          (1 eV = 1.6x10 ^ J )
A     ampere             ( electric current )
T     tesla              ( magnetic field strength )
s     second
min   minute
h     hour
d     day
a     year
g     gram
1     liter
m     meter
ppm   parts per million
multiplication factors :
                               10
                                   -3
                         m
                                   3
                         k     10
                                   6
                         M     10
                                   9
                         G     10
                                   12
                         T     10
 ---pagebreak---                                                           51 .
Abbreviations
ALARA         as low as reasonably achieveable
ALI           allowable limit of intake
ΒΗΡ           biological hazard potential
CEC           Commission of the European Communities
CEGB          Central Electricity Generating Board ( UK )
D             deuterium
DEMO          demonstration reactor
D-D           deuten ium deuter ium
D -T          deuterium - tritium
FCTR          First Commercial-sized Tokamak Reactor
HWR           heavy water reactor
ICRP          International Commission on Radiological Protection
INTOR         International Tokamak Reactor
LWR           Light Water Reactor
MARS          Mirror Advanced Reactor Study
MEI           most exposed individual
MPC           maximum permissible concentration
NET           Next European Torus
NII           Nuclear Installations Inspectorate ( UK )
NRC           Nuclear Regulatory Commission ( USA )
R+D           research and development
T             tritium
 ---pagebreak---                                                                                    52
           THE ECONOMIC PROSPECTS OF NUCLEAR FUSION       A 1986 VIEWPOINT
W.R. Spears            The NET Team , c / o Max Planck Institut fur Plasmaphysik ,
                       Boltzmannstraße 2 , D-8046 Garching bei München .
R. Biinde              The NET Team , c / o Max Planck Institut fur Plasmaphysik .
                       Boltzmannstraße 2 , D-8046 Garching bei München .
G   Grieger            Max-Planck Institut fur Plasmaphysik , Boltzmannstrage 2 ,
                       D-8046 Garching bei München .
P.E. Grohnheit         Riso National Laboratory , DK-4000 Roskilde
J. Pericart            EDF ~ Centre des Renardières , BP No.1 ,
                       77250 Moret sur Loing , France .
                                       CONTENTS
0.          SUMMARY                                                            54
1 .         INTRODUCTION                                                       57
2.          REVIEW OF PUBLISHED REACTOR COSTS AND COSTING STUDIES              58
3-          GENERATION COST SENSITIVITY                                        69
4.          DEVELOPMENT POTENTIAL FOR FUSION                                   78
5.          COMPARISON WITH OTHER POWER SYSTEMS                                82
6.          CONCLUSIONS                                                        91
7.          REFERENCES                                                         92
8.          GLOSSARY OF TERMS & DEFINITIONS                                    98
 ---pagebreak---                                                                                     53 .
ACKNOWLED GEMEN TS
         The authors would particularly like to thank Dr R          Hancox ( UKAEA ) for
carrying out the research and contributing the basic text of section 2 .
        The authors are also very grateful for the comments and suggestions of
Drs C.M. Braams ( FOM ), B Brunelli ( ENEA ), G. Casini ( JRO Ispra ), J. Darvas
( CEC ), A. Gibson ( JET ), H.H. Hennies ( KfK ), G. Kessler ( KfK ), A. Malein ( CEC ),
D   Palumbo ( CEC ), R S. Pease ( UKAEA ), F. Prevot ( CEA ), J. Raeder ( NET ) and R.
Toschi ( NET ).
 ---pagebreak---                                                                               54
0.   SUMMARY
          This report summarises todays best estimates of the cost of power
generation from nuclear fusion      These estimates can only be rough since the
earliest commercialisation date is well into the 21st century and since
development up to now has concentrated on making fusion work , not in making it
cheap . An understanding of the technical and economic feasibility of fusion
will not exist until at least the next generation of experiments , like NET in
Europe , have been operated .
       Despite these qualifications    in the last ten years several conceptual
design studies of power producing fusion reactors have been undertaken .    Such
studies are necessary since they show where fusion development is heading
thus guiding both plasma physics and reactor technology development programmes
along   reasonable  paths .   These r.tudies produce estimates of the cost of
constructing the reactors or of generating electricity , which indicate that
the economic viability of fusion is a possible ,      but by no means certain ,
outcome of the present research programme .
       For tokamaks ( the most advanced confinement method ), the direct capital
cost   in these studies varies over a factor of nearly 3 while for other
confinement schemes the range is a factor of 5 .        This indicates the wide
variety of possible methods for tackling the technological problems of fusion
and the uncertainty over the most desirable design solutions .       These costs
apply to fully commercialised designs , not the first device of a series .
Usually the tenth device of its kind is costed to take advantage of the
economic benefit of the gain with experience of manufacturing and construction
know-how .
-        As an alternative to cost   in these studies    it is also possible to
estimate the energy expended in all the processes involved in manufacturing ,
constructing and operating the power station .      Such studies show an energy
expenditure in constructing a fusion station twice that for a fission plant .
However for fission , considerable energy must be expended in producing fuel
for the plant during its lifetime whereas for fusion this item is minuscule .
The apparent fusion disadvantage is more than outweighed by this advantage .
      As part of the design definition of NET , cost methods suitable for first -
of-a kind devices have also recently been evolved . These indicate the levels *j
 ---pagebreak---                                                                                            55 .
of cost to be expected early in the deployment of commercial-Scale fusion
reactors when the manufacturing and construction design base is still growing .
Such costing methods rely heavily on design solutions proposed for NET . These
may not be the ones chosen , for technical and economic reasons , when
commercial reactors are finally designed .             For a prototype commercial-sized
reactor of 1200 MW so ( sent out ) typical of present-day plant sizes , with
plastna physics only relying on a plausible extrapolation of the results from
present-day experiments , the estimated generation cost is about 2-3 times that
for thermal fission stations beginning operation in 1995 .                      Under series
production of fully commercialised designs ( e.g. the tenth device after the
prototype ), this gap can be significantly reduced or even closed . In addition ,
a considerable     reduction    in   the cost could be achieved by a significant
increase in the ability to confine plasma and reduction in the unit cost of
design solutions , with only a modest increase in levels of power sent out .
         The present fusion programmes worldwide are geared towards solving
problems of scientific principle .         In the past , they have almost exclusively
been directed at increasing the understanding of plasma physics but , as a
consequence   of    physics    progress ,    are   now   increasingly     concentrating      on
technological feasibility .        The target of these programmes is to produce a
working   demonstration     power    reactor .      Such   a   device   would    need   to   be
technically improved and simplified to arrive at a desirable and economically
competitive   end   product .     The   combination     of   several   of   the   innovations
proposed up to now might result in substantial economic benefits .                   Most are
aimed at increasing plasma power density using theoretically feasible plasma
physics and advanced superconductors .          In this respect device compactness has
a part to play , but only to the extent that technological design margins are
not eroded and the good safety characteristics of the fusion power plant
compromised .   Many proposals , whose benefits are impossible even to estimate
today    are   not   just    applicable     to   tokamaks     but   to   toroidal    magnetic
confinement generally .
      By the time fusion power is commercially available , coal ,, fission breeder
and solar photovoltaic power stations will be the likely competitors . Solar
photovoltaic power ' costs are predicted to be a factor of 2.5-M higher than
thermal fission .    Coal , whose present electricity generation cost in baseload
is up to 60 ? higher than thermal fission plants , is expected to maintain , or
even increase , this cost disadvantage .          Fast breeders , which at present are
linked by their fuel cycle to thermal fission stations and are only just
 ---pagebreak---                                                                           56 .
beginning their evolution from the prototype commercial - sized device , although
initially ( in the first-of its kind device ) expected to have power costs up to
100 ? higher than that from thermal fission , are predicted to attain a much
more competitive generation cost compared with thermal fission , when they are
introduced on a full commercial scale .   Predictions for thermal fission depend
on the economic conditions prevailing in the middle of the next century and
extend over a factor of 2 ( Even for systems starting operation in 1995 the
cost for thermal fission can only be predicted within a factor of 1.5 ).
Fusion power thus fits alongside these estimates and from this point of view
should be able to penetrate      the market   in the   future as a large scale
generating technology .
      There are also a number of somewhat intangible but potentially beneficial
effects of electricity generation with fusion ,     in addition to those items
considered in present costings . These include security of fuel availability
( deuterium and lithium are spread widely and plentifully on earth ), low fuel
price dependence , an internal fuel cycle ( extensive off-site reprocessing
systems and their associated logistics are .     in principle , unnecessary and ,
even if needed for economic reasons , are much less than in fission ), the
potential for reduced waste hazard ( through materials optimised for fusion ),
and reduced scale of possible accidents .   To what extent these items will have
an economic impact and add to the desirability of fusion power is impossible
to estimate until more progress is made .
        The development cost for fusion power is a tiny fraction of todays
expenditure for energy supply which , given the virtually inexhaustable nature
of the fuels and their worldwide distribution , and the potential for high
environmental acceptability , should produce a highly desirable payoff .
 ---pagebreak---                                                                            57 .
1 . INTRODUCTION
       The aim of this report is to describe todays view of the cost of the end
result from the fusion development programme , in so far as it can presently be
quantified .    This is a difficult task since its earliest commercialisation
date   is well   into the next century ,      after a considerable development and
proving programme .    In todays position we are still far from the commercial
end product .     Any predictions made here must therefore be understood as
representing a considerable range around the quoted values .       Furthermore , the
programme of development up to now has concentrated on making fusion work , not
making it cheap , and there is likely to be considerable improvement in the
cost predictions once there is a greater understanding of what needs to be
done technologically .    This will not come about until the next generation of
experiments , like NET in Europe , have been operated .
        The report reviews what has been said in the past about fusion costs
( section 2 ) and describes the sensitivity of generation cost to assumptions in
section 3 . for f irst - of - a - kind tokamaks .  The potential for improving on
present conceptions of what makes a viable reactor is discussed in section M
and fusion is compared with its competitors in section 5 .       A full glossary of
terms and definitions is given in section 8 .
 ---pagebreak---                                                                                      58 .
2.  REVIEW OF PUBLISHED REACTOR COSTS AND COSTING STUDIES
          In the last ten years several conceptual design studies of power
producing fusion reactors or fusion based power stations have been published .
Many of these studies have included estimates of the cost of constructing the
reactors    or  of   generating  electricity ,  and  these  published estimates are
reviewed in the following section .
2.1 Capital costs
         Direct capital costs per unit output for most published commercial
reactor designs are shown in table 2.1 .          The direct capital costs are the
major contributor to the total cost and therefore form a convenient basis for
comparing different designs .      Table 2.1 also shows the relative direct capital
cost of each design normalized to Starfire and adjusted for inflation .                   ( In
the case of the Culham Mk II reactor , the standardized exchange rate defined
by Ashby / 22 / was used to convert the cost to dollars .)
      A number of conclusions may be drawn from the information in the table :
2.1.1   Historical variations
        Early studies such as the Princeton tokamak reactor of 1974 and the
University of Wisconsin tokamak reactors ( UWMAK I and II ) of 1975 , gave lower
direct capital costs than the more recent NUWMAK and Starfire tokamak studies
completed in the period 1979-80 , this being due to the more realistic physics
and engineering bases of the recent studies .
2.1.2   Design uncertainties
      Costs based on recent studies still show considerable variations .            Whilst
the turbine and electrical plant can be costed accurately on the basis of
manufacturing experience , the cost of the fusion reactor itself is uncertain
both because of unresolved physics issues and because of novel manufacturing
requirements .     This is illustrated in table 2.2 which compares the costs of
the reactor plant with the total station cost for some of the power stations
listed in table 2.1 .      The ratio of reactor plant cost to total direct cost
varies from 37$ to 76$ .       Further causes of variation include the effects of
scale ,  and   whether   the reactor    is costed as    the  f irst - of - a- kind or the
 ---pagebreak---                                                                                 59 .
benefits of previous production experience are assumed .             For the above
reasons , comparisons with existing power systems auoh ns fission reactora can
be misleading .
2.1.3    41 ternatives to the DT-toka»afc
        Table 2.1 also show3 estimated direct capital costs for several power
stations based on plasma confinement systems other than the DT-tokamak .             In
general the plasma physios basis for these reactor designs is less well
developed than for the tokaaak . Within the present accuracy , all the costs
are of the same order as for Starfire .
2.1.4    Alternative fuels
       Only one study , Wildcat , has been based on a fuel cycle other than D-T .
This design , based on a D-D fuel cycle , is conceptually similar to Starfire
but requires substantially better plasma confinement in view of the lower
reaction cross-section . As a result the capital oost and cost of electricity
are nearly twice those of Starfire .
2.2 . Cost sensitivity
          Several studies / 23 _ 29 / have  investigated how the cost of a fusion
reactor    varies   with one   or more parameters ,   both  to assess    the  relative
importance of that parameter or to establish its optimum value .       These studies
have utilized both simplified analytical models / 23 , 24 , 25 / which provide
insight     into  the  inter-relationship between parameters ,   and more detailed
computer models / 26 , 27 /.   The main results are as follows
2.2.1    Physics parameters
       The major physics parameters affecting the cost of a tokamak reactor are
the ratio ( 6 ) of the plasma pressure confined to the magnetic pressure
applied , and the plasma aurrent for a given magentic field ( i.e. the inverse
rotational transform of the field lines , q - see glossary ). A plasma pressure
of approaching 10% relative to the toroidal magnetic field pressure is
desirable , but recent predictions of the physical limit are somewhat below
this   level .     A high current for a given field       is essential ,   leading to
requirements for plasma shaping .          By contrast , plasma confinement times
predicted in devices of the scale of a commercial reactor appear adequate .
 ---pagebreak---                                                                              60 .
2.2.2    Engineering parameters
       For unit sizes above 600 MW e , the unit cost of a fusion reactor follows
the two-thirds power law common in engineering production .        Larger units are
therefore more economic , but if too large there may be limits of acceptance .
The first wall power loading has a strong influence on unit costs and there is
an optimum value which is a compromise between the desire to reduce general
reactor material quantities as far as possible , without making the design too
complex or incurring penalties from too frequent maintenance periods .            This
optimum is usually in the range 3 to 6 MW/ m , depending on the predicted life
of the wall before radiation damaged material must be replaced .          In smaller
unit sizes , the total thickness of the blanket and shield on the inboard side
of  toroidal    reactors   significantly  affects   costs  because   it  limits    the
achievable     wall   loading .     The  peak   magnetic   field   achievable     with
superconducting coils ,    or supportable with practical structures ,      is not a
major constraint in a tokamak unless the plasma pressure ratio , 3 is low .
2.2.3   Compact reactors
      One simple way of comparing the economics of alternative power sources is
through the power produced per unit mass of the system .          The cost of many
power   sources   is roughly    related to their mass ,   since variations due to
special materials of complex design do not predominate , and for this reason
compact systems are economically attractive .         For fusion reactors a rough
target for the mass power density of 100 kWg/ tonne has been suggested / 30/,
and several designs of compact reactors exist approaching       this value as shown
in figure 2.1 / 31 /.   In this respect the Reversed Field Pinch has an advantage
because of its high plasma pressure ratio (3 - 25% ), whereas for tokamaks only
designs with non-superconducting magnets to allow high-field operation can
approach this mass power density .         This question is considered again in
section 3-
        As already indicated in table 2.2 the ratio of reactor plant cost to
total direct cost is significantly higher for a fusion reactor than for a PWR .
Figure 2.2 shows a correlation between this ratio and the unit capital cost ,
which suggests that the estimated capital cost of a fusion reactor should be
reduced by a factor 2 to compete with a present day PWR . This reduction
corresponds to a factor 4 in mass utilization . These conclusions , however ,
take no account of the low fuel costs of fusion which may considerably reduce
thes^ factors .
 ---pagebreak---                                                                                  61 .
2.3  Electricity generating costs
      in several, studies the direct capital costs have been used as the basis
of generating cost estimates , as quoted in table 2.3 .        These are dealt with
more fully In section 5 .
2.4  Energy accounting
        An alternative to considering the electricity generating costs is to
calculate the energy expended in all the processes which are involved in the
manufacture ,     construction  and   operation   of   the  system .     This  energy
expenditure includes mining and refining the raw materials - including the
fuels - as well as the production , transport , and erection of the plant and
buildings .      One advantage of energy accounting is that        it should not be
influenced    by relative    wage  and  price   changes .   Another   very  important
advantage in relation to energy accounting for power stations is that the
ratio of energy expended to the energy generated during the life of the
station is an easily understood and convenient measure of the value of the
project .    Th'e major difficulty in the assessment is the calculation of the
energy expenditure in each activity , which is often poorly defined and is in a
Variety Of different forms . Conversely the payback time , in spite of being
widely    used ,   is a misleading measure    because    it is highly   sensitive   to
arbitrary assumptions in its definition .
      Some results of a recent detailed study by BUnde / 32 , 33 . 34 /, in which
two fusion power plants were compared with two LWR fission reactor power
plants , are given in table 2.4 . The energy expenditure on construction of a
fusion power station is a factor of two greater than that for a PWR station ,
which is consistent with capital cost estimates . The overall energy input for
the fission station , however , is significantly increased by the energy
required to provide fuel both for the start of operation and for life-time
refuelling .      the figures quoted in table 2.4 are the most optimistic for
fission and the most pessimistic for fusion of the cases considered .               An
earlier study by Tboulfanidis / 35 / gave similar results , shown in table 2.5 ,
but it may be hoted that the fusion energy inputs were calculated on the basis
of the UtiMftK-III which is been in table 2.1 to be the most expensive of the
American tokamak reactor designs .
 ---pagebreak---                                                                                 62 .
2.5   Discussion
       In discussing the existing literature of fusion economics it must firstly
be stated that all cost estimates are based on outline designs which assume
favourable solutions to outstanding physics questions . Whilst the cost of
individual components can be estimated from other engineering applications ,
not all details of the components are known , and so the costs quoted here are
only the best possible indications at the present state of fusion development .
By comparison ,     other   energy   systems   such as   fission   reactor  based   power
stations are well defined and can be much more accurately costed , although
still dependent on financial assumptions and resource availabilities .
      Sensitivity studies have allowed present reactor designs to be optimised ,
within the constraints of present understanding .          The extent to which changes
in parameters could lead to lower capital costs is well understood .            In terms
of physical limitations , the plasma pressure ratio 3 is most important .              In
terms of engineering constraints ,         any factor which permits a higher power
density will be important .          Present designs are therefore tending to more
compact reactors , with increased emphasis on materials properties and high
magnetic fields .
      There have been very few new commercial tokamak reactor design studies in
the   past   five   years ,  not   only  because of    the   present  emphasis on next
generation devices      such as NET or       INTOR , but  because   there have been no
significant changes in physics understanding since the Starfire study which
would    change   the  engineering concept      and hence    the  estimated  cost .    In
contrast to the tokamak situation , there have been several recent studies of
reactors based on other confinement geometries .           Of these , the tandem mirror
( MARS )  study suggests that there is no obvious economic advantage .                The
Reversed Field Pinch , however , has the potential to be the basis of a more
compact ,   and hence cheaper ,     reactor but has a weaker physics basis .          The
stellarator has been the basis of several studies , which indicate costs in the
same range as for the tokamak .
      This viewpoint has not covered inertially based reactor systems / e.g. 20 ,
21 /, for which much of the target physics is classified information and for
which the cost of the driver systems is very uncertain .             Nor has it covered
fission-fusion hybrid systems / e.g. 36 / for which reactor designs are less
well developed , and costs depend to a large extent on the value of the fissile
fuel produced and on the cost of safety for this complicated system .
 ---pagebreak---                                                                                  63
                      TABLE 2.1 : SUMMARY OF REACTOR STUDIES
                                                               Specific             Relative
Year of     Year of MW       Name                              Direct               capital
                       e
publication costing net                                        Capital              cost
                                                               cost                  ( corrected
                                                               ( $ / kW e )            for
                                                               ( in year             inflation )
                                                               of costing )
                      DT-Tokamaks :
1974        1974    2030     PPLP / 1 /                                      433    0.4 )
1 97b       1974    1474     UWMAK I / 2 /                                   723    0.78
1975        1975    1709     UWMAK-II / 3 /                                  706    0.69
1976        1975    1985     UWMAK-III / 4 /                                1154    1.14
1976        1976    2500     Culham I / 5 /                                  750    0.70
1979        1978      660    NUWMAK / b /                                   1279    1 . 05
1980        1977    1200     Culham II B / 7,8,9 /                          1442    1 . 28
1980        1980    1200     Starfire / 10 /                                1439    1
                      Others =
1978        1976      492    Standard mirror / 1 1 /                        4510    4.22
1979        1979      750    RFPR ( Reversed field pinch )     712 /        1104    0.84
1980        1980    1530     WITAMIR ( Tandem mirror ) / 1 3 /              1348    0.94
1981        1980      812    Wildcat ( D-D tokamak ) / 1 4 /                2725    1 . 89
1981        1981    121 4    EBTR ( Bumpy torus ) /1 5 /                    1737    1.14
1982        1982    1882     UWTOR-M ( Stellarator ) /1 6 /                 1422    0.88
1983        1980    1 660    MRS-IIA ( Stellarator ) /1 7 /                 1482    1 . 03
1983        1980    1302     MRS-IIB ( Stellarator ) / 1 7 /                1265    0.88
1984        1980    1200     MARS ( Tandem mirror ) / 1 8 /                 1970    1 . 37
1985        1980    1000     CRFPR 20 ( Compact RFP ) /1 9 /                1111    0.77
 1985       1984    3784     Hiball II ( Heavy-ion beam ) 720,21 / 1347             0.74
 ---pagebreak---                                                                  64
              TABLE 2.2 ;: REACTOR PLANT COSTS
           Reactor          Direct    Total    Ratio       Ratio
            ( $M )          capital   Capital  Reactor /   Dir . cap ./
                               ( $M )  ( $M )  Dir . cap . Total cap .
PPPL        606                880     1215     0.69        0.72
UWMAK-I     574              1 066     1433     0.54        0.74
UWMAK-II    775              1207      1615     0.64        0.75
UWMAK-III   812              2290               0.35
NUWMAK      534                844     1140     0.63        0.74
Starf ire   969              1727      2400     0.56        0.72
Culham IIB  656                911     1824     0.72        0.50
RFPR        397                828              0.48
WITAMIR    1565              2063      2785     0.76        0.74
Wildcat    1497              2213      3076     0.68        0.72
MRS-IIA    1687              2460      3695     0.69        0.67
MRS-IIB     968              1647      2473     0.59        0.67
EBTR       1426              2109      2872     0.68        0.73
UWTOR-M    1765              261 1     3758     0.68        0.69
MARS       1517              2365      3266     0.64        0.72
CRFPR.20 .  415              1112      1515     0.37        0.73
PWR                                            0.25-0.32
 ---pagebreak---                                                                                 65 .
                  TABLE 2.3 : COST OF ELECTRICITY - ( mills-1 980 / kWh )
                               Starf ire   CRFPR.20        Mars
Annual capital charge           30.44       22.79          42.56
Operation and maintenance        2.46        4.11            2.63
Component replacement            2.20        1 . 00         0.69
Fuel                             0.04        0.03           0.36
Total                           35.15       27.93          46.24
The annual capital charge is set at 10% of the total capital cost , in constant ( zero
inflation ) money over a 30 year operating life . Plant availability is different in
each study ( between 75-80% ).
 ---pagebreak---       TABLE 2 . -4 : ENERGY INPUT AND OUTPUT OVER 30 YEAR LIFE ( from ref 34 )
                                                            Fusion     Fission
Construction of power plant             ( MWh,th./MW e ) +   4082       2160
Construction of fuel installations      ( MWh,th. /MW e ) +    16        789
                                                                             *
Fuel for first operation                <"“hth/HWe /             3       399
                                                                             *
Fuel for lifetime operation             ( MWh th /MWe )*       87       5554
Total energy input                      (MWh th
                                              , . /MW e ) +  4188       8902
Energy generated                        (MWh th
                                              . . /MW e ) + 6.3x1 0^    6.3x10
Energy gain                                                   150         70
#
   Assuming centrifuge enrichment of ore with a 0.2% uranium content .
+
   MWh^^ always means thermal energy and/ or primary energy equivalent of
   electrical energy , and MWg refers to electrical power sent out .
 ---pagebreak---               TABLE 2.5 : ENERGY GAINS FOR POWER PLANTS ( from Ref 35 )
                                              EG 1     EG2       EG3
Coal Plant                                    5-7      6-9       53-93
PWR ( diffusion enrichment )                  3-5      7-5       15
P WR ( centrifuge enrichment )                10       13        80
Fusion plant                                   5        7        64
EG1 = Electrical energy out /equivalent thermal energy in .
EG2 = Electrical energy out / total energy in .
        Electrical energy out / electrical energy in .
 ---pagebreak---                                                                                          68 .
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                                           FUSION POWER CORE
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FIGURE 2.1 Specific direct capital cost as a function of mass utilisation in
            the fusion power core ( from reference 30-
                        -1-1-1-1-r -1-1-r
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                            REACTOR PLANT EQUIPMENT ( RPE )
                                TOTAL DIRECT tOST ( TDC )
FIGURE  2.2   Specific direct capital cost as a function of the cost ratio
              between reactor plant equipment and total direct cost ( from
              reference 31 •
 ---pagebreak---                                                                               69 .
3.   GENERATION COST SENSITIVITY
        As pointed out in the previous section there have been very few recent
assessments of commercial reactor's because of the present emphasis on the next
step in the programme of development .        As part of this work in Europe , an
extensive model of the cost scaling of reactor systems is under development as
a design aid in the choice of NET parameters .        This model has been built up
using the expertise gained in the studies reported in section 2 and has now
been extensively reviewed by Motor Columbus Engineers Inc .           who have wide
experience of power plant construction worldwide .       Modifications suggested by
them have been incorporated in the model as it stands today / 37 /, and it has
been    extended   to  analyse   electricity  generation   costs   along   the     lines
recommended in the UNIPEDE study / 38 /.
       This model is used here as the basis for describing the cost sensitivity
of reactor parameters , since it represents the latest , and therefore hopefully
the most accurate , assessment within Europe of reactor costs for first - of - a-
kind , DT-based tokamaks .     As such , the results reported below should not be
taken to be indicative of reactor costs in a mature industry .           In any case ,
extrapolation of currently perceived NET design solutions into the commercial
reactor regime has low credibility since NET itself will be the test bed for
developing such reactor relevant design solutions .       Inevitably , in all areas ,
both learning in manufacture and improvement in design will also drive costs
down in future devices from levels predicted today .        Furthermore , within the
present    modelling , no  attempt  has  been  made to minimise     non-direct     costs
( operation and maintenance especially ) to increase commercial acceptability ,
and this results in a further overestimation of fusion costs .
3.1   Generation Cost Usage
       One of the advantages claimed by fusion is that it has low fuel costs to
offset against probably high capital costs .        When comparing the merits of
fusion with its competitors it is therefore essential to consider all costs
incurred from the start of construction to ultimate decommissioning when making
a judgement .    This can only be done by the use of generation costs ( G ), also
known as cost of electricity , which properly account for the influence of
capital , operating and maintenance , fuel , decommissioning and interest charges .
The assumptions implicit in the costs reported here are listed in table 3-1 .
Only direct , operation and maintenance , and fuel costs are calculated in
detail , with other non-direct costs amounting to 58$ of D.
 ---pagebreak---                                                                                70
3-2    Generation Cost vs. Beta Level and Mass Expenditure
        The plasma pressure ratio , B , can be related to basic Tokamak parameters
by the equation B($ ) = gI(MA) /a(m)B(T ) where I , a and B are plasma current ,
minor radius and toroidal field respectively and g is a constant known as the
" beta level ".     To minimise the amount of plasma needed for a given output
power ,   B and hence g must be maximised ,       particularly since  its square is
proportional to the plasma power density .       One of the major efforts in fusion
is therefore to maximise the beta level subject to any other constraints that
might apply .
          For a device of fixed power sent out and beta level there exist an
infinite number of possible designs with different dimensions .       A minimum cost
device can be chosen from this infinite set .       The variation in generation cost
of such minimum cost devices can then be shown as a function of the power sent
out and beta level .      This is done here using parameters predicted by SUPERCOIL
/ 39 / over a wide range of values of power sent out and beta level .            This
analysis / 40 / extends an earlier analysis based on the capital cost only / 41 /.
Figure 3-1     shows the results , relative to the cost of one particular design
point ( the reference point , PCSR-E ( prototype commercial-sized reactor ), is a
1200 MW SO device with a value of g ( 3-5 ) consistent with present day
experiments ), indicating a decreasing cost benefit as both beta level and power
sent out are raised but that certain minimum levels of these parameters are
worthwhile attaining .      Also shown is the wall loading that should be achieved
to gain access to the cost minima at each value of power sent out and beta
level .    ( In reality , since cost minima are fairly flat as a function of wall
loading ,    small  reductions   in  wall  loading  from  the  values  shown may   be
tolerated without much cost increase ).
         Under the stimulus of studies recently carried out in the USA / 30 / the
same results are replotted in figure 3.2 as a function of " mass expenditure "
( ME ) on the fusion power core ( FPC ), i.e. the mass of material required for the
torus ( first wall /blanket/ shield ), magnets ( toroidal and poloidal field ) and
their respective support structures ,       divided by the power sent out .     This
variable is equal to the " mass utilization " multiplied by the overall plant
efficiency ( typically 30% ) , and is inversely proportional to the mass power
density ( 100 kWg/ tonne = 10 tonnes/MWg ), both these terms having been mentioned
in section 2 .     Figure 3.2 also shows absolute generation cost values for these
first-of- a- kind stations in 1984 ECU (1 ECU-1984 - 0.822$-1984 ).
 ---pagebreak---                                                                               71 .
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FIGURE 3-1    : Generation cost of minimum-cost devices as a function of beta
                level at different values of power sent out , and the corresponding
                wall loading levels required .
 ---pagebreak---                                                                                            72 .
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                                                               FUSION POWER
                                                          C0RE CONTRIBUTION
                0 -^-1-1-1-1-'- 0
                  0        10        20      30         40      50      60    70
                           MASS EXPENDITURE I tonnes / MW.„)       MWJ
FIGURE 3.2 : Correlation between generating cost , neutron wall loading and
                mass expenditure for minimum-cost devices at given values of g
                and power sent out
 ---pagebreak---                                                                                       73 .
        The most striking features of figure 3.2 are the direct proportionality
between generating cost and mass expenditure and the wide range of cost that
can    occur   with    different    assumptions     about   g   and   P so .    ( The    direct
proportionality would have been distorted somewhat if availability had been
related to wall loading but this was not thought reasonable to do here since a
utility will prescribe a desirable availability , like that shown in table 3 • 1 »
and all design solutions must , satisfy it ).
           The   results   of   figure    3.2  show   that  FPC   cost   curves   are    almoat
superimposed     indicating    the   strong   dependence of its costing on mass .             A
typical unit cost is around 50 ECU / kg and this is independent of P 30 and g .
However , the accessible range of values of ME varies considerably with g and
P so . Although it only -     directly contributes about 1 5-35% to the direct costs
( 30$ for PSCR-E ), the FPC has an indirect effect on the rest of the plant .
This can be seen by the direct cost contribution curves which have now become
separated , since costs depending on power sent out , and fixed costs , have been
added in .     However , the change in slope of the curve indicates a " knock-on"
effect of FPC mass , which occurs mainly via the building costs since , under
present assumptions , building size is strongly related to FPC dimensions .                 The
FPC thus influences 50-80$ of the direct costs ( 71$ for PCSR-E ).              Furthermore ,
at least 60$ of non-direct costs depend on direct costs and this produces the
further amplifying effect on the slopes of the lines shown in the generating
cost curves .     The FPC then influences between ^ 0-75$ of the generating cost
( 65$ for PCSR-E ) although it only directly contributes 8-18$ ( or 13-20$ if
first wall and blanket replacements are included ).
        These results show the strong influence of the FPC on costs .                 However ,
this is partly a figment of the cost models used at present and is strongly
affected by items not usually considered in the fusion programme ( e.g. building
design for fusion plants ). This , combined with the strong variation in costs
that can be achieved with improved physics attainment , represented here by ' g' ,
makes costing of fusion reactors at this stage , highly speculative .
3-3   Directions for Improvement
       The above results do not indicate any hard target for the competitiveness
of fusion , such as the 100 kWg/tonne mentioned in section 2 , although any
improvement     which   lowers    mass   expenditure    may   result   in    a reduction     in
generation cost .     At present all that can be said is that there is considerable
 ---pagebreak---                                                                                   74 .
uncertainty in costs of DT tokamak fusion caused by the lack of knowledge of
the physics and technology particularly of the KF’C in a reactor . Despite this ,
current estimates of the absolute costs , shown in figure 3*2 , indicate that the
PCSR-E design point would be rather expensive as an end point of the tokamak
development programme .    It is therefore worthwhile to speculate how the cost of
the end point device would be affected by future developments .
3.3.1   Direct cost réduction
      To accomplish this , inherently cheaper technological solutions than those
proposed for the engineering design problems of NET would have to be found .           In
the present PCSR-E design , the major direct cost items / 42 / are the fusion
power core ( 30$ ), buildings ( 19$ ) and the cooling / generating system ( 12$ ). The
latter two items have not yet been optimised even for NET , so it is reasonable
to expect considerable improvements by the time commercial reactors are being
designed .    For the fusion power cor *;, magnet costs , which are strongly driven
by specific conductor costs , make up more than half the total .        A significant
reduction of these specific costs under the mass-production of superconducting
cable needed for fusion reactors is therefore to be expected , irrespective of
any cheaper design solutions that may be         implemented .    As a guideline ,      a
generation cost reduction of 15$ ( without change in mass expenditure ) can be
achieved by reducing specific costs of all items in only the FPC by 50$ .
3.3*2   Improved plasma physics at constant power sent out
      This is represented here by the factor g .       A 15$ reduction in generating
costs is achievable with a 60$ increase in g .        A consequent 20$ reduction in
mass expenditure occurs due to this Increase in compactness .       This approach has
its limitations , however , as g has to be doubled again to reduce costs by a
further 15$ .    However , these calculations have been carried out using a fixed
plasma configuration , and innovations in this area ( see section 4 ) which
improve the plasma beta at constant g and which have the advantage of making
the device more compact , may , despite possible extra costs due to the use of
more exotic configurations , have a beneficial effect overall on cost .
3-3*3   Raised P 30 without g increase
           Increasing compactness    is not  the only method of decreasing mass
expenditure .    A 15$ reduction in generating cost would be achieved by a 40$
 ---pagebreak---                                                                              75 .
increase in power sent out without increasing g , as shown in figure 3*1 •        The
corresponding mass expenditure decrease would be 16% . However , this increased
power sent out would have to be acceptable to the utilities . Here there are
differences , with , for instance , 1500 MWgo becoming the new European standard ,
whereas in the USA , 300-600 MWgQ units are thought to be more desirable for
their future energy needs .
3 . 1* Sensitivity to Assumptions
        In producing the results quoted here , certain basic assumptions have been
made .    The sensitivity of the cost of PCSR-E to changes in these assumptions is
shown in table 3*2 for the most sensitive parameters .          The sensitivity is
defined as the relative change in the costs , divided by a given relative change
in the parameter ,     all other parameters in the table remaining fixed .        The
sensitivity is quoted relative to that for variations         in g .  Three plasma
physics parameters head the list and they are not really independent ( as
assumed in the sensitivity analysis ) since g and q depend on the radial
profiles     of  plasma   density and  temperature  in  a way   which can  only    be
determined after extensive experimentation on reactor-level plasmas .         These
profiles are      implicitly included  in f which is also a function of plasma
operating temperature .
       Stress levels in the toroidal field coils are less important .    The use of
better quality materials in superconducting coil manufacture may ease this
limit towards higher values ,      but many superconducting materials are strain
limited and this may provide a nearby limit .      Also blanket thickness is not a
major cost driver .       This is fortunate since adequate space must always be
allowed for tritium breeding .
3*5    Discussion
        The results given above indicate that generating cost must be used with
extreme caution as a measure of the future worth of fusion power from DT-driven
tokamaks as it strongly depends on the FPC cost , which is poorly known at this
stage . It is therefore too early to draw hard and fast conclusions from this
analysis and such conclusions must wait until more is known about reactor
design solutions and their technology , that is , at the end of operation of NET .
 ---pagebreak---                                                                             76 .
       Even though generating cost values are uncertain , it is apparent that
factors of 2 can result from future research and development activities . There
appears to be a benefit in systems which either reduce mass expenditure , by
possessing higher g and / or operating at increased levels of power sent out , or
reduce fusion power core costs by the use of cheaper design solutions .          This
clearly points the direction for future development but the strength of the
incentive cannot yet be clearly quantified .  It must also be remembered that in
a mature fusion economy , learning will significantly reduce costs / 1 0 / over the
absolute values shown here .
        However , before fusion can be   introduced on a large scale ,     the cost
difference between fusion and its competitors must be small or even negative .
That fusion has the development potential to accomplish this is demonstrated in
the following section .
 ---pagebreak---                                                                       77
        TABLE 3.1 : LEVELISED GENERATION COST ASSUMPTIONS
Plant lifetime                                  25 years
Availability - Year 1                           4000 hours / yr
                Year 2                          5000 hours / yr
                Year 3-25                       6600 hours / yr
Discount rate                                   5$
Indirect costs                                  29 $ of D
Interest during construction                    23$ of D
Decommissioning costs                           20$ of D , discounted
         TABLE 3-2 : SUMMARY OF MOST SENSITIVE PARAMETERS
                                                           Relative
Parameter                           Value                  Sensitivity
Beta level , g                       3.5                        - 1.0
Inverse rotational transform q       2.2                          0.8
Fusion power density ratio , f       1.5                        - 0.5
Blanket thickness                    0.55 / 0.85 m                0.3
Toroidal field stress level          160 MN /m2                 -0.2
 ---pagebreak---                                                                                      78 .
4.    DEVELOPMENT POTENTIAL FOR FUSION
           The present fusion programmes world-wide are scientific programmes
orientated towards solving problems of principle . In the past , the programmes
concentrated on physics questions because the largest hurdle to be overcome was
seen there but , as a consequence of the progress made in physics , a gradual
transition has been taking place for some years now to increasingly include
questions of technology as well .
        The target of the programmes is a demonstration reactor to prove by its
successful operation that working solutions have been found for all problem
areas .    However , these solutions , if applied without any further improvement ,
would result     in a commercial reactor more costly than perhaps necessary .
Therefore the demonstration of basic feasibility has to be followed by a period
of technical improvement ( i.e. innovation and simplification of the design ) to
arrive at a desirable and economically competitive end product .              Such a step ¬
wise .procedure is advisable , especially since many of the expected improvements
at the reactor level would have no or only negligible impact on present-day
experiments .
         In order to substantiate this argument , an activity on reactor concept
innovations was started within the INTOR frame and the first results will be
reported here .
4.1    Reactor concept innovations
       At the request of the IFRC ( International Fusion Research Council ) an IAEA
Specialists' Meeting was held on 1 3 _ 1 7 January 1986 at Agency headquarters in
Vienna / 43 /.    The purpose of this meeting was to identify innovations that
would significantly improve the prospects that fusion reactor development would
lead to an attractive end product - a viable and economically competitive
fusion reactor , and to limit the initial activity to the Tokamak concept .               A
worldwide call for innovative proposals was made prior to the meeting via the
INTOR Workshop .     About 120 proposals on innovations were received and underwent
a first analysis .     They were nearly equally distributed among nine categories :
( i ) impurity control , ( ii ) beta and confinement enhancement , ( iii ) heating and
current drive , ( iv ) advanced magnets , ( v ) plasma engineering , ( vi ) configuration
and maintenance , ( vii ) advanced blankets / first walls / shields , ( viii ) advanced
materials , and ( ix ) innovative concepts .       Categories ( i ) to ( iii ) are in the
 ---pagebreak---                                                                                   79
physics field , and ( iv)-(viii ) in the field of engineering . As expected from
the early concentration of the fusion programme on physics questions , the
physics innovations mainly consisted of anticipated results of present
activities     promising     plasma  conditions    suitable   for   reactor   application ,
whereas     many    of    the   engineering    innovations    were    orientated    towards
improvements     of   the   end  product with    no essential     impact  on the    present
generation of experiments .        This will become apparent from the results of the
Workshop summarized in the next section .
4.2   Results of the Workshop on Reactor Concept Innovations
4.2.1    General
       By combination of a large number of the proposed innovations , substantial
improvements seem to be possible , even if the single ones alone might only
produce moderate effects .       This conclusion holds even if some of the proposals
in the end would turn out not            to be feasible .      Furthermore , many of the
proposals are not restricted to Tokamaks but applicable to toroidal magnetic
confinement in general .
4.2.2    Increase in plasma power density
        There were a considerable number of proposals aiming at increasing the
plasma    power   density .     They range   from using     indentation and the second
stability regime , to increasing the magnetic field by using advanced super ¬
conductors allowing both higher field and higher current density , and they also
include sophisticated feedback circuits to improve plasma stability .                  Here
combination looks promising .         If all of them work it is expected that the
limitation in power density will then be set by the acceptable wall load .
4.2.3 Plasma heating
          Compared to the presently used systems , high energy ( about 0.5 MeV )
neutral beam injection should allow the beam power density to be increased by
an order of magnitude above that of today 's systems and , simultaneously , the
distance between beam sources and plasma to be increased to 30 m or so ( high
beam collimation ) .       This should not only allow the blanket coverage to be
increased but also the beam sources to be put into regions with nearly no
neutron irradiation .        In addition , these beams could perhaps also be used for
 ---pagebreak---                                                                            80 .
active impurity control and current drive .     Present plasmas are too small in
cross-section for such beams to be applicable .
4.2.4  Trends
       After having discussed the proposals on advanced Tokamak concepts , the
Workshop recommend    to put  emphasis on  improving upon   the present  line of
moderate elongation , moderate aspect ratio configurations rather than switching
over to very elongated or very low aspect ratio configurations .
4.2.5 System Aspects
         There was one proposal of potentially high influence on the reactor
concept . It exploits the extremely high plasma temperature ( above 100 million
degrees ) unique to fusion power by replacing the usual balance of plant by in-
situ MHD power conversion .    MHD circuits are introduced directly behind the
blanket such that the toroidal magnetic field existing anyway can be used for
the MHD process . The plasma electron temperature will be raised to above 30 keV
so that half the alpha power will be converted into synchrotron radiation which
will be used to create the necessary non-equilibrium ionization within the MHD
medium at acceptable operating temperature .   By this method the neutron energy
could be absorbed by high ( but still manageable ) temperature pebble beds and
then exploited by the MHD process . This proposal claims considerable savings
in the balance of the plant .      The concept is also applicable to magnetic
confinement in general and not restricted to Tokamaks .
4.2.6 Summary on reactor concept innovations
          The Workshop has clearly shown that there are enough ideas for
significantly improving the end product above previous perceptions . Nearly one
half of the proposals received were selected for deeper studies on their
prospects of final feasibility .       This provides a large potential for
substantial improvements .
4.3  Stellarators and Reversed Field Pinches
      In Europe it was concluded at a very early stage that toroidal magnetic
confinement offered the best chance of leading to a viable fusion reactor , and
practically all the European fusion effort was concentrated on this class of
 ---pagebreak--- systems with the Tokamak being the main approach .            Therefore , the above
sections dealt with the prospects of the Tokamak as the ultimate fusion reactor
concept .    There are , however , substantial possibilities of improving on the
Tokamak where it encounters difficulties in its physics and engineering .
Stellarators and Reversed Field Pinches are being developed in Europe with
these prospects in mind .       According to European plans the concept selection
will be made after NET operation .
          The Stellarator line of magnetic confinement uses external electric
currents to produce the magnetic field in which a ring of plasma is passively
contained .    The successful operation of the Wendelstein Stellarators and of a
few other machines in other countries have made the Stellarator line a very
serious   contender  with   the Tokamak as  the basis for a future fusion reactor .
The transfer of the Tokamak plasma current into external coil currents for
producing the necessary poloidal field components allows the Stellarator to
work with only one single coil system , to dispense with any transformer or
current    drive  system ,  to  be free of disruptions ,  and  to use   steady-state
operation as an inherent property .      Once ignited it works by re-fuelling and
exhaust alone .     Present work aims at establishing beta values predicted by
theory and solving the impurity problem .
     Reversed Field Pinches , on the other hand , use plasma currents higher than
those of a Tokamak .      The magnetic field configuration produced in this way is
expected to relax into a minimum-energy state promising very high values of the
plasma pressure stably confined by the RFP fields .         Experiments in Culham ,
Padua , and elsewhere in the world have shown that the basic processes work .
This concept offers the advantage of arriving at the burning state by ohmic
heating alone .     Present work aims at establishing the RFP configuration at
higher plasma parameters and at reducing the transport losses to acceptable
values .
 ---pagebreak---                                                                                         82 .
5.   COMPARISON WITH OTHER POWER SYSTEMS
          If fusion     power   is   to be    introduced on a large scale         it must be
competitive with baseload generating technologies .                Today these technologies
are the conventional coal-fired and nuclear thermal power stations .                   By the
mid-21st    Century when      nuclear     fusion   can  be   expected   to  be   commercially
available ,    fast breeder nuclear power and solar photovoltaic conversion are
also likely to have reached commercial maturity .
5. 1  Comparison validity
          It   could be argued       that   coal-fired   plants and nuclear      plants will
undoubtedly change in many ways during the next 50 years or so , making any
reference     to   their   present     state    irrelevant .     However ,  some   long   term
tendencies of these changes can be inferred :
         - For coal-fired plants ,         increasingly difficult exploitation of fuel
resources and the strengthening of anti-pollution standards will lead to higher
prices .     In addition , worries about the increase in atmospheric CO^ could
curtail the use of fossil-fuels in power generation .
         - For thermal fission reactors , a number of technological changes are
still possible .     Higher fuel utilisation would be particularly stimulated by an
increase of the uranium ore price .
       In the long term , the uranium price will undoubtedly increase , although
neither    the   time scale nor       the   slope of this      increase  is  known and    they
obviously depend on the worldwide development of nuclear energy .                    With the
present state of the art , multiplying the price of fuel by a factor of 10
induces a factor of about 2 in the generating cost of thermal fission reactors .
      Other types of reactors , like the HTR with a thorium cycle , or molten salt
reactors ,   could also appear        in the meantime .      In the case of fast breeder
reactors , the investment cost of the French Superphenix plant is about twice
the price of a French PWR .        This is expected to reduce significantly for future
commercial fast breeder reactor plants / ^ ^ , A5 /.
          The above uncertainties          indicate the difficulty in telling in what
direction and to what extent the present price of nuclear energy will change
half a century ahead .        Therefore comparisons of fusion with present costs of
these systems can only give guidance , since it must be remembered that the
 ---pagebreak---                                                                                     83 .
prion of    prenont day    systems may    increase   cons idnrabi y over   the timescale
envisaged for the introduction of fusion .
5.2  Non-quantif ied économie characteristics
         There are a number of somewhat intangible but potentially beneficial
effects    of   an   electricity  generation     network    with   fusion   as   a   major
constituent .    These include :
        - Security of fuel availability .          Deuterium and     lithium are spread
widely and plentifully , a guarantee against a geopolitical crisis .
        - Low fuel price dependence allows even low fuel-content resources to be
exploited and ,    in the very long term , keeps at a low level the influence on
generation costs of fuel price escalation .
        - The fuel cycle is internal to the power plant , so the fuel supply does
not depend in principle on extensive off-site reprocessing systems and their
associated logistics . Even if recycling of lithium proves to be desirable from
an economic standpoint ,     this is much less expensive and hazardous than with
fission .
            Without    the  need  for    fuel  reprocessing     there   is   considerable
difficulty    in   the  diversion of   materials    for  the   construction   of   nuclear
weapons without detection .
        - Opportunity for reduced waste hazard by developing low activation
materials ( materials presently proposed are optimised for use in fission ),
leading to a lower impact on society .
        - The reduced scale of possible accidents .
5.3  Quantitative cost comparison
        Generation cost has been used in several studies by the OECD /Nuclear
Energy Agency / 46 , 47 /, UNIPEDE / 38 /, and in national comparisons of coal-fired
and nuclear generation of electricity . These results are shown in table 5.1 ,
and transferred to 1984 US $ for comparison with the other technologies .
      The generation costs of nuclear fission and coal-fired power stations are
illustrated by appropriate high and low estimates for the different generating
cost components taken from the OECD / NEA reports .           The fuel costs , however ,
include price escalations within the time horizon ( 2020 ).                 The cost of
electricity from fast breeder reactors must be within the cost range for coal
and thermal fission , if this technology is to penetrate the market on a large
scale , so this is not included in the tab In .
 ---pagebreak---                                                                                 84 .
         Solar energy appears to be a possible challenger of fusion in the middle
of the next century , at least in Southern Europe .      Two processes are currently
under development : thermodynamic cycles ( with mirrors and boilers ) and direct
conversion ( photovoltaic cells ). The probability that thermodynamic cycles can
be a valuable long term solution is limited , considering its vulnerability to
weathering .       The prospects are better for direct conversion .      The price of
direct conversion is sensitive to cost and efficiency of photovoltaic cells ,
for which significant improvements are possible .      However , even if zero cost is
assumed       for   photovoltaic   cells  and  several  values      taken   for   their
efficiencies , the minimum generation cost is still about 20 mills / kWh / 48 /.
Two solar photovoltaic generation studies with realistic prices for the cells
/ 50 ,   51 / are quoted in table 5.1     and they quantify expected reductions of
investment costs .       No estimates are made for operation and maintenance cost ,
these being considered negligible .
          The basic conclusion that can be drawn from table 5.1       is that all the
estimates are of the same order of magnitude , and that the numerical values of
the cost ranges of these technologies are overlapping .
         The most recent estimate of fusion power costs , PCSR-E , which is a first -
of - a- kind study and does not assume improvements beyond the present physics
base , shows costs that are three times higher than those of the Starfire study
from 1980 , which was a tenth- of- a- kind study .       Under learning assumptions
typically assumed for Starfire , cost reductions of between 30 and 50% over
first-off costs are readily obtainable .        Fission costs that are estimated on
uniform assumptions show a range from 19 to 53 mills 1984 per kWh , which has a
significant overlap with the 29 to 86 mills per kWh range for fusion . Since
any cost       calculation so   far ahead  in the future  is bound to be extremely
uncertain ,     this should not necessarily lead to the conclusion at this stage
that the one will be eventually more expensive than the other .
        Within the calculated cost range of these technologies that already exist ,
namely coal and thermal fission , ranging from 20 to 80 mills-1984 per kWh , it
seems likely that both nuclear fusion and solar photovoltaic will be able to
penetrate in the future as large-scale generating technologies .
 ---pagebreak---                                                                                      85 .
TABLE 5.1       : ESTIMATES OF ELECTRICITY GENERATION COSTS IN MILLS-1 984 / kWh
                  BY MID 21st CENTURY FOR LARGE SCALE BASE LOAD TECHNOLOGIES
                                                                       y
         Discount rate 5%                             Invest        0&M     Fuel     Total
Fusion                                0
   Starfire ( tenth of a kind )“"                      25.9          3.3     0.0       29.2
   CRFPR.20 ( not first of a kind ) ¿                  19.4          6.1     0.0       24.5
   MARS ( tenth of a kind ) ¿                          36.2          4.0     0.5       40.7
   PCSR-E ( first of a kindK                           70.6         15.0     0.7       86.4
Thermal Fission                             ,
   OECD /NEA low estimates ( France )                  10            4       5         19
   0ECD /NEA high estimates ( USAT                     32            5      16         53
Coal                                      ,
   OECD/NEA low estimates ( Italyl 0                     6.9         2.8    24.6       34.4
   OECD /NEA high estimates ( USA)'                    14.0          4.8    63.2       82.0
                          g
Solar photovoltaic
   USDOE Price Goal 1990
   ( 1 . 1 0$-1 980 /W ) Northern Europe               89                              89
                         Southern Europe               54                              54
   EC Study
   (2 ECU-1 980 /W ) Northern Europe                  164                             164
                         Southern Europe               98                              98
Notes
1.       $ 1 984 = 0.833 $ 1980 = 1.21 ECU 1984
2        As in section 2 but assuming annual capital charge 7.1 ? ( interest 5% / year ,
         lifetime 25 years ) instead of 10% .
3        As in section 3
4.       French investment and O&M costs plus parameters of once-through nuclear
         fuel    cycle giving     lowest    fuel costs ;    no escalation  in uranium price
         ($ 32/lb U30g ) / 46 , 47/.
5.       Central US .      investment   and O&M costs       plus parameters of once-through
         nuclear fuel cycle giving highest fuel costs ; uranium price escalation 4%
         p.a . from 1995 to 2020 ($ 85 / lb U 3o0 »a ) / 46 , 47 /.
6.       Italian investment & 0&M costs plus coal price after 2020 2.4 $ / GJ / 46 ,
         47 /.
7.       Central U.S investment and O&M costs plus German indigenous coal , coal -
         price after 2020 4.7 $ / GJ / 46 , 47 /.
8.       Annual capital charge 7-1% ( interest rate 5% / year , lifetime 25 years )..
         Load factor for Denmark 0.12 , for southern Italy 0.2 / 49 , 50 , 51 /.
 ---pagebreak---                                                                                     86
5. 4   Criticism of the economic potential of fusion
       In parallel with the extensive literature containing fusion reactor design
studies with detailed cost estimates , there have been several publications / 52-
58 / which have sought to demonstrate through general arguments that fusion
power will be uneconomic .        These publications argue that fusion devices can
achieve only a low power density , need a long energy payback time , require
highly     complex   but reliable    design   solutions ,   have  an   end-product   with
undesirable     features  and    therefore   that  the    present   strategy  of   fusion
development is incorrect .
5.4.1     Power density
        With regard to power density , it is certainly very likely that the power
density in the fusion power core ( see glossary ) will be considerably lower
( typically 30-40 times ) than inside a fission reactor pressure vessel . Even if
it were sensible to use the same cost per unit volume for both systems , and
even if the fission reactor pressure vessel were to amount to the high figure
of 7% of the construction cost of a fission plant , this power density factor
would only lead to an increased construction cost of fusion over fission of 3 _ 4
times .    That solely power - density- based comparisons are not very reasonable can
be seen by examining fission itself , where typical power densities in a PWR ,
                                                                  3
AGR and Magnox reactors are around 15 , i and 0.4 MW^/m                respectively / 59/
whereas the construction and generation cost differences are within a factor of
2 / 60 /.
          In fact , topologically a fusion reactor most resembles a coal or oil
plant ,   in that it has a single combustion chamber surrounded by a heat sink .
Of course , in the case of fusion , this heat sink must be much thicker than with
a coal plant to absorb neutrons ,         and the combustion chamber must be under
vacuum and filled with magnetic field , and this leads leads to greater expense
for the fusion " furnace ".       However , the power density averaged over a coal
                                              3
combustion chamber is about 0.1 MWtf( /m /61 / compared to the typical fusion
power core value /59/ of 0.5 MW^/m^ expected in a reactor .
        In addition , the construction cost difference between coal and fission is
in contradiction to the difference in their power densities , again showing the
weakness of power density in comparing different power generation systems .
Power density is only a useful indication of cost trends when changes are made
 ---pagebreak---                                                                                  87 .
to a single design concept of one particular power generation system , as in
section     3. and  it  is not  realistic to use     it as the only yardstick for
comparisons of different types of systems .      It should also be realised that the
low power density of fusion may turn out to be a considerable advantage due to
its tendency to produce safety benefits .
5.4.2    Energy payback ( Net Energy Gain )
        As far as energy payback time is concerned , it is important to consider
lifetime energy requirements for construction ,        fuelling and operating power
plants and their output as a function of time in order to see the full picture
/ 32 , 33 /.   When this is done , energy payback time ( i.e. the time after the
commissioning date to recover the energy expended up to that point ) turns out
to be a rather misleading term to use , and should be replaced by the net energy
gain over the lifetime of the plant .       As was demonstrated in section 2 ( Table
2.4 ) fission has considerable energy expenditure on replacement fuel after
commissioning and this is not present with fusion .          In fact , the net energy
gain over the lifetime can turn out to be higher for fusion than fission .
5-4.3    Маззез
       That fusion can hope to be eventually competitive in price with fission is
shown clearly by comparisons of the material masses involved in both plant
types / 62 /. The ratio of masses between the presently conceived fusion power
core ( including lithium-containing breeder ) and a PWR reactor pressure vessel
( including fuel ) is around a factor of 30 . However , when the full plant is
considered , the mass of metals in the plant ( which are the highest cost and
energy-using components of the plant ) is around 30$ higher for fusion .
5.4.4    Complexity
        It has also been argued that fusion involves much greater complexity than
fission , and that this will both push up component costs and reduce system
availability , both having an effect on generation costs . This argument cannot
yet be conclusively refuted , but because of the lower power densities in fusion
plants compared to fission plants , fewer safety systems , whose failure would
interfere with plant availability , will be required . For comparison , todays
aircraft have many more systems and are much more complex , yet they are now
much more reliable than in earlier times .        By analogy , fusion ought similarly
 ---pagebreak---                                                                                       88
to be able to cope with the complexity of Its systems without an excessive cost
penalty .
5.4.5      Undesirable Characteristics
         Fusion has also been criticized for having undesirable qualities in the
end-product reactor .          These centre around the use of lithium and tritium , the
presence of high energy neutrons , and pulsed operation .
        As far as lithium is concerned , the European strategy excludes its use in
the metallic form in which it presents a fire hazard .                   From the resource
viewpoint lithium is not a serious restraint on the expansion of fusion , since
a typical 1200 megawatt reactor lithium lifetime requirement ( of which 1 / 10th
is consumed ) is around 100t of enriched lithium / 10 / compared with world
reserves ( on land ) estimated in 1970 at 180 Mt /63 /.                 Taking account of
enrichment ,     but     without    considering    the  possibility  of   recycling  unused
lithium , 500 fusion plants would take around 500 years to consume 5% of the
world land-based resources . This is less than but comparable to the predicted
timescale for consumption of energy reserves in the most well-endowed European
countries , so it might be argued that the development of fusion is therefore
unnecessary .         However ,   the purpose of the present programme is to develop
fusion , so as to be able to choose the best system at any given time , bearing
in mind the problems that may arise with alternative power generation methods
( e.g. C02 with fossil fuels ).
          Furthermore , sea-borne lithium resources are nearly 20000 times larger
than land-borne and in energy terms 40 times larger than sea-borne uranium
/ 57 /). Lithium also occurs at 500-1000 times the concentration of uranium / 64 ,
65 / making extraction more economically viable .               In addition , recycling of
unused lithium might be contemplated as a means of stretching resources by a
further order of magnitude .           Also , within the above half-millenniumm a greater
understanding        of  the    fusion  process and a desire      to optimise  the process
further is likely to lead to an evolution away from dependence on tritium ( and
hence on lithium ), to use possibly pure deuterium as a fuel or even an isotope
                  ■s
of helium ( He ) found throughout the solar system / 66 /. For the relatively
near term , however , it should be noted that even now there is considerable
knowledge of how to handle tritium at the concentrations required for fusion ,
under a commercial reactor operating environment , it being a by-product of the
irradiation of heavy water in CANDU reactors .
 ---pagebreak---                                                                                      89 .
       With regard to high energy neutrons in the fusion process , this is the
price paid for having clean reaction products , and gives an advantage ,
especially when comparison is made with the long term disposal of fission
products .     ( This point is considered further in the companion report on
Environmental Aspects of Fusion ). It is worth noting however that no practical
fusion   fuel    for   a   man-made   power   source   is   completely   neutron-free   and
therefore    there     is    always   some   residual   radioactivity     associated  with
structural materials surrounding the reactor .             It is by developing the most
suitable    surrounding       materials ,   having   very    low  levels    of  long-lived
radioactivity ,      that fusion will reach its full potential , and the costs of
developing or manufacturing these materials is not thought at this stage to be
prohibitive / 67 /-
      Steady state operation of a fusion device might be desirable both from an
operational viewpoint and to reduce the fatigue experienced by the reactor
subsystems .      The    principle   has   already   been   demonstrated   experimentally ,
although at this early stage of its development there are doubts about its
economic viability on a commercial scale .         In the end , its implementation will
depend on the relative effects on generation cost of the efficiency of the
method used for maintaining steady state operation and of the increased quality
of fatigue-resistant materials and components .          In any case , living with cyclic
fatigue is not a unique problem for fusion ,              it being commonplace in many
complex structures today .
5.4.6  Strategy
         The strategy       and justification    for developing fusion has also been
questioned / 56 / implying that the likely return from fusion is small compared
to the investment on its development .          Although it is impossible to say today
with absolute certainty that the present development programme will result in
the successful implementation of fusion power ( it being the purpose of the
programme to find out whether this is possible ), the potential long-term return
if fusion were implemented would be enormous because of the long time over
which this return would be made .            As a proportion of generation costs for
fusion reactors over this long timescale , development costs can only be a
minuscule proportion .
          The critisism has also been made / 54 / that , by concentrating on DT
Tokamak fusion , prospects are weakened for ever developing better alternative
 ---pagebreak---                                                                                   90 .
fusion concepts .     Even proponents of DT fusion realise that their present
reactor concepts will have to be improved upon to make them as highly desirable
as fusion was initially claimed to be , but realise that the best way to find
out how to make such improvements is to pursue at least one line of research
vigorously towards the commercialization phase .        DT Tokamak fusion looks from
the present viewpoint to be able to achieve the earliest commercialisation date
but other confinement methods are not being neglected .           In fact about 1 0% of
the  worldwide   and   European   fusion   budget  is  being  spent    on research and
development of alternatives to the tokamak / 68 /.         Whether DT or more exotic
fuels can economically be used in such confinement schemes will depend on the
confinement physics attained .      In any case the status of such alternatives to
the  Tokamak   is   continually    being   re-examined   and  a    check-point  on     the
development status of such schemes is already planned in the European programme
before proceeding to a demonstration fusion reactor .         Concentrating on the DT
Tokamak line at this stage is intended to produce information which would be
valid for whatever confinement concept is pursued further at that time .
     In summary , therefore , the information presented by the critics of fusion
is  often  highly   selective ,   and  the   conclusions are    not   supported by     the
detailed studies .     It  is   true that the low power density of many present
designs leads to high capital costs , but the estimated cost of electricity from
fusion power stations is not so much greater than forecast costs from existing
or other alternative energy sources that fusion can be dismissed on economic
grounds .
 ---pagebreak---                                                                                  91 .
6.   CONCLUSIONS
       Since the earliest commercialisation date for fusion power looks from the
present perspective to be around the middle of the next century , any prediction
today of its economic prospects is rather uncertain .        However , this has not led
to the development of fusion without consideration of its ultimate economic
potential as is witnessed by the considerable number of power reactor studies
whose results are recorded in this report .        By the very nature of our present
understanding of fusion and its technology , these studies give rather a wide
range of results .     They do prove extremely useful , however , in identifying
general trends for future development .      It is clear of course that if a fusion
reactor had    to be constructed     today , using   the   presently available plasma
parameters   with  their   established   scalings and using presently established
technologies , that reactor would have an electricity cost in the upper range of
the   projections  for   other   systems .    However ,   fusion   physics  and   fusion
technology have developed by orders of magnitude over the last 20 years .           This
history and the present experience in fusion research lead to the belief that
the development potential for fusion will , over the comming decades , result in
considerable improvements in the relationship between the generation cost for
fusion and that of other systems .
       Not only is it impossible to forecast the economic conditions , it is also
difficult to fully appreciate now the improvements which will undoubtedly occur
during the further development of the fusion reactor system .         Examples given in
the previous sections show that such improvements can also be expected from
innovations   which are    not  necessary on    present-generation systems .       Their
impact will only become significant if         integrated into full-scale reactors .
The programmes on Stellarators and Reversed Field Pinches could also have an
important influence .     In any case , the development cost for fusion power is
only a small fraction of todays expenditure for energy supply .            Finally , the
use of fossil fuel will eventually have to be restricted to those applications
where there is no alternative , such as transport .               The increasing C02~
accumulation may otherwise lead to difficulties .        It is therefore essential to
have more than one high-potential energy source available working without any
CC>2 production , and thus in all respects environmentally acceptable , and the
ultimate goal for fusion reactor development is to satisfy this need .
 ---pagebreak---                                                                                   92 .
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/8/       An analysis of the estimated capital cost for a fusion reactor ,
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 ---pagebreak---                                                                                93 .
/ 1 2/ The reversed field pinch reactor ( RFPR ) concept .    R.L. Hagenson et al .
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/ 1 3/ Witamir-I , A University of Wisconsin Tandem mirror reactor design , B.
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/ 1 6/ UWTOR-M , A conceptual modular stellarator power reactor
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/ 20 / Hiball - a conceptual heavy ion beam driven fusion reactor study ,
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/ 21 / Hiball-II . An improved conceptual heavy ion beam driven fusion reactor
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/ 22 / A currency exchange rate for use in technical comparisons .
       D.E.T.F. Ashby .      CLM-R245 , May 1984 .
/ 23 / Cost sensitivity analysis of possible fusion power plants , R. Bunde ,
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/ 24 / Scaling of tokamak reactor costs , W.R. Spears and J.A. Wesson , Nuclear
       Fusion , Vol 20(12 ), pp 1525-1532 , December 1980 .
 ---pagebreak---                                                                                       94 .
/ 25 / Tokamak and reversed field pinch reactor cost scaling P.I.H Cooke ,
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/ 26 / Factors affecting the minimum capital cost of a tokamak reactor
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       1214 , September 1980 . ( Also CLM-P623 ).
/ 27 / Cost scaling of tokamaks , J. Sheffield and A. Gibson , Nuclear Fusion 15 ,
       pp 677-685 , 1975 .
/ 28 / Cost assessment of a generic magnetic fusion reactor ,
       J. Sheffield et al .          Oak Ridge National Laboratory Report ORNL / TM-9311
       ( 1986 ).
/ 29 / Generic magnetic fusion reactor cost assessment , J. Sheffield , Journal
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/ 30 / Report on high power density fusion systems ( MFAC , Panel X ) May 1985 .
/31 /  Compact Fusion Reactors , R.A. Krakowski , R. Hagenson , Los Alamos Report
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/ 32 / Evaluation of          the  energy   required   for  constructing and operating a
       fusion power plant , R. BUnde , Proc . 12th Symposium on Fusion Technology ,
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/ 33 / NET energy gain from DT fusion , R. BUnde , Proc . 13th Symposium on Fusion
       Technology , Varese , pp 181 – 1 88 . September 1984 .
/ 34 / The potential net energy gain from D-T fusion power plants , R. BUnde ,
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/ 35 / Energy analysis of coal , fission and fusion power plants ,
       N. Tsoulfanidis .         Nuclear Technology / Fusion , Vol 1 , pp 238-254 ,
       April 1981 .
/ 36 / The      tokamak      hydrid    reactor , J.L.    Kelly   and   R.P.  Rose , Nuclear
       Engineering and Design 63(2 ), pp 395-421 , March 1981 .
/ 37 / The SCAN-2 cost model , NET report EUR- FU/XII - 80/86/62.
 ---pagebreak---                                                                                         95 .
/ 38 /   Electricity Generation Costs Assessments made in 1984 for stations to be
         commissioned in 1995 . Moynet G. , UNIPEDE Study , 1985 .
/ 39 /   A Model for the Computation Design of Tokamaks - Part I : general
         OverView , Borrass , K. , NET Report EUR - FU/XII - 361 / 85 / 42
/ 40 /   Reactor Beyond NET , Spears W.R. , to be published in Proc . IAEA Tech .
         Ctte Mtg . & Workshop on Fusion Reactor Design & Technology , Yalta , USSR ,
         26 May - 6 June 1986 .
/ 41 /   DEMO & FCTR Parameters , Spears , W.R. , NET Report EUR - FU/XII - 361 / 85 / 41
/ 42 /   Reactor Cost Driving Items , Spears , W.R. ,        to be published in Fusion
         Technology , Proc . 14th Symposium , Avignon , France , September 1986 .
/ 43 /   IAEA Tec doc / 373 , 1986 .
/ 44 / • Status of Liquid Metal Fast Breeder Reactors , Technical Reports Series
         No . 246 , IAEA , Vienna , 1985 .
/ 45 /   Nucleonics Week , January 23rd 1986 .
/ 46 /   Projected Costs of Generating Electricity from Nuclear and Coal-fired
         power stations for commissioning in 1995 .       OECD / NEA . Paris 1986 .
/ 47 /   The economics of the nuclear fuel cycle , OECD /NEA , Paris 1985 .
/ 48 /   Minimum cost of photovoltaic energy for a utility grid and general
         features    of   a  generating    plant  using  costless      solar   cells . Madet ,
         D. , Fourth E.C. Photovoltaic Solar Energy Conference , Stresa , 10-14 May
         1982 .
/ 49 /   Solceller i et fremtidigt dansk energisystem .
         Nielsen , L.D. , in Riso National Laboratory , Den teknologiske udvickling
         og dens betydning for udformningen af det fremtidige energisystem ,
         Roskilde , Denmark , 1984 .
/ 50 /   Photovoltaics      Program    Overview .   P.D.   Maycock ,       Proc .  3rd   E.C.
         Photovoltaic Solar Energy Conférence , Cannes ,          France , 27 - 31 October ,
         1980 . pp. 10-17 .
 ---pagebreak---                                                                                        96 .
/51 /  M.R. Starr , The potential for photovoltaics in Europe .             Proc . 4th E.C.
       Photovoltaic Solar Energy Conference , Stresa , Italy , 10-14 May , 1982 .
       pp . 40-50 .
/ 52 / Neutron wall    loading ,   power density and pay-back time K.H. Schmitter ,
       Proc . 11 Symposium on Fusion Technology , Oxford , pp 1255-1259 .
       September 1980 .
/ 53 / The    fusion dilemma ,   R.  Carruthers ,    Interdisciplinary Science Reviews
       6(2 ) , pp 127-141 , 1981 .
/ 54 / The trouble with fusion , L.M. Lidsky , Technology Review ( MIT ),
       October 1983 .
/ 55 / Some critical observations on the prospects of fusion power ,
       D. Pfirsch and K.H. Schmitter , Proc . 4th Int . Conf . on Energy Options ,
       London , pp 350-355 , April 1984 .
/ 56 / Models for the assessment of research and development - Why does fusion
       get    such a good    press ?  C.W.    Hope ,  Proc . 4th Int .   Conf .   on Energy
       Options , London , April 1984 , pp 356-358 .
/ 57 / Fusion Thermonucleaire Contrôlée -
       La grande illusion , André Ertaud , Revue Generale Nucléaire , 1985 , No . 3 ~
       Mai-Juin .
/ 58 / Kernfusion , Rudolf Wienecke , Bild der Wissenschaft 3 / 81 .
/ 59 / Small fusion reactors : problems , promise and pathways , Krakowski , R.A. ,
       Hagenson ,   R.L. ,  Miller ,   R.L. ,   Fusion   Technology    1984 ,   Proc .   13th
       Symposium pp 45-58 .
/ 60 / Fission , Fusion and the Energy Crisis ( 2nd Edition ) Hunt , S.E. , Pergamom
       Press 1980 , Chapter 8 .
/61 /  Didcot Power Station , Techieal Publications Department , CEGB Midlands
       Region .
 ---pagebreak---                                                                                     97 .
/ 62 / The potential net energy gain from DT fusion power plants , ' Bilnde , R. ,
       Nuclear Engineering and Des l gn / Fus J on , 3 ( 1985 ) 1 36 .
/ 63 / Fusion Research , Dolan , T.J. , Pergamorn Press 1982 .
/ 64 / Controlled Thermonuclear Fusion , J. Raeder et al . Wiley & Sons ( 1986 )
/ 65 / Encyclopaedia Britannica .
                              3
/ 66 / Lunar   Source   of He    for  Commercial       Fusion Power ,  Willenburg ,  L.J. ,
       Santarius , J.F. , Kulcinski , G.L.        Fusion Technology 10 ( 1986 ) pp.167 -
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/ 67 / Private communication G.J. Butterworth , ( 1986 ).
/ 68 / Long term planning towards a Demonstration Fusion Reactor G. Grieger
       ( Chairman ) et al.   EUR FU XII / 708 / 77 / LTP50 ( 1977 ).
/ 69 / Fusion reactor design studies - standard unit costs and cost scaling
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/ 70 / The costs of      Generating Electricity         in Nuclear   & Coal  Fired   Power
       Stations .   Report by Expert Group of NEA / OECD , 1983-
 ---pagebreak---                                                                                     98 .
8.     GLOSSARY OF TERMS AND DEFINITION .'
Direct Capital Cost
         The direct capital cost of a fusion power station includes the purchase of
the site , structures and site facilities , the reactor plant , and the turbine
and electrical plant ( Items 20-26 in the standard US-DOE accounting system
/ 69 / ) .
Spécifie Direct Capital Cost (= Unit Direct Cost )
         Direct capital cost per unit electrical power sent out (P              )
Indirect Capital Cost
           Project management , design , services , licensing and all personnel costs
during construction .
Generation Cost
         According to OECD / NEA / 70 /:
            " the   ideal  calculation will   take . iccount  of the   time   flows
            of    money   expended   on  constructing     the  station ,   on   its
            operation ,     on  its   fuel  and   on    subsequent    spent    fuel
            management and station decommissioning ...
            These costs will be discounted back to a selected base date
            and added together to arrive at a total cost in present
            worth terms .
            If the total present worth cost is divided by the sum of the
            discounted     annual   electricity   output    over  the   station’s
            life ,   a levelised generation cost       is obtained in constant
            monetary units .     If each kWh sent out from the station over
            its   lifetime was sold for this " levelised cost " the         income
            in present worth terms would exactly equal the total present
            worth costs of construction and operation ."
 ---pagebreak---                                                                                                  99 .
      TI u ; l <; V' - 1 i ■/<!<! i'<;n<‘r.it, lori c:osl 1:1 UK.'I'OI'OI'II oxpre:J:3i;d by
                           D   3  I   + Z   +   M +  F +  R
                           N                              * ,
                           I P so A n 8.76 / ( 1 + d ) n
                         n=1
where N is the plant lifetime in years , Pgo is the rated power sent out by the
plant ( MW ) and An is the plant average availability in year n . The cost items
in the numerator are direct ( D ) and indirect ( I ) capital costs , interest during
construction ( Z ), operation and maintenance costs ( M ), fuel costs ( F ) and
decommissioning ( R ), all discounted to the date of commissioning using the
discount rate d .
Fusion Power Core ( FPC )
        Torus ( first wall / blanket / shield ), Magnets ( toroidal and poloidal field )
and their respective support structure .
Mass Expenditure ( ME )
      The mass of material needed for the FPC divided by the power sent out .
β
      Ratio of plasma kinetic pressure to the presssure of the toroidal magnetic
field confining it .
q
       A measure of the twist of the field line - the number of times the field
lines pass round the major circumf erence before returning to the starting point
in the minor circumference .                       To resist gross instabilities this must be greater
than 2 at the plasma edge and above unity on axis .
g
      The      beta          level ,      i.e.      the   coefficient        in   the    scaling
6 (?) = g I ( MA) / a(m)B(T ) where I is the plasma current , a the minor plasma
radius and B the toroidal field on the plasma axis .
 ---pagebreak---                                                                      2„lJ
      The ratio of mean plasma fusion power density and the product 3 B (B i£
toroidal field at the plasma centre ).    It measures the extent to which the
fusion reaction rate at the average plasma temperature is modified by spatia ]
variations in plasma temperature and density .