ARCHIVES HISTORIQUES 
DE LA COMMISSION 
 
 
 
COLLECTION RELIEE DES 
DOCUMENTS "COM" 
 
 
 
 
COM ( 83) 550 
 
 
 
Vol. 1983/0 208 
  
 
Disclaimer  
Conformém ent au règlement (CEE, Euratom) n° 354/83 du Conseil du 1er février 1983 
concernant l'ouverture au public des archives historiques de la Communauté économique 
européenne et de la Communauté européenne de l'énergie atomique (JO L 43 du 15.2.1983, 
p. 1), tel que modifié par le règlement (CE, Euratom) n° 1700/2003 du 22 septembre 2003 
(JO L 243 du 27.9.2003, p. 1), ce dossier est ouvert au public. Le cas échéant, les documents 
classifiés présents dans ce dossier ont été déclassifiés conformément à l'article 5 dudit 
règlem ent. 
In accordance with Council Regulation (EEC, Euratom) No 354/83 of 1 February 1983 
concerning the opening to the public of the historical archives of the European Econom ic 
Communit y and the European Atom ic Energy Communit y (OJ L 43, 15.2.1983, p. 1), as 
amended by Regulation (EC, Euratom) No 1700/2003 of 22 September 2003 (OJ L 243, 
27.9.2003, p. 1), this file is open to the public. Where necessary, classified documents in this 
file have been declassified in conformity with Article 5 of the aforementioned regulation. 
In Übereinstimmung mit der Verordnung (EWG, Euratom) Nr. 354/83 des Rates vom  1. 
Februar 1983 über die Freigabe der historischen Archive der Europäischen 
Wirtschaftsgemeinschaft und der Europäischen Atomgem einschaft (ABI. L 43 vom  15.2.1983, 
S. 1), geändert durch die Verordnung (EG, Euratom) Nr. 1700/2003 vom 22. Septem ber 2003 
(ABI. L 243 vom 27.9.2003, S. 1), ist diese Datei der Öffentlichkeit zugänglich. Soweit 
erforderlich, wurden die Verschlusssachen in dieser Datei in Übereinstimmung mit Artikel 5 
der genannten Verordnung freigegeben. 
COMMISSION  OF THE EUROPEAN COMMUNITIES 
COM(83)  550 final 
Brussels,  21 September  1983 
PROJECTS  OF EUROPEAN  SIGNIFICANCE 
(Second  Progress Report from  the Commission to the  Council) 
COM(83)  550 final 

- 1 -
TABLE OF CONTENTS 
Page 
PROJECTS  OF EUROPEAN  SIGNIFICANCE  2 
I. A TRITIUM  HANDLING LABORATORY AT  J.R.C.-ISPRA 
Second progress report  1.1 
1. Introduction  1.4 
2. Executive  Summary 1.6 
3. Main report 1.18 
II. A VIBRATING TABLE  AT J.R.C.-ISPRA 
Second progress report  II.1 
III. PERSPECTIVES  OF HOSTING AT THE  J.R.C.-ISPRA 
ESTABLISHMENT  A FUSION MACHINE  LIKE IGNITOR  -
Second progress report  III.1 

- 2 -
Commission  of the  European  Communities 
Joint Research Center 
PROJECTS  OF EUROPEAN  SIGNIFICANCE 
1. Introduction 
Three preliminary studies  were begun in March 1983 following  the 
Council  meeting of the 10th of that  month in which the  Commission 
was requested to study projects  suitable  for siting  at the JRC 
Ispra Establishment. 
Terms of reference included examination of : 
- opportunity and  need at Community  level; 
- feasibility  at Ispra; 
- investment costs. 
The subjects  examined  are : 
~ a Tritium  Handling Laboratory; 
- a Vibrating  Table; 
- Ignitor. 
These subjects being of  differing  natures,  different  elements  have had 
to be taken  into account in  each case and the rate of progress has not 
been uniform. 
2. Tritium  Handling Laboratory. 
The second progress report on the  feasibility  of constructing in Ispra 
a Tritium  Handling Laboratory has reached the necessary completeness as 
to constitute  a complete  technical  study. This report has  been already 
presented to the Consultative  Committee  for Fusion  Programme  (CCFP), 
which has introduced the  tritium  in its  action  of "First Round  Projects" 
in the frame of the  1982-1986  Fusion  Technology  Programme. 

- 3 -
The opportunity and the  need of a European  Tritium  Handling Laboratory 
to serve the  fusion  reactor research  programme,  already stressed by the 
Fusion Review  Panel held in 1981, is  well demonstrated. The report 
analyzes in  detail  the expected  problems  related  to tritium  in the 
design-studies  of the post-JET experimental reactor  INTOR and of 
NET (Next  European  Torus). The  tritium  cycle in a  fusion  reactor has 
operational  and safety requirements which are  different  from and  much 
more complex  than those found in  military  plants for  tritium  production. 
JRC-ISPRA  can candidate for the construction and operation of such a 
Laboratory on the basis of past and present experience in  hydrogen-
isotopes  technology. 
The facility  to be placed at the  disposal  of all the  fusion  associa­
tions, would be a flexible  multipurpose  installation,  capable of 
handling  up to 100  grams  of Tritium.  It would be mostly dedicated 
to study  tritium-related  safety problems, to provide conditions to 
handle and  test components  as well as tritiated  parts of the  fusion 
reactors. 
The total  staffing  of the laboratory  when in operation is evaluated 
in a maximum  of 25 persons (of which 7  professionals).  The previous 
evaluation  of the  installation  costs of the Laboratory,  about 10 Mio 
ECUS (1983 value), is being evaluated  more precisely  by an external 
engineering  bureau. The  results  of this study  will be available 
by September  15th. 
3. Vibrating  Table 
The second prograss report on  vibrating  tables is  first of all a 
detailed  study on the  needs  of this kind of  installations  as well as 
of their  present  availability  in the  Community  and in the world. Also, 
current  research on seismic  analysis  is reviewed, in order to assess 
current  trends in  this field. 

Most Member  States consider that present  norms for seismic design 
may be adequately  fulfiLled  using already  available  experimental  and 
analytical  tools.  Nevertheless, there  is a shared  feeling  in the 
Community  that a shaking  table of larger  capacity than  existing  ones 
might be beneficial  as a cooperative  European  effort,  in order to 
standardize  ground  motion  measurements,  thus creating  a European  data 
base, and develop  and test refined  models  and computer  codes, to 
enable a better understanding  of the  phenomena  and improve  safety  norms. 
No major problems  seem to exist for the  siting of such a table in 
JRC-Ispra.  However,  study contracts  have been launched  in order to 
investigate  by static and dynamic  geotechnical  tests the adequacy  of 
the preselected  site in the  Ispra Establishment  to host the installation. 
The JRC will continue  in the  next years  the accurate  technical  and finan­
cial assessment  of this operation, which presents aspects  of great 
interest  for the  Community's  needs,  and is  well placed  in the  general 
safety motivation  of the  Joint Research Center. 
4. Igni tor 
In this second progress report,  the three main items relevant  for the 
siting of IGNITOR  in JRC-Ispra  : electrical  supply system,  main building 
and auxiliary  systems  and licensing,  have been further  investigated.  In 
particular,  installation  costs and the  annual charge  for the  electrical 
power supply  have been assessed. 
The installation  in the  ESSOR containment  building  may offer  practical 
advantages,  due to  existing  facilities  in this area. The licensing 
procedure  does not seem to present  major  problems. 
The final evaluation  of all the costs involved  in the  Ignitor  installa­
tion, as well as the  complete  technical  study  will be the  object of a 
final report, which  will be ready at the end of  September  1983. The Joint 
Research Center considers  in this way to  have fulfilled  its obligation 
to prepare  a feasibility  study for the  siting of IGNITOR  in Ispra. It 
will proceed  further  in the  case a request  in this sense would be formulated 
by at least one of the  Fusion Associations. 

COMMISSION OF THE EUROPEAN COMMUNITIES 
JOINT RESEARCH CENTER 
I. A TRITIUM  HANDLING LABORATORY 
AT 
J.R.C. - ISPRA 
Second progress  report 

I. 2 
TABLE OF CONTENTS 
page 
1. INTRODUCTION I.  4 
2. EXECUTIVE  SUMMARY  I.  6 
2. 1 General  Remarks  I.  7 
2. 2 The NET  requirements  I.  7 
2. 3 Review of the  existing  Know-how and of the  activity  in 
progress  I.  8 
2. 4 JRC - experience  and present  activities  I.  9 
2.:5 Objectives  of the JRC  - Tritium Laboratory  1.10 
2. 6 Description  of the  Laboratory  1.11 
2. 7 Location  1.12 
2. 8 Licensing  procedures  1.12 
2. 9 Planning  and resources  1.13 
2.10 Conclusions  1.14 
3. MAIN REPORT !.1 8 
3. 1 Remarks  related  to fusion reactors  as breeder  reactors  1.19 
3. 2 NET and  power reactor  requirements  related  to tritium  1.21 
3.2.1 Introduction  1.21 
3.2.2 Tritium  permeation  and inventory  in structural 
materials  1.21 
3.2.3 Tritium processing  systems  1.23 
3.2.4 Effects of tritium  contamination  during standard 
operation,  maintenance  and accidental conditions  1.26 
3.2.5 Tritium inventory  1.28 
3.2.6 Time scale requirements  for R&D 1.29 
./. 

I. 3 
3. 3 Review  of the  existing  know-how  and of the  activities  in 
progress 1.30 
3.3.1 Experience from  fission  reactors 1.30 
3.3.2 Activities  in support to Fusion 1.31 
3. 4 Past and present  activities  at JRC 1.38 
3.4.1 Operative experience in  tritium  handling and 
monitoring 1.38 
3.4.2 Research  activities  in fusion  technology  related 
to tritium  problems  1.41 
3. 5 Objectives of the JRC  Tritium  Laboratory I.47 
3.5.1 Motivation and  orientation  I.47 
3.5.2 Experimental  program  proposed,  related  to the JRC -
Fusion  activity  1.48 
3. 6 Description of the laboratory 1.58 
3.6.1 Building  1.58 
3.6.2 Tritium  containment  systems  1.58 
3.6.3 Siting  of the laboratory 1.60 
3.6.4 Licensing procedures 1.61 
3.6.5 Planning and resources I.70 
3.6.6 Cost evaluation 1.71 
3. 7 Conclusions 1.72 
APPENDIX  1 : 
First round projects on  tritium  set up by the Fusion 
technology Steering  Committee  I.73 

I. 4 
1. Introduction 
Tritium  has been  utilized for many  years  already in the countries of the European 
Community  in the field of biological, biochemical and chemical research and  even 
for industrial and commercial applications. The confinement and conditioning of 
tritiated  waste  waters  arising in nuciear fuel reprocessing  have  also  been  objectives 
of research  efforts  in the Community. 
Tritium  will be used as a fuel in fusion reactors, but its handling in large quantities 
implies the development of advanced and  safe  technologies.  Anyway, tritium is  pro­
duced in macroquantities in classified installations of the nuciear weapon  states,  i.e. 
France  and Great  Britain.  These two countries  have  declared, in a  statement  at the 
Atomic  Questions  Group  in December 1981 (ref. 1), to be disposed to the  release 
of know-how in tritium  technology  relevant to fusion. The  statement  of the UK 
delegate,  which the French  delegation  declares  to support, without reservations,  ends 
as follows: 
"I am authorized to  state  that  the United Kingdom would be happy to  assist  the 
"Commission and the Community oy making available to it information concerning 
"its axperience with tritium  technology  relevant to the Community's programme. 
"I would  hope  that  this  would help to  define  more  closely  the areas  where  new 
"work  needs  to be undertaken." 
The concern, however,  expressed  by the "European Fusion Review Panel" in its 
report of December 1981,  "that  only a pan'of the know-how required in  this  field 
(namely the manipulation of tritium for the purpose of fusion  technology)  exists 
already in the classified laboratories", is understandable, if one  takes  into account 
the quite different  goals  and requirements such  techniques  have  been  developed for. 
The panel supports its  statement  by the fact  that  the United  States  have  built a  civil 
tritium  laboratory at Los Alamos, dedicated to fusion. Such a laboratory was  felt 
to be mainly  necessary  to demonstrate, in relation to regulatory requirements, the 
safety  of the tritium manipulation  techniques  for the  civil  application of fusion. 
This  laborator/, however, is not completely open to international collaboration. 
The panel recommends the construction of a tritium  test  facility  devoted  to the de­
velopment of  technologies  in the frame of the Community's Fusion Programme, in 
consideration of the fact  that  Europe is in an extremely  weak  position in  this  field. 
The panel points cut furthermore the opportunity to deal already now with the 
aspects  of safety  and the environmental impact of fusion anticipating the problems 
of its industrial application (ref. 2). Among the principal  items  of these  safety  as­
pects  the handling of tritium is mentioned. 
Similar  concern for tritium related  safety  aspects  was expressed  by an  IAEA  workshop 
(ref. 3), attributing high priority to research and development on tritium problems. 
The Commission of the European Communities, aware of the problem,  have  introduced 
the tritium question in its  "First  Round Projects" aimed at sharing  R&D  activities  in the 
Community  in the frame of the  1982-1986  Fusion Programme (ref. 4). 

I. 5 
References 
1. Statement of the  British  and  French  delegations at the Atomic Questions  Group  offering the 
release of know-how in  tritium  technology relevant to fusion.  Annex  A and  Β to document 
ΑΤΟ  98, 10  December  1981. 
2. Rapport  du "European  Fusion Review  Panel",  Commission  des Communautés Européennes, 
CRE3T-EN/216b,  SEC(8 1) 1933,  Brussels,  3 December  1981. 
3. IAEA  workshop on Fusion Safety,  23-27  March,  1981. 
4. European  Fusion  Technology  Programme  1982-1986,  Tasks  Tl toT6.(see appendix  1). 

I. 6 
2. EXECUTIVE  SUMMARY 

I. 7 
2.1 General Remarks 
The basic factors of the tritium cycle of a post-JET experimental reactor 
(INTOR)  and of a power fusion reactor are summarized in Table 1 and  Fig.l. 
This cycle has a number of requirements  which  are new as compared to the 
ones  related to  those  found in military plants for tritium production. In 
this last case, in particular,  there  is no plasma chamber, no exhaust pro­
cessing nor recovery from coolant. Furthermore  there  is no requirement for 
tritium  breeding as in the power fusion plant. This indicates  that  there 
will be many components and processing units, in the tritium  fuel  cycle of 
a fusion reactor,  that  have no equivalent in the technology developed so 
far, and  that  require specific  R&D. 
Table  1: Tritium  Balance in Fusion Reactors 
Tritium  flow Inventory Tritium  burned Tritium  breeding 
rate g/d kg kg/a rate 
INTOR^ 1600 4-5 12 0.6 
POWER  , . 
REACTOR  ' 8000 10-20 120 1.05 
(1) stage  III,  50%  availability  ; (2) 1600  Mwe,  80%  availability 
2.2 The NET requirements 
In order to identify the specific research and development  needs  in Europe, 
a detailed analysis of the expected problems related to tritium in NET 
(Next  European Torus) has been performed. For this analysis reference is 
made above all to the design study of  INTOR  (phase  1 and phase  2a). 
The main topics investigated are  : 
a) tritium permeation, inventory and outgassing in structural materials 
(first  wall  and reactor components); 
b) systems related to the various  aspects  of the  fuel  cycle (tritium reco­
very from exhaust plasma, coolant and breeder;  fuelling,  waste  treat­
ment and disposal ) ; 
c) reactor contamination during standard operation, maintenance and  acci­
dental conditions; 
d) processing and disposal of structures and components as waste. 

I. 8 
Regarding  topic a), the problems related to tritium implantation and 
permeation through the first  wall  are complex and the available'data 
insufficient.  An effort is required to extend the existing data on per­
meation and outgassing · of H/D to tritium, and to study the combined 
effects of neutron irradiation and tritium. 
Regarding  topic b), different schemes have been proposed for plasma ex­
haust  neutral beam processing,  fuelling,  coolant detritiation and  tri­
tium recovery from blanket. The need to develop large tritium-compati­
ble components has been recognized. This applies in particular to pumps 
(cryogenic-and  turbomolecular pumps) as  well  as to valves and other  cir­
cuit components. An important problem  will  be that  of reweldability of 
steel  with  high tritium content. 
A number of specific  needs  related to the single flow-sheets have been 
identified.  The development of  apparatus  is considered important in or­
der to optimise the various processes. This applies, in particular to 
systems for gas purification  (cold  traps, catalytic oxidation) and to 
isotopic  separation units (cryogenic  distillation,  gas chromatography, 
electrolysis,  etc.) for the plasma  exhaast  In these  fields the experien­
ce accumulated in  fission  reactors and military applications  will  be of 
particular help. 
2.3 Review of the existing know-how and of the activity in progress 
The available knowledge on tritium problems and in particular on tritium 
technology is mainly based on the operation experience of  fission  reactors 
(in particular heavy water and  HTGR  reactors) and on the confinement and 
conditioning  of waste water arising in  fission  fuel  reprocessing plants, 
whereas the information coming from the military applications is for the 
moment  still  limited. 
Heavy  water reactors give problems related to handling of large volumes of 
tritiated water and to its processing. Important experience has been gained 
both in Europe and world-wide (mainly in Canada) on  these  problems. 
A detritiation system capable to extract up to 160.000 Ci/a of tritium from 
heavy water is in operation at the high  flux  reactor of Grenoble, based on 
a process developed by CEA. 
In HTGRs  the problems related to tritium permeation from the primary coolant 
circuit  have been investigated and solutions to reduce such permeation by 
coatings have been proposed. 
Tritium  is also a by-product of nuclear (fission) power generation where it 
becomes a concern essentially at the  spent  fuel  reprocessing  stage.  Research 
activities  on the  treatment,  conditioning and disposal of tritiated waste 
waters have been shared in the member countries of the EC in the framework 
of the "Management and Storage of Radioactive Waste" programme (DG XII). 
In the framework of the "Fusion Programme" some actions have been undertaken 
in Europe to deal  with  tritium problems, namely : 

I. 9 
- conceptual design studies on the  fuel  cycle of experimental and 
power reactors; 
- design of the tritium plant for the JET-operation  with  tritium; 
- definition of a first experimental research programme. 
The Joint European Torus is scheduled to  operate  in its second pha­
se with  tritium. So a tritium plant  will  be installed in the JET area 
to purify the burned  gases  coming from the plasma chamber and from the 
neutral injectors. The JET staff is now engaged on detail specifica­
tion of the  Tritium  Plant incorporating the essential process elements. 
A research programme to investigate the tritium problems related to NET 
and power reactor development has been launched recently by the Commis­
sion in the framework of the Fusion Technology actions decided by the 
Fusion  Technology Steering Committe. Study contracts have been propo­
sed by the European Fusion Associations. The  areas  considered are : 
- fuel  clean-up system; 
- tritium recovery from waste streams; 
- tritium detector; 
- electrolytic eel 1; 
- decontamination system; 
- industrial development of large components; 
- tritium recovery from blanket  (liquid  and  solid  breeder). 
The total amount of money already allocated is about 11  MECU.  The experi­
mental work  will  cover the period 1984-1986. 
In USA the construction of the  Tritium  System Test  Assemvly  (TSTA)  at 
Los Alamos is now complete and the commissioning  tests  with  tritium are 
expected to  start  in autumn this year. Other laboratories such as  Mound 
Laboratory  and Lawrence Livermore Laboratory are engaged in experimental 
research. Design and specifications of components for the tritium  test 
plant of the Princeton Large Tokamak  TFTR  have been also performed. 
In Japan a  Tritium  Process Laboratory of multipurpose type is in construc­
tion.  The operation is planned for 1985. 
2.4 JRC-experience and  present  activities 
At JRC-Ispra two heavy water  test  reactors have been operated for many 
years, the  Ispra-1  reactor and the  ESSOR  reactor. Therefore the Centre is 
well equipped for tritium monitoring, health physics and  in-field  control 
analysis.  It has experience on maintenance and component replacement for 
tritium-contaminated water loops and hardware. 
In the framework of the JRC Hydrogen Programme, now approaching its con­
clusions,  a well  acknowledged experience on problems related to hydrogen 
production,  handling, corrosion etc., has been acquired. 
In the framework of the  JRC-Fusion  Technology Programme a number of  activi­
ties related to tritium are on-going at Ispra;  these  activities, started 
in 1978, are of  three  types: 
- conceptual design and flow-sheet  assessments  related to the  fuel  cycle 
of experimental and power reactors; 

I. 10 
- measurement.of basic data on H/D Interaction  with  metals and breeding 
materials; 
- experiments  with  H/D to investigate the feasibility of some  aspects 
of the processes of the  fuel  cycle (electrolytic  cell  development, 
tritium  recovery from  liquid  breeders). 
2.5 Objectives of the  JRC-Tritium  Laboratory 
The analysis of the expected problems in NET and power reactors and of 
the existing know-how calls for a strong effort in the coming years. 
On the basis of the multi-facet situation previously decribed it is belie­
ved that  a Tritium  Laboratory at the JRC-Ispra, dedicated primarily 
to the safety problems of tritium technology in fusion fits  well  to the 
planned activities  within  the Member Countries of the Community. There 
are common objectives to be achieved in this  field,  to which  the labora­
tory  will  be devoted, such as : 
. acquire experience on handling tritium in complex systems in connection 
with the safety of operators during routine operation (general and in­
dividual  protective  means  as  well  as with  the protection of the  envi­
ronment  (waste  treatment,  emergency clean-up, monitoring); 
. provide the conditions to experience procedures for handling tritiated 
parts  of the reactor and to optimize process components in relation to 
their operational safety and tritium inventory. In particular : 
a) execute experiments based on tritium loaded mock-ups reproducing re­
newable  parts  of the NET reactor and aimed at decontamination and/or 
further  treatment  and conditioning (boots, seals, weldings) in ad­
verse operating conditions and by remote handling; 
b) test  new concepts and components developed in the European laborato­
ries,  such as gas chromatography, electrolyzers and catalyzers,  tri­
tium recovery systems from blanket; 
c) check the  validity  of physico-chemical data extrapolated to the  tri­
tium case  from H/D behaviour. The  reliability  of such data is of  pri­
mary importance for the proper design of containment and clean-up sy­
stems. 
To cope  with  these  requirements, a  flexible  multipurpose installation ra­
ther than a  facility  devoted to a specific problem or system  seems  to be 
the most suited.  Given  the objectives of the laboratory, a total tritium 
inventory  not exceeding 100 g can be taken as reference value for the de­
sign. 
The laboratory utilisation  will  be oriented towards two goals: 
- to host research groups from the European Associations not disposing of 
facilities  for tritium handling, 
- to perform tritium related experiments in the frame of the  JRC-Fusion 
Technology  programme. They are : 

I. 11 
. an investigation of the outgasing of D/T in first  wall  structures; 
. a simulation by mock-ups of the handling and  treatment  of typical  reac­
tor components (as for instance the blanket and divertor/1 imiter  seg­
ments  ) ; 
. the optimisation of some process systems such as the electrolytic  cells 
and the gas chromatography technique for isotope separation; 
. the  tritium  recovery  from  liquid  breeders. 
These items correspond to the work already in progress at Ispra in  "cold" 
conditions  (i.e.  with  hydrogen and deuterium) and/or to the design  activi­
ty being pursued in the  INTOR/NET  frame work. 
These experimental activities  imply  the  availability  of adequate  and re­
liable  systems and procedures, for health protection and safety,  which  can 
not therefore  represent  a research item "a  priori".  It is intended, however, 
to evaluate and to  qualify  by their day-to-day performance in practical  con­
ditions,  systems and procedures such as  : 
. monitoring systems for area and surface contamination, 
. protective  clothing  and equipment, 
. dosimetry and operator surveillance, 
. environmental dispersion and  control, 
. secondary containment clean-up and  gaseous  waste  treatment  systems, 
. procedures for intervention and maintenance, 
. glove-box techniques, 
. tritium-compatible materials and equipment. 
The results  will  help to formulate recommendations for future  facilities 
and in particular for the appropriate design of  fusion  reactor systems. 
2.6 Description  of the Laboratory 
A layout of the laboratory is shown in  Fig. 2. It includes: 
- a Hall  for Process Development (HPD.) of about  3000 m  volume, equipped 
for housing glove boxes  with  inert atmosphere to  allow  the installation 
and handling of equipment containing large amounts of  tritium.  The  H.P.D. 
will be air-tight.with.interlocked doors and be kept  slightly  below at-
. mospherlc  pressure  by an appropriate "hot ventilation" systems. In  case 
9f abnormal  tritium  release  from  the secondary containment to the atmo­
sphere  of the  H.P.D. , an Emergency Clean-up  (E.C.U.)  system, automatical­
ly actuated,  will  be able to reduce the  tritium  concentration in the ope­
rational  area and the  tritium  release to the environment to permissible 
levels.  The  H.P.D.  wi^l include an additional  facility  (expanded  area) 
consisting  of a  500 m  steel box  fully  tight and separately ventilated. 
This  will  provide a containment for large components and  will  allow  ex­
perimental  tests  on their detritiation and handling. 
3 
- A set of laboratories (about  450 m )  for experiments  with  low-tritium 
inventory. 
- A laboratory for tritiated  liquid  waste conditioning and containing the 
gaseous  waste  treatment  (G.W.T.)  systems. 

I. 12 
In order to comply  with  the existing regulations for nuclear  facilities 
the laboratory  will  also include health physics room, health physics store 
and monitoring station. 
2.7 Location 
The utilisation of the  ESSOR  dome as tertiary containment implies some 
draw-backs  that  might easily off-set the advantages of having the  building 
available  with  air conditioning ventilation and stack. The huge volume of 
the dome (2 38.000 m ) and the  nature  of construction materials  would  re­
quire the realisation of another containment  shell  inside the dome in order 
to keep the emergency clean-up  within  a reasonable size both for investment 
costs and operational  efficiency.  The utilisation of the  ESSOR  bunkers  would 
appear practical  only  for the installation of an already  full  developed and 
well  defined  pilot  installation to be operated remotely (such as cryogenic 
or water distillation) due to the  difficulties  for access to the bunkers 
and their particular  shape  (unproportionally high  with  respect to their 
surface area). 
A third potential area  within  the  ESSOR  complex is situated near the stack and 
the reactor dome. Two possibilities exist. One of  these  would  allow  the incor­
poration  of the already existing  building  for the Super-Sara power supply sy­
stem. The second  possibility  is to rationalize the lay-out by constructing a 
new building  near the stack (Fig. 3). 
2.8 Licensing  Procedures 
According  to the Italian regulations for nuclear installations, the licensing 
procedure for a laboratory handling bulk tritium amounts beyond 100 Ci is 
normally  established by  Article  55 of the LAW  D.P.R.  13.2.1964, n.185  which 
normally  implies a strict and lengthly procedure. A favourable situation e-
merges if such a laboratory is built at the  JRC-Ispra,  in the area covered 
by the existing operation license of the  ESSOR  reactor,  utilising  facilities 
of the latter such as stack, gaseous effluents monitoring system, power sup­
ply (including emergency supply), and others. 
In a meeting  with  representatives  from  the legal and technical authorities for 
nuclear licenses in Italy, the former considered  that  a tritium laboratory 
should  therefore be considered as a part of the  ESSOR  complex and as such be / , 
approved by the procedure foreseen for  "Modification  of existing nuclear in­
stallations"  (starting  from  art. 42 of  D.P.R.  185). In this case the authori­
zation  for the construction  could  possibly be obtained  within  less than one 
year. 
Another  advantage is  that  the radioactive release  limits  for  ESSOR  remain 
valid  also for the tritium laboratory. Indeed the hazard potential repre­
sented by the total inventory of tritium in the laboratory (100 g)  will  be 
lower  than the existing reference value for the External Emergency 
Plan  of the Centre:  This  plan refers to a maximum credible accident of the 
ESSOR  with  an expected dose of 4 rem to adults  living  in the  nearest  villa­
ge, whereas the maximum release of tritium (even in oxide form)  from  the 80 m 
high  stack of  ESSOR  would  give a dose of less than 1 rem. 

I. 13 
2.9 Planning and resources 
Starting  from a positive  Council  decision, the  following  steps  can be envisa­
ged for the realisation of the tritium laboratory: 
A) Preparation of the preliminary safety report to be transmitted  with  the 
request  for  "Modification  of the  ESSOR  license" to the  Ministry  for In­
dustry and Trade; 
B) Approval  of the  request  by the safety authorities  together  with  identifi­
cation  of safety relevant items ("progetti particolareggiati" ); 
C) Executive design of the laboratory (excluding experimental equipment); 
D) Construction and mounting of structures and components; 
E) Functional testing and commissioning; 
F) Experimental equipment preparation. 
It is intended to perform the  phases  C to E by an Architect-Engineer  firm 
from  outside, the task of  which  will  be to  prepare  the laboratory, excluded 
the scientific equipment. 
The JRC staffing required for the supervision of such a contract and for the 
Implementation of  phase  A and B  will  be  limited  in the  initial  period to two 
professionals and an  adequate  number of technicians.  With  the completion 
of the construction of the  facility  and its  cold  commissioning, other techni­
cians and operators  will  be added, up to a maximum of about  14  persons,  inclu­
ding  mechanics, electronics and ventilation specialists. 
In parallel the design, preparation and testing of research specific equip­
ment  will  be performed (point F) . This task  will  be assigned to the groups 
already  involved  at JRC in the  measurements  with  H/D to evaluate processes 
for Tritium  Handling (Fusion Safety Programme). 
On this basis the total staffing when the laboratory is in operation is eva­
luated in a maximum of  25 persons,  7 of which  will  be professionals. These 
figures do not take into account, of course, scientific and technical groups 
from  outside  that  might be interested in performing specific experiments in the 
Ispra laboratory. 
In Table  2 a break-down of the schedule related to the  phases  A to F is given 
' together  with  that  of the overall manpower accounted for the laboratory (ta­
king  into account the  JRC's  supports and overhead, a factor of  2.2  has been 
applied  to the above indicated purely laboratory staff). 

I. H 
TABLE  2 
TIME  AND MEN POWER  RESOURCES  FOR  TRITIUM  LABORATORY 
1 2 Years 
3 4 5 
A 
B 
C 
1 
D 
E 
F Prelim.  Safety report A 
P.S.R.  approval 
Executive  design - contractor select. 
- project elab. 
Construct  + Mount - construct, selectis. 
- construct + mount 
Funct.  testing 
Experim.  equip. - projects 
- preparation 
- cold,  tests 
- install., hot tests B« 1 
F*  C-» ' 
D — 
D«  
——· E< • · 
Men power - A to E 
- F 8 
6 20 
6 20 
10 22 
17 30 
25 
Total 14 26 30 39 55 
A preliminary cost evaluation is being  made  by an external engineering 
bureau. The results  will  be available by September 15th. 
2.10 Conclusions 
The analysis . performed shows  that  the problems raised by the  presence 
of tritium in the fuel cycle of fusion  reactors  are complex. 
The data  and information already available or expected from classified work 
do not seem sufficient to solve  these  problems. A vigorous experimental pro­
gramme is needed to  meet  the requirements of NET and  those  of large scale power 
reactors. 
On this basis it has been shown  that  a laboratory for tritium handling at 
0RC could give a significant contribution to the European effort in this area. 
The main objective of such a laboratory would be  that  of providing a basis to 
test  components and operation procedures in support of the NET design and con­
struction. The objective could be achieved in the  following  way : 
- by making experimental verifications of concepts related to fuel cycle and 
reactor operations, according to proposals which  will  be  made  by the Euro­
pean Associations and by the NET team; 

I. 15 
- by  pursuing  the  tritium-oriented activity started  at  JRC  in the  frame­
work  of its  fusion technology programme. 
This  will  imply  a close collaboration  with  the  European fusion research 
centres  and the  NET team starting  with  the  design phase,similar  to that 
being  done  in the  other  areas  of the  Fusion Technology activity where 
JRC is  involved. 
In doing this, particular attention  will  be  payed  to acquiring experience 
in the operational  safety  aspects  related  to  tritium,  a task  which  is in 
harmony  with  the  general vocation  of the  JRC,  as it has  been recently appro­
ved by the  Council  of  Ministers. 

The small 
arrows indicate 
losses  due to 
leakage  / and 
decay  V 
In a seiJ'-iUstaming re; 
Di · Pi - pi > q 
Fig.l  : Scheme  of the  tritium cycle 
in a fusion  reactor Fig. 2 : Tritium  laboratory  lay-out 

Fig.  3 :  General lay-out of the  ESSOR  complex 

I. 18 
3. MAIN REPORT 

I. 19 
3.1 Remarks related  to fusion  reactors  as breeder reactors 
A 1 GWe fusion power plant  will  bum 120 kg of  tritium  per  year. The total production 
of tritium from  CANDU  reactors  in 1980 was (~ 6  GWe installed)  1.4  kg. It is  clear  that, 
if fusion power  will  have  to play  a significant role  in the  energy economy,  it must  be self-
sustaining.  Each  reactor must produce more tritium than  the  amount needed  to fuel the 
plasma  in order  to be  self-sufficient.  The  excess tritium  will be required 
. to  return  the  initial  supply after some years  of operation; 
. to  compensate  for  losses  due to  leakage  and  radioactive decay; 
. to  supply tritium  for new  power plants. 
The tritium breeding ratio (TBR) required  to achieve  these  goals  is ideally 1.05. There  are 
reasons  to believe  that  this value  is not  easy  to obtain,  and  will heavily affect  the  overall 
design  of the  reactor. The basic  parameters  of the  tritium cycle  of INTOR  and.  of a  power 
fusion  reactor  are  summarized  in Table  1. 
TABLE  1 - Tritium  balance  in fusion  reactors 
Tritium  flow Inventory Tritium  burned Tritium  breeding 
rate (g/d) (kg) (kg/a) ratio 
INTOR1 1600 5-6 12 0.6-0.7 
Power  reactor 2 8000 10-20 120 1.05 
1 stage III,  50%  availability. 
2 1000  MWe,  80%  availability. 
The main units  and  processing  steps  of the  fuel  cycle  of a  fusion reactor  are  shown 
in Fig.  1. 
The rectangles indicate  the  processing  steps,  and the  circles  the  units  of the  reactor 
where tritium  will  flow.  The  arrows from each  part  of the  system indicate leakages 
and radioactive decay.  The  requirement  of a  TBR  greater  than unity  means  that  the 
flow rate  into  the  storage  exceeds  the  flow  rate  from  the  storage  into  the  plasma, 
allowing  for  leakage  and  decay.  In a  military plant  for  tritium production  there  is no 
plasma chamber,  no exhaust processing,  nor  recovery from coolant. Furthermore,  there 
is no constraint  similar  to that  on the  TBR. 
In both  cases  the  optimisation  of the  tritium cycle requires  minimizing  the  radioactive 
hazard  and the  leakages,  but in the  case  of a  fusion reactor,  the  constraint  of TBR  > 1 
may be achieved only  with  a careful reduction  of the  radioactive decay,  which  means 
minimizing  inventories  and  reprocessing times. 
This  rough picture shows  that  there  are  many physical phenomena, components, 
processing units,  in the  tritium  fuel  cycle  of a  fusion reactor,  that  have  no equivalent 
in the  technology developed  so far, and  that  require specific  R&D. 

The small 
arrows indicate 
losses  due to 
leakage  j/ and 
decay y 
In a self-sustaining  reactor 
pi + p2 + p3 > q 
- Scheme of die  tritium  cycle  til a fusion  reactor. 

I. 21 
3.2 NET and power reactor requirements related to tritium 
3.2.1 Introduction 
In the Next European Torus the presence of tritium  will  play in important role. 
Therefore it is important to identify the most relevant problems related to tritium 
"n such a type of machine and evaluate the research and development needed during 
the various  phases  of design, construction and operation. The solution of the pro­
blems related to tritium for NET  will apply to a large extent also to the fusion power 
reactors. Indeed, all the main tritium systems of  these  power stations  will  already be 
present in the  NET-machine.  For some problems, e.g. the permeation through the 
first wall  and blanket structures, differences  will  arise  from  the operating temperature 
(higher in case of a-power reactor) or  from  the irradiation time. A scale-up of the 
basic components, such as pumps, valves, etc., is also to be expected. However, the 
research related to NET should, enable to extrapolate the results to the conditions 
of a power reactor. 
A comprehensive assessment,of..these problems has been performed during the  INTOR  (refs.l:2"> 
design  study. Therefore, we  will  largely refer to this work to identity the  critical 
issues and the required research and development. 
The main topics to be investigated can be subdivided as  follows: 
. tritium permeation, inventory and outgassing in structural materials (first  wail 
and reactor components); 
. systems related to the various  aspects  of the  fuel  cycle  (tritium recover/  from 
exhaust plasma, coolant and. breeder;  fuelling,  waste purification and disposal); 
. reactor contamination during standard operation, maintenance and accidental con­
ditions; 
. waste processing and disposal. 
In the  following  an outline  will  be given of the state-of-the-art in  these  areas;  then 
the experimental effort and the related time scale needed to meet the  target  re­
quirements for the engineering design and construction of NET  will be evaluated. 
3.2.1  Tritium  permeation and inventory in structural materials 
Tritium  permeation through first  wall,  limiter  or divertor collectors is governed by the im­
plantation  of the charge exchange neutral  flux  in the near surface zone, by the recombina­
tion of these  atoms into molecules at the plasma/first  wall  and coolant/first  wail  interfaces 
and by  diffusion  in the bulk of the  wall. 
Where  no tritium implantation is present (e.g. in the breeding blanket  cooling  ducts), 
the modellisation of the phenomenon is somewhat simpler, since the permeation is . -.. ..— 
governed by  better  known physico-chemical phenomena. 
In the models adopted for calculation of the implantation of charge exchange neu­
tral flux,  the surface boundary condition is characterized by a recombination constant 
which  depends on the activation energy for migration, the  heat  of solution and the 

I. 22 
Slevert's  constant The surface cleanliness is taken into account by a molecular 
sticking  coefficient. On the wail-coolant side, the presence of oxide layers and 
hydrogen  produced by radiolytic decomposition of water originates complex sur­
face conditions  which  are approximated by recombination  limited  kinetics or 
diffusion  limited  conditions. The  diffusion  inside, the layer is described by a tem­
perature-dependent  diffusion  coefficient and also considering a thermal gradient 
(Soret effect). Neutron damage of the  wall  is taken into account  with  a trap  con­
centration and. a detrapping energy. 
In INTOR,  the calculated permeation  rates  for the first  wall  (SS-wall  surface: 400 mJ, 
Tinner  200 -  300°C,  TjUter:  100°C,  t = 0.5 - 1.3 cm) cover a broad range of values, 
from  10"- to 0.2 g/a. The highest permeation  rates  are obtained for dirty inner sur­
face and clean outer surface (Le. low recombination  rate  on plasma side and high 
recombination  rate  on coolant side). 
In the same way, the. tritium, inventory in the- first  wail  depends on the recombina-
tive and trapping effects- In the  INTOR  case, values  from  0.7 to 7 kg have been 
evaluated.  Both  tritium permeation and Inventory decrease  with  the  wail  thickness. 
Variations  in duty cycle have  small  influence. As far as the divertor/limiter  collec-. 
tors are concerned, tritium permeation and inventory vaiues have been found of 
the same order of magnitude as in the previous case. 
The relative importance of the tritium permeation and tritium trapping in the  wall 
is strongly related to the temperature of the outer (coolant side) surface.- In the 
INTOR  case, this temperature is low  (100°C)  whereas for a power reactor it  will 
be much higher (~  300°C)  so increasing strongly the tritium permeation into the 
coolant  and reducing possibly the inventory. In the case of NET the problem of 
first wall  coolant temperature is  still  open, a value between  100°C  and  300°C 
could  be the selected one.  This  calls for experimental research over a large range 
of temperatures both on the inner and outer  wall  side. 
The question of tritium outgassing during the  dwell  time of the pulsed operation 
?.nd after the reactor shut-down has been recognized as being one of the most  critical. 
The release of D and T during the  dwell  time  will  influence the vacuum conditions 
and then the pumping requirements. The vacuum chamber has to be pumped down 
to the pre-shot pressure (~ 10"5 torr); at the same time the gas release  from  the 
walls  has to be reduced to a  level  low enough  that  it does not appreciably influence 
the D-T species mixture during current  initiation  and ohmic heating. Tae calculations 
performed  during the  INTOR  study have shown  that,  depending on. the hypotheses 
of the mechanisms for gas desorption and surface conditions,  these  requirements 
can or cannot be satisfied. The same uncertainties apply to the case of reactor shut­
down.  The outgassing  rate  knowledge is fundamental to establish the operations 
(waiting  time, baking, pumping) before removing the reactor components inside 
the vacuum vessel. A preliminary calculation has shown  that  the quantity of tritium 
retained in the first  wall  could  be significantly  reduced by outgassing. A  350°C  baking 
for 100 hours should release about 90% of the tritium trapped in a 13 mm SS first 
wall.  However, neither recombinative, nor trapping effects were taken into account 
and these  results must be reviewed. 

I. 23 
3.2.3  Tritium  processing systems 
Vacuum  requirements and pumping 
A number of vacuum systems  will  be needed in NET and power reactors: many of 
them  will  be placed in a tritium environment.  This  will  apply in particular to the 
plasma  exhaust, plasma heating (neutral injection) and tritium recovery  from  blanket. 
In INTOR  Phase-1 an evaluation of the pumping requirements for the torus vacuum 
system was performed (ref. 3).  This  system must perform the  following  operations: 
. to pump down the torus chamber to an  initial  base  pressure of  Torr  before 
the first  start-up  and to a  base  pressure of about 3-10"5  Torr  during  dwell  time; 
. to exhaust a gas quantity equal to  that  for refuelling during the burn time. 
A nominal pumping speed of 104' Is"1 was calculated as needed for cleaning before 
start-up  whereas during  dwell  and burn time the required pumping value was 
=s 4· 10s Is"1.  Assuming  12 pumping units, each of them should have a speed of 
= 3-104" Is"1. Concerning the pump selection, the suggestion was made to use  com­
pound  cryopumps equipped  with  microporus sorbents on the cryosorption panels 
for the plasma exhaust during the burn-time, whereas pumping before  start-up  and 
during  the cleaning process  could  be provided by turbomolecuiar or mechanical 
pumps. For the cryopumps the most important problems to be solved are: 
. the He-pumping  rate  (compared to  that  for D and T); 
. the  heat  load  and radiation effects; 
. the tritium storage in the cryopanels. 
For turbomolecuiar pumps, the problems  will  regard the industrial scale-up (the 
maximum  speed is now 3.5· 103 Is'1), the effects of the electromagnetic  field  and 
the tritium compatibility (no  oils  and greases). The same applies to the mechanical 
pumps. The other vacuum systems  will  call  for similar  requirements on the pumping 
side. 
Plasma  exhaust and neutral beam processing 
Various  schemes have been proposed to process the plasma exhaust pumped-out 
through the divertor/limiter ducts  from  the vacuum torus chamber. Three are the 
main  steps  in this process: 
. gas purification; 
. He-separation; 
. . hydrogen-isotopes separation. 
A flow-sheet for plasma exhaust and neutral beam processing systems has been pro­
posed during the  INTOR-Phase  HA study (ref.4). In this scheme the two  streams 
coming  out  from  the torus and the neutral beam injectors are  purified  on cold  traps 
(palladium-silver  membranes or  getters  may also be used). H, D, T are  liquefied  in 
a condensor, whereas gaseous He is separated and  sent  to other processing lines 
(e.g. the blanket recovery system). The hydrogen isotopes are separated in cryodis-
tillation  columns  which  can be connected to other process systems, such as the 
blanket recovery and the coolant reprocessing. 

I. 24 
.Alternative  modes for plasma exhaust reprocessing have been proposed (see chapter 
where alter the separation of the impurities, the exhaust plasma is  oxidized  to 
(H.D,T) ?0 in a catalytic recombiner and fed to the hydrogen isotopes separation 
system. Here, the  liquid  stream is fed discontinuousiy to a separative electrolytic 
cell, where a protium-enriched fraction of the isotopes is electrolysed. The  gaseous 
isotopes are separated by gas-chromatography. The non-eiectrolysed residue consti­
tuted mainly by DTO is  sent  to a second electrolytic  cell,  operating at low  liquid 
inventory  for its transformation to D,T. 
Fuelling 
In INTOR  fuelling  into the plasma chamber can be provided by gas puffing and/or 
a pellet injection system. Twelve ports are available for the gas injection.  Both  con­
tinuous and pulsed  flow  control of the selected gas are provided. Two pellet  fuel 
injectors are located on opposite sides of the torus. Remotely controlled selector 
valves  determine the  fuel  gas admitted to the liquefiers,  which  are located inside the 
insulated  fuel  injector. The design pellet injection velocity is 2-103 ms"1 (.pneu­
matic or centrifugal injectors can be used, depending on their development). 
Tritium  recovery from blanket 
Solid  Breeders 
For the  INTOR  reference design  (Li 2C as a breeder at 400 -  600°C,  water as a 
coolant), tritium may be extracted using helium as a sweeping gas (ref. 4). The pro­
cess envisages a countercurrent  cooling  of the helium  which  emerges  from the' 
blanket at =  600°C,  containing tritium at = 0.1 Pa of partial pressure. In the ana­
lysis  the hypothesis was made to recover  daily  all the tritium bred (50 g/d); the 
corresponding helium  flow  rate  was calculated to be 7800 m3/h, subdivided among 
the twelve blanket  segments  considered in the design.  Tritium  is first completely 
oxidized  on a catalytic bed (=  200°C).  In a  subsequent  step  it is cooled to  liquid 
nitrogen  temperature  to trap  tritiated water as frost.  Helium  is heated to s  400°C 
before re-entering into the blanket. The trapped tritiated water is then recovered 
in liquid  form and decomposed in an electrolytic  cell.  Tritium  is sent  for purifica­
tion to the plasma exhaust processing line and oxygen is  sent  back to the catalytic 
bed. 
This  scheme has been developed for the  case  where helium is used only for tritium 
recovery, whereas the  cooling  is afforded by water. Proposals to use helium both 
for cooling  and tritium recovery have been also worked out, in particular for  appli­
cation  to power reactor systems (ref. 5). In this  case  a similar  approach for detri-
tiation  of the gas can be devised, even if the difficulties are expected to be larger 
as a consequence of the higher  pressure  and  mass  flow  of the helium coolant, the 
lower  concentration of tritium and T2 O in the coolant and the advisability of 
avoiding  cooling  the helium to very low  temperatures  in the extraction system in 
order to maintain a reasonable thermohydraulic efficiency for electricity generation. 

I. 25 
Liquid  Breeders 
Various  schemes for tritium recovery from  liquid  breeders  have been proposed, in 
particular from  Lil7Pb83  which  looks attractive for  NET-applications.  These include 
either a direct pumping in vacuum from the breeder or the use of a purge gas in 
contact  with  the  liquid.  A more detailed description of the  advantages  and disad­
vantages  of these  methods and of the research needed to  demonstrate  them is given in 
chapter  5. 
• Coolant detritiation 
During  the  INTOR  assessment  several processes for water detritiation have been in­
vestigated. The most appropriate are described hereunder: 
a) Vapour Phase Catalytic Exchange/Cryogenic  Distillation  (VPCE/CD).  Tritium  is 
transferred from light or heavy water to elemental form in a catalytic exchange 
reactor. Cryogenic  distillation  is used for the isotopic separation of the ele­
mental forms. 
b) Direct  Electroiysis/Cryogenic  Distillation  (DEL/CD).  Water is electrolysed, to 
hydrogen and oxygen. Hydrogen is  distilled  in a cryogenic column for isotopic 
separation. Detritiated hydrogen is recombined  with  oxygen from the electro­
lytic  cell  and recycled to the coolant loop. 
c) Combined Electrolysis Catalytic Exchange/Cryogenic  Distillation  (CECE/CD).  Here 
electrolysis  and catalysis perform a preliminary isotopic separation and produce 
elemental hydrogen, then the process is completed by cryogenic  distillation. 
d) Water  Distillation/Electrolysis/Cryogenic  Distillation  (WD/EL/CD).  Water  distilla-  • • 
tion produces slightly enriched water to an electroiyser. Hie elemental hydro­
gen is fed to a cryogenic  distillation  system for the  final  isotopic separation. 
A comparative analysis of  these  processes has been performed, assuming a tritium 
permeation of 103  Ci/d  (0.1  g/d) through the first  wall  and a tritium concentration 
in the coolant of  1 CI/1.  Correspondingly, the volume processed was  1000  1/d. 
The results of the comparison are as  follows: 
. the  CECE/CD  and the  WD/EL/CD  seem the most promising processes from an 
economical  point of  view; 
. CEC/CD  can detritiate water to low levels  which  could  be unrestrictedly released 
to the environment, and accepts feeds  with  different tritium concentrations  which 
should  permit to process  together  several tritium systems. The disadvantages are 
a limited  operating experience and the necessity to  operate  in the front end tritiated 
water at higher concentrations than the other processes; 
. WD/EL/CD  has many years of commercial experience in heavy water production. 
This  process can accept feeds at different tritium concentrations. In the design 
option  used for the economic analysis, water  would  not be detritiated to levels 
allowing  unrestricted release to environment. In this  case  multiple processing  would 
be required; 

I. 26 
. DEL/CD  seems  feasible since  there  is practical experience in low  level  tritiated 
water electrolyses and hydrogen isotopes cryogenic  distillation,  also if they were 
not used  together  on a large scale; 
. VPCE/CD  does  not appear  suitable for low  level  light water purification. It  could 
be of interest in  case  heavy water were the coolant in the  firs:  wall  system. 
2.4 Effects of tritium contamination during standard operation, maintenance and  
accidental  conditions 
J^ac^crJ^U  ji^^uat^n_^stejm 
The determination of the tritium concentration in the main reactor volumes (vacuum 
chamber, reactor  hall,  etc.)  will  be an important item of the NET design and it  will 
represent  the basis to optimize the various components and reactor operations during 
normal  and abnormal conditions. In  INTOR  a rough evaluation of all the sources of 
tritium  in the reactor room has been made  which  has led to the  following  global 
values for the tritium release: 
. standard operation 10 - 30  Ci/day 
. maintenance up to 103  Ci/day 
. accident (single release) up to 105 Ci 
The safety criteria taken during the  INTOR  design study for the tritium contamina­
tion are: 
. 10.000 Ci/a released to the environment during standard operation; 
. 5-10"6  Ci/m 3 maximum allowable concentration for unprotected workers. This 
limit  could  be raised to up to 5-10"*  Ci/m 3 in case  workers are equipped  with 
protective suits. 
Various  strategies  can be devised in order to  meet  these  targets.  They  will  depend 
also on other hypotheses, in particular on the minimum time required for human 
intervention  after the reactor shut-down. In setting-up  these  targets,  one has also to 
take into account the other sources of radioactivity, mainly  those  related to neutron 
activation. 
The first evaluations made during the  INTOR  study (reactor  hull  volume s  1.5-103  m3 
have shown  that:  i 
. during standard operation, if the release  will  be a 10 Ci/d, the reactor room air 
could  be ventilated directly to the stack; 
. for higher tritium standard leakages, an air detritiation system is needed. Such a 
system  will  consist of blowers,  pre-heaters,  catalytic reactors, after-cooler and water 
removal  systems. The required  flow  rate  and cost was evaluated as a function of 
the tritium concentration  level.  Typically,  for a release of 30 Ci/d, a cost of 75 MS 
has to be expected for the air detritiation system to  meet  the conditions for worker 
in the reactor  hall  (5 juG/m3). A reduction of a factor of 10 in cost was found 
if the concentration  level  could  be increased to up to 500  ,^Ci/m 3. The correspon­
ding total costs for a ten-year operation  would  be = 200 and 16 MS, respectively. 
This  is an incentive to reduce as much as possible the air detritiation volumes, for 
example by subdividing the reactor  hall  in various rooms. 

I. 27 
. In case  of maintenance or accidental conditions, the air detriation system required 
for enabling the worker's access  within  24 hours after the reactor shut-down 
results too expensive and impractical. Such a short time delay  would  also require 
a torus extra-shield of 1.5 m thickness, as compared to  that  strictly needed for 
superconducting  coil  protection  from  neutron and gamma irradiation.  Alterna­
tively,  one  will  have to rely on protective suits for the workers  (which,  however, 
will imply  more  limited  efficiency) or on remote handling operation. 
Plasjjia  J/£cuu  m^£sjjej_  j;o_nt  arnjrurrion  ^t^majn  t^n^jicej^^nj  e_rn  als_ 
As mentioned before, the plasma vacuum chamber  will  be contaminated after the 
reactor shut-down as a consequence of the tritium desorbed  from  its surfaces and 
of the neutron activation of its structures. In  INTOR,  40 - 100  Ci/cm 3 (107 rem/h 
of contact dose) are expected in the first  wall,  an average value of 10  Ci/cm 3 in 
the blanket structures,  with  a total shut-down activity of 109 Ci.  Possibly  a part 
of this activity  will be associated to dusty material (1 mm erosion of  INTOR  first 
wall implies about 3 tons). The extreme  working  conditions of the plasma vacuum 
chamber  will  require automated and remote inspection and maintenance systems. 
They  must be capable to reach the whole inner surface of the plasma chamber to 
carry  out "in situ" repair and to substitute  small  pieces.  More  important repairs or 
the complete substitution of reactor components  will  require the disconnection of 
large and heavy (up to 150 tons) structures and their transport to hot  cells. 
It can be shown  that  the most  difficult  items are: 
. control and repair systems for the reactor inner components  with  visualisation 
capabilities  and/or  with  fully  automated operation; 
. disconnection/connection of welded and bolted flanges; 
. transport and precise positioning of large and heavy pieces. 
All the previous interventions  cause  a loss of containment and the design philosophy 
for maintenance must reduce to the minimum possible the spreading of contamina­
tion due to tritium and erosion products. Possible solutions are to employ provisional 
tunnels or transfer flasks between the containment enclosures (e.g. torus, hot  cells). 
Waste processing and disposal 
Gaseous  wastes 
Gaseous tritiated  wastes  coming out  from  the reactor system (plasma chamber, 
neutral beam lines, blanket, tritium  storage)  must be detritiated before their release 
to the environment. The processing system must be apt to  treat  organic compounds. 
The process proposed for  INTOR  foresees an uranium bed (s  600°C)  where water 
and organic compounds are decomposed in uranium compounds and hydrogen. 
Hydrogen  diffuses through a palladium  silver  membrane and it is  sent  to the plasma 
exhaust reprocessing system. The detritiated  gases  are released via the stack. 

I. 28 
Reactor  components  conditioning 
Maintenance  of reactor components  in the hot  cells  will include decontamination 
procedures  and  methods  to extract soaked tritium.  In particular, thermal processing 
under vacuum  can be  carried  out at  higher  temperatures  than  those  allowed  in the 
torus.  Final  disposal  of solid  waste  can  require volume reduction  by compactation 
or melting. 
3.2.5  Tritium  inventory 
In Table  1 is predated  a break-down  of the  tritium inventory  of INTOR-Phase Ha, 
for the  case  of solid  breeders.  A more detailed analysis  of the  single items should 
consider  the  following  points: 
. for the  plasma exhaust system,  the  main contribution  to the  inventory comes 
from the  cryodistillation  (120  g), from  the  cryopumps  (120  g) and  from  the gas 
storage  (120  g) units; 
-. the  process system  of blanket  itself  involves about  1 kg of  tritium,  the  main con­
tributions coming from  the  liquid  nitrogen  cold  trap  (700  g). This figure applies 
to solid  blanket; 
. the  fuelling  system  has an  inventory  of 200 g, but  this is considered  an upper  limit; 
. for  liquid  breeders,  the  inventory  will be of the  order  of 100  to 200  g. 
During  the  INTOR  workshop  an attempt  was  made  to distinguish between  the so-
called  vulnerable  and  non-vulnerable inventory.  A tentative subdivision  was  1.4 kg 
(vulnerable)  and  2.9 - 4.3  kg (non-vulnerable).  In the  safety analyses  no more than 
10 g were considered possible  for  release  to the  environment from  a single accident. 
However,  this figure  seems  today  too  optimistic. 
Table  1 - Tritium  inventory  (solid  breeder)  - INTOR-Phase  lia 
Plasma  exhaust  and  neutral  injectors 
Breeding  tritium 
Primary  coolant  detritiation 
Wastes 
Atmosphere 
Neutral  injectors  and  torus  feed 370 g 
746 g 
65 g 
not estimated 
1 o 
245 s 
Storage 
Breeding  blanket 
First  wail 1427 g 
2300  g 
500-  1000 g 
100- 1000 g 
Total 4330- 5730  g 

I. 29 
3.2.6  Time  scale  requirements for  R&D 
Assuming  for NET: 
. beginning of engineering design 1987 - 1988 
. beginning construction 1992 - 1993 
. start operation 1998 - 1999 
the following time  scale  for R&D related to tritium problems can be evaluated: 
. basic information on H/D simulation data for permeation, fuel cycle  processes, 
etc.: 1984 - 1986; 
. check of  these  data with  tritium:  1986 - 1989; 
. large components development:  1985 - 1993; 
. tests  of NET sub-system with  tritium:  1988 - 1993; 
. development of  waste  disposal  systems:  1989 - 1995. 
References  . 
1. G. Casirü  (EC),  D. Leaer_(EC),  Rogers  (USA),  G. Shatalcv  (USSR),  T. Suzuki  { apan),  INTOR, 
Phase Il-a,  Final  Report,  Ch.8:  Tritium  and Blanket, under publication. 
2. European  Contribution  to LNTOR,  Phase Il-a,  Ch.8,  Tritium,  Blanket and Safety, edited by 
G. Casini,P.  Rocco,  EUR  FU  BRU/XII-132/82/EDV  30 (December  1982). 
3. European  Contribution  to LNTOR,  Phase I,  Ch. XI,  Vacuum  and  Tritium  Systems, edited by 
D.: Leger  and  F. Reiter,.  FUR  FU  BRU/XII-132/82/EDV  2 (April  1982). 
4. D. Leger,  M.H.  Plouzeniicc,  INTOR  Tritium  Systems  CEA  report  DCAEA/SECF/82-103/DL/AM, 
June  1982  (also  reported in ref.  2). 
5. Conceptual Design of an Electricity Generating  Tritium  Breeding Blanket Sector for  1NTOR/NE1 , 
edited by  Bond  and P. Reynolds,  NET-contract  085/81-12/FU (UK), final report (May  1983). 

I. 30 
3.3 Review  of the  existing know-how  and of the  activities  in progress 
The available knowledge  on tritium problems  and in  particular  on tritium technology 
is mainly based  on the  operation experience  of fission  reactors  (in  particular heavy 
water  and  HTGR  reactors)  and on the  confinement  and  conditioning  of waste water 
arising  in fission  fuel  reprocessing plants, whereas  the  information coming from  the 
military  applications  is for the  moment  still  limited. 
In the  following  the  existing knowledge  on tritium  is briefly  revised  and the  main 
research activities related  to fusion  in progress  or scheduled  in Europe  and  throughout 
the world  are  outlined. 
3.3.1  Experience from  fission  reactors 
In heavy water plants, tritium  is produced  in the  moderator  by neutron absorption. 
This  implies  the solution  of a  number  of problems related  to the  handling  of large 
volumes  of tritiated water  and to its  processing. Important experience  has  been 
gained both  in Europe  and  outside  in the  world  (mainly  in Canada)  on these  pro­
blems.  A detritiation system from heavy water capable  to extract  up to  160,000 
Ci/a of tritium  is in  operation  at the  high  flux  reactor  of Grenoble, based  on a 
process developed  by  CEA.  It applies cryodistillation  of elemental hydrogen  iso­
topes  for their separation.  Due to the  know-how acquired  by the  operation  of this 
facility,  this technique  represents  todav  the  first choice  for  tritium separation 
and recovery  in fusion reactors.(ref.  1). 
In HTGRs  tritium  is formed  in the He  primary coolant from  which  it permeates 
partially  through  the  heat  exchanger walls into  the  water  of the  steam generator  at 
levels  which  can be  kept  within  acceptable  limits  without particular purification 
procedures.  In the HTGR  concept  as heat  source  for  energy carriers (hydrogen, 
hydrocarbons),  the  tritium permeation  rate  was  required  to be  reduced  for  safety 
reasons. Tests performed  at KFA  Jülich  have shown  that  by coating ferritic steel 
with an oxide layer,  a reduction factor  of about  100  was  obtainable. These oxide 
films  showed, however,  a low stability  with  temperature  changes. Better behaviour 
was found  with  oxide layers on Incoloy. 
Tritium  is also a by-product  of nuclear (fission) power generation, where  it becomes 
a concern  for  LWRs  at the  spent  fuel  reprocessing  stage.  Table  1 gives  an overview 
of T inventories  and  waste volumes arisings  in a  reprocessing  facility.  Research acti­
vities  on the  treatment,  conditioning  and  disposal  of tritiated waste waters have been 
shared  in the  member countries  of the  EC by the  "Indirect  Actions  Program"  of the 
CEC in the  frame  of the  "Management  and  Storage  of Radioactive Waste" program 
(DG XII),  as summarized  in Table  2. 
Reference 
1. J  . Chatoux,  W. Eisermann,  Atomwirtschaft,  Vol.  14 (1969), 25-31. 

I. 31 
Table 1 - Assumed data for the  description  of  3H-waste  management  alternatives 
Reprocessing  capacity tU/a 1500 
3H-inventory of the  spent  LWR-fuel  elements mCi/a 1 
3H-inventory of the cladding  hulls MCi/a 0.5 
3H-inventory of first  solvent  extraction  cycle MCi/a 0.5 
Volume of tritiated  waste  water  (16 mCi/1) m3/a 32-103 
Volume of tritiated  waste  water  with  recycling and 3H-scrubbing 
(166 mCi/1) m3/a 3000 
0/7 -  activity  of the tritiated  water  (300  m 3/a) mCi/1 0.1 
Heavy  metal  inventory  of the tritiated  water  (uranium) g/m3 1 
(plutonium) mg/m3 1 
(water/cement)  ratio — 0.5 
Table  2 - Research  contracts  on tritium bearing  waste  from  reprocessing 
Program Contract Nr. Contractor CEC  cuntrib. Title 
period KECU 
75/79 090-78-7  WAS B CEN  Mol 371· Separation of T from aqu.  effluents 
80/S2 WAS-159-B(D) CEN  Mol 393· 
83/84 292· » 
75/79 088-78-1  WAS D KFK  Karlsruhe 
TU Karlsruhe 83' T-enrichm. on hydrophobic catal. 
75/79 046-77-10  WAS D UKAEA  Harwell 25· T Immobilisation 
75/79 075-79-7  WAS UK 30· T immob. by incorporatin in inorg.  solids 
80/84 KFK  Karlsruhe 58· Characterisation of  conditioned  cladding 
material (T  release) 
3.3.2  Activities  in support to  Fusion 
Eujop_ ean_ activi  ties 
Design  studies for JET operation  with  tritium 
The Joint European Torus is scheduled to operate in its second phase  with  tritium 
in order to investigate the thermonuclear plasma regime. For this it is intended to 
bum  a 50/50 at% mixture of deuterium-tritium; the planned number of shots  with 
D-T plasma is 10".  Considering  the low burn-up of  fuel  and the problems related 
lo an out-site reprocessing of it, it was decided to process the spent gases  in-situ 
after each shot before being reinjected in the torus. So a  tritium  plant  will  be in­
stalled  in the JET area  with  the  following  main  functions: 
. to  purify  the gases  coming  from  the plasma chamber and  from  the neutral injectors; 
. to make their isotopic separation. 

I. 32 
The plant is characterised by the  following  operational requirements which strongly 
influence design: 
. low throughput of  about  2 kg during  life 
. short  life of about  2 years 
. unchanging dr.ty 
. parallel circuits permit replacement without interrupting operation 
. active handling of components and  waste  disposal are available off-site 
. small size 
. cryogenics, ultra-high vacuum and  remote  handling facilities are all available  within 
JET. Cryogenic dumping  will  be used extensively. 
The gas purification  will  be performed by  means  of cold  traps  which  will allow to 
eliminate  gaseous  impurities without making  solid  wastes. 
The isotopic separation by gas chromatography has been chosen in order to obtain 
the best  process efficiency and the lowest tritium inventory. 
The whole processing unit, including the gas analysis  apparatus,  will  be placed in 
glove-boxes. The glove-box  atmosphere  will  permanently be processed by two clean­
up systems. The first one is intended for glove-boxes with high contamination such as 
that for repairing and maintenance,  whereas  the second one  will  be related to glove-
boxes with low contamination such as  those  for the process. 
The tritium room  will  be normally ventilated by a routine ventilation and air condi­
tioning  system  whereas  in case  of contamination, an Emergency Clean-up System  will 
be needed. Ail the  atmosphere  processing systems  will  be positioned in "caissons". 
All these  systems  will  be located at different floors of a building which  will be for­
bidden to workers during the pulse operation of the machine. An annex building is 
planned for services, laboratories, process control and safety. 
The JET staff is now engaged in detailed specification for the tritium plant incor­
porating the  following  essential process elements: 
. removal of chemically active impurities from  torus  and neutral injector  exhaust  ga: 
. removal of helium 
. removal of protium 
. isotope separation of T and D 
. storage  of separated  hydrogen isotopes 
. storage  of waste 
. pumps 
. valves 
. instruments for process control and inventory monitoring 
. input/output interfaces 
EUR-association  research  activity 
A research programme to investigate the tritium problems related to fusion power 
reactor development has been launched recently by the Commission in the frame 
of the Fusion Technology Programme  1982 - 1986  set up by the Fusion Technolo­
gy Steering Committee. In Appendix I are indicated the  areas  related to tritium 

I. 33 
for which  study contracts have been proposed by the European Fusion Associations 
together  witli  the state  of the art of the  décisions  about the acceptance of  thèse 
proposais. The  areas  considered are: • 
. fuel  clean-up system; 
tntmm recovery trom waste streams; 
. tritium detector; 
. electrolytic  cell; 
. decontamination system; 
. industrial development of large components; 
. tritium recovery from blanket  (liquid  and  solid  breeders). 
The proposais are based in  gênerai  on previous  expérience  in fission  reactors; in 
some cases,  référence  is made to  spécifie  experiments carried out in the frame of the 
fusion  activity prior to 1983, and to the conceptual design work carried out in 
Europe on fusion reactor Systems. 
Activiries  outside Europe 
Los Alamos labomtory 
The TSTA  facility  has been completed and  tested  with  hydrogen.  Opération  with 
tritium  is scheduled by the end of 1983. The principal objectives of this  facility  are: 
. to demonstrate  the fuel  cycle for DT fusion power Systems; 
. to develop and evaluate personnel and environmental protection  Systems; 
. to provide a  facility  that  will yield  a reliable data  base  for tritium handling Systems; 
. to develop tritium compatible components  with  long-term  reliability. 
A schematic block diagram of the  fuel  supply and exhausted plasma recovery Systems 
is represented in  Fig.  2, indicating  that  the followina  main sub-svstems are reauired: 
. plasma chamber  évacuation  system by  cryogénie  pump; 
. transfer pumping; 
. fuel  clean-up removing ail the impurities (He included); 
. séparation  of hydrogen isotopes by  cryogénie  distillation; 
. waste  treatment  and disposai; 
. fuel  injection; 
. safety control and monitoring; 
. emergency clean-up. 
Other Systems or sub-systems connected  with  the plant are: 
. the impurity simulation; 
. the neutral beam injection; 
. the gas analysis; 
. the  gênerai  data acquisition and control. 

I. 34 
In this plant, in  which  500 moles of DT are treated per day,  three  levels of tritium 
containment are considered: 
. the primary  level  constituted by the material of construction of the  apparatus  con; 
taining  tritium; 
. the secondary  level  constituted by glove-boxes for major tritium handling hardware 
and components; 
. the tertiary  level  constituted by the large volume room maintained at tritium 
concentration below  1-10" 5 Ci/m" 3 by a normal and an emergency (operating in 
case of accident) dpan-up system. 
This  plant, locate^ in an old  buil'ding  of 3000 m3 room volume, should maintain 
a tritium inventory below 200 g. 
Other US laboratories 
Recently,  Mound Laboratory has focused most of its tritium technology development 
on tritium containment and environmental control..The  principal  goals are: 
. to prevent any tritium release to the environment and recover for  re-use  all 
tritium  released  within  the laboratory; 
. to demonstrate the  CECE/CD  water detritiation system. An electrolytic  cell is 
fabricated  by G.E. and has been  working  for a  long  time  with  tritium con­
centration of 10  Ci/cm 3 . 
. to develop a system for injecting into the torus of  TFTR; 
. to design containers for tritiated  liquid  waste. 
The principal  activities pursued at Lawrence L.L. are: 
. the tritium recovery  from  atmospheric release, by using catalytic oxidation units; 
. the development of organic  getters  which,  in the presence of air,  minimize  the 
formation  of more hazardous tritiated water; 
. the measurement and correlation of the physical and chemical properties of the 
hydrogen  isotopes below 30 K. 
At the Princeton Plasma Physics Laboratory, in  view  of the  TFTR  operation  with 
tritium,  a tritium systems area (Fig. 3) has been constructed in accordance  with  the 
requirements of storage of tritium on-site and  with  no attempt  to recover,  purify  and 
re-use  tritium. The supply of tritium, according to the  Mound  Laboratory  tests,  will 
be made automatically and rapidly for a maximum of 400 Curies or 0.4 g.s"1 in 
15 ms.  Many  precautions are taken in order to provide barriers to prevent  passage 
of particulate matter into the torus and means for removal of He3 and impurities. 
Major  components  will  be fabricated  from  316 stainless steel using all welded con­
structions except for the connections required for the maintenance of components and 
for replacement of removable items. The whole area is supplied  with  a clean-up sys­
tem which,  in the main  line,  is similar  to the  TSTA,  but in smaller dimensions as 
the room atmosphere and the tritium inventory are strongly  limited. 

I. 35 
Fig. 2 - Flow  schematic of  TSTA  process system 
Japanese  activities 
A Tritium  Process  Laboratory (TPL) is now under construction and  will  be in opera­
tion in 1985.  It is conceived as a multipurpose  facility,  with  the  following  main de­
sign characteristics: 
. to apply multiple barrier containment and tritium removal systems; 
. to make zoning  division  clear and to  install  proper ventilation systems;  ' 
. to select materials and components suitable for tritium handling. 
The dimensions of the  building  are about  48 x 26 m  of which  the controlled area 
is 1080  m 2. 
In the experimental room,  which  is 15 m  high, the glove-boxes are arranged. The  rest 
of the  building  is composed of an equipment room where various types of tritium 
removal  systems are furnished (Fig.  4). Each  block of glove-boxes has a function of 
individual  ventilation and  pressure  control and is provided by airtight boxes  with 
accommodated vacuum pumps. There are  three  containment systems: The first one 
is composed of the material of the  apparatus;  the second one is the tritium removal 
system  which  is formed by: 

I. 36 
Fig. 3- Tritium  systems-general 
arrangement at  TFTR. 
Fi" 4 layout  of TPL 
(Japan). 1 fdowhoi for  irulvtu 
2 0 nvr ho » for no rip 
3 gjnvcbo* for crvoyrnc ujiullauon 
4 gioveoo»  for  pna« 1 equilibrium muturcmeni 
5 g) nvx bo  » for recover* 
6 ^ o *c bo  « for thermal  diiTuion 
7 JJOVTDOI  for componeni trjt 
Ä pctvKt»n  lor purification 
9 povr bö 1 for lolidificanon 
10 (loveoo»  for  iniium  "rno^aJ  teil 11 evacuatile  »lo^t-bo*  for transfer 
12 e va cu ibi c pJovrhot for transfer 
13 pai·.hot fur loJûina  jnd  uni^adm? 
|4 pii fot telline  upcr>(>):cntc  cniumn 
15 pat for  drajrijpr 
16 hehum  reì'nuerator 
17 conimi  panel 
18 power tuppK 
19 commi pj ne  li 

I. 37 
. a glove-box purification system for removing tritium and oxygen; 
. a pressure  control of each block in order to prevent tritium leak from the glove-
boxes and to avoid the radiation exposure of operators during normal and off-
normal  operation. 
Tae third  containment system is formed of a clean-up system for the room atmos­
phere  during the normal operation. In  case  of accident an emergency ciean-up is 
automatically  put into operation. 
As far as the planned program is concerned, to  date  the glove-boxes in the airtight 
glove-box  room  will  contain units for tritium measurement and analysis systems, 
tritium  storage  systems, cryogenic  distillation  systems, liquid-vapour equilibrium 
measurement and thermal  diffusion  systems, tritium purification and trapping systems, 
liquid  waste  solidification  systems and. tritium containment and removal testing systems. 
Canadian activities 
In Canada a fusion fuels technology project has been set up, focused on supply of 
tritium  and tritium technology,  with  the mandate  to extend and  adapt  existing ' 
CANDU  tritium technology for use in international fusion programs. The production 
rate of tritium in this country was about 1.4 kg/a already in 1980 and it  will 
produce  still  much more consistent stock piles of .tritium in function of its nuclear 
power programme. 
It has acquired an outstanding operational experience and know-how in handling 
of tritium oxide bearing fluids and components, in dynamic and static (leak:-tight) 
containment design, in occupational exposure and environmental dispersion models. 
It has developed to commercial maturity highly efficient monitoring systems, 
protective clothing and remote handling techniques and devices. Two facilities  will 
become operational in the  near  future for T separation from heavy water, the first 
one for commercial use based on the CEA process by isotopic exchange in the 
vapour phase, the second one as a pilot plant based on isotopic exchange in the 
liquid  phase  (Canadian development).  Both  apply cryogenic  distillation  for final  T puri­
fication  and  getters  for long-term T storage. 

I. 38 
3.4 Past and  present  activities at JRC 
3.4.1  Operative experience in tritium handling and monitoring 
Introduction 
The first nuclear  installation  generating tritium oxide bearing effluents has been 
commissioned  at the JRC in  1959.  Since then the monitoring of tritium has been 
carried  out systematically on samples of air and water (environment), on  biological 
materials and on plant effluents.  During  this time period, the procedures for sampling 
and measurement have  bee.i  continuously revised and adapted according to the tech­
nical  evolution in order to improve their sensitivity. 
. Triti um generating  ins tallations · 
ISJPR.4.-1 , D 20 cooled and moderated research reactor of  5 MWth  nominal power, 
in operation from  1959  to 1973.  The  IX O inventory was ss  7' m 2 with  a final  T-
activity  of 2. l,Ci/l.  T' ^monitoring in the reactorand  ..its effluents was pursued 
until  1977. 
ESSOR  research reactor  (critical  in 1967)  moderated and cooled by D 20. in two 
distinct  loops.  Nominal  power  25 and  17 MWth  in the feeding and experimental 
zones,, respectively, of the core. The D 70 inventories in  these  two zones are  8 and 
12 nv 3, respectively. In the latter tritium has reached an activity  level  of 3.2 Ci/1. 
The total tritium content in the heavy water at Ispra is about  7 grams. 
Tritium  monitoring 
Types  of samples 
Air from operation  areas,  from  gaseous  effluents of experimental loops, from off-
gas systems, from outside area of the installations. 
Water from  liquid  effluents and. from  cooling  water of experimental loops, from 
waste streams, from irradiated  fuel  elements  storage  pools, from rain and surface 
water, from the water table and water supply system. 
Process  fluids,  like  D20, He, No. 
Food  chain and environment,  like  vegetables, fruit,  milk,  soil,  etc. 
Biological  samples,  especially urine for radiological surveillance of operators. 
Surface  contamination control in operation  areas. 
Applied  techniques and  procedures 
Detailed  information on this item is given in a series of reports (refs.1-6). Presently 
the preferential measuring technique applied is  that  by liquid  scintillators adapted to 
the type of sample preparation (distillation,  lyophilisation,  centrifugation, electro­
lytic  isotope enrichment). 
The on-line control of operation  areas  and  gaseous  effluents contamination levels are 

I. 39 
performed by  ionisation  chambers with or without compensation of external radia­
tion fields, depending on the specific requirements. The "in-situ" surface contamina­
tion control is performed by window-less gas  flow  proportional counters. Sampling 
for environmental air- contamination control is performed by bubbling the air through 
distilled  water or passing it over H 2 O selective molecular sieves columns. Chemical 
or cryogenic techniques have been abandoned. 
Sensitivir?  of the  measurements 
Ionisation  flow  chambers, direct  measurement:  5-10" 6 + 5-1 0" 5 Ci/m 3 air. 
Liquid  scintillators: 
air sampling:  10" â - 1CT 12 Ci/m 3; by applying electrolytic enrichment a  limit  of 
5-10"14  Ci/m 3 is attained, which allows to evaluate the natural atmospheric tritium 
background. 
. environmental water and materials samples:  ÎCT^-IO" 7 Ci/m 3; with electrolytic. 
enrichment the  limit  is brought to  about  5-1CT9'  Ci/m 3, which again is sufficient for 
the evaluation of the natural background levels. 
. biological  samples (urine): 1CP  /xCi/ml;  this value  represents  less than  0.1%  of 
the "maximum body burden", ie. the  level  leading to the maximum annual  dose 
limit  for occupationally exposed workers. 
Laboratories and equipment at the JRC-Ispra. 
Process  analyses  laboratory 
This  radiochemical laboratory in the  ESSOR  complex is equipped for  treatment  and 
measurement  by liquid  scintillation of samples with a relatively high T content and 
has acquired a considerable practical experience in this  field. 
Health  physics  laboratory 
Measurements at low T levels are performed in this laboratory which disposes in par­
ticular  of a  pre-treatment  section by  lyophilisation,  electrolytic enrichment (in-house 
project and construction), etc., and various  liquid  scintillators. The staff of this labora­
tory has a large experience in handling samples of various  nature,  in performing the 
measurements  and in evaluating the results (see annual reports, ref.  7). 
In-field control 
Portable instrumentation is currently applied by the Health Physics  Division  for in­
field control of tritium in air and surface contamination. This  Division  has under 
its control also the  local  monitoring instrumentation in the installations. 

I. 40 
Radio tossicological laboratory 
This  laboratory,  divided, into  a radiochemical section for samples preparation and 
an analytical section  with  liquid  scintillation  counting equipment, has performed 
more  than 9000  control measurements on  internai  tritium  contamination of workers. 
Biological  laboratory 
Research  activities are going on in this laboratory on the micro-distribution of T in 
living  organisms and plants and on its  biological  effects  (rets.  8 - 12). 
Experienc e in operation and, maintenance  with  tritium  at the JRC 
The operation of the  ISPRA-1  and. the  ESSOR  reactor systems has required  conti­
nuously  interventions for maintenance and replacement, on a routine basis, as  well 
as exceptionally on T contaminated heavy water loops and components, in an 
atmosphere  with  tritium  oxide contamination. The protective  means  for the opera­
tors have been carefully selected. JRC technicians, in collaboration  with,  supply, 
industry-  for suits,, masks and other protective  clothing,  have developed or  modified, 
such  items in order to improve their performance and  adapt  them to the special 
requirements of operation on  tritium  contaminated heavy water loops. A close 
exchange of information exists on  these  items  with  other nuclear  centres  with  simi­
lar problems. 
References 
1. V. Camera, G.  Dominici,  Determinazione di tritio (DTO) nell'aria, con  scintillalori  liquidi, 
EUR-Ispra-5 14 (10.1463). 
2. D. D'Adamo,  M. De Borfoci,  G. Dominici,  Prelievo di vapore d'acqua tritiaia mediante selacci 
molecolari  e misure in scintillatore liquido con coincidenza  multiple,  Giornale  di Fisica  Fumitaria, 
Vol. 14.N.1,  pp.31-37,  1/3, 1970. 
3. G. Donunici,  A. Fenzi,  E. Momiroci,  Concentrazione  di tritio  in alcune  acque  naturali dell'Italia 
Settentrionale, Energia Nucleare,  Voi.  26,  N.ll,  11 - 1979. 
4. R. Cazzaniga, G.  Dominici,  Tritio  in acque  naturali della zona  di Ispra,  EUR-7349it,  198 1. 
5. G. Dominici,  Misure  di tritio  in campioni acquosi ambientali, Bollettino  delTAIPF,  N.50/51, 
Anno  IX,  Nov.-Dec,  1982. 
6. G. Dominici,  La misura  dell tritio in aria  al CCR-Ispra  (in corso di pubblicazione). 
7. Series of annual  reports  on "Misure di  radioattività  ambientale",  1978  - 1982  (EUR-6632Ì, 
72S0Ì,  7958L  3572Ì). 
8. Margaret  Meriini  and  R Van  Steenwinkel,  The evolution  of biological  studies  on tritium toxicity, 
1979,  Eur-6369EN,  JRC-Ispra, preprint. 
9. Libero  Cerici,  Margaret  Meriini  and  M.J.  Carroll,  The effects  of tritiated compounds  on the  eariy 
mouse embryo  in culture. Campagnari, F., Van Steenwinkel R.  and  Meriini  Margaret,  Molecular 
aspects  of tritium contamination  of DNA.  Tritium-hydrogen exchanges  at the  C-8 position  of 
purines.  Both  talks were  presented  at the  Meeting  of the  Committee  for Internal Emitters  of the 
European Late Effects Project,  St. Bartholomew's  Hospital,  London, September  1980. 
10. L.  Clerici,  M J. Carroll,  Margaret  Meriini,  Luisa  Vercellini  and  F. Campagnari, The toxicity  of 
tritium:  the  effects  of tritiated amino acids  on preimplanted mouse embryos. Int  J.Rad.Biol.  (in press). 
11. Margaret  Meriini  and Libero  Clerici,  Toxicologicai  aspects  of tritium.  Talk  to be  presented  at the 
2nd Int.Symp.on  Environ-Pollution  and its  Impact  on Life in the Mediterranean Region. Iraklion, 
Crete, Greece, September 6-9,  1983. 
12. Paper  in preparation: The  uptake  and  incorporation  of tritiated precursor macromolecuies  in the 
ceils of the  mouse embryos. L.  Cerici  and  Margaret  Meriini. 

I. 41 
3.4.2  Research activities in fusion technology related to tritium problems 
For the past  several years a series of activities is going on in several research groups 
and laboratories of the JRC-Ispra on tritium problems related to fusion technology. 
These activities concern basic parameter studies such as hydrogen-metal interaction, 
as well  as technological aspects, and  impiy  paper as.well as experimental work. 
The latter is obviously performed up to now  with  the lighter isotopes H and D 
for simulation of and extrapolation to tritium behaviour pending verification  with 
the radioactive isotope. 
^_tuxiie£_on_  hydrogeii  Jsotapes-metal interaction 
Since  about ten years, the surface thermodynamics laboratory is active in fusion tech­
nology  studying problems of plasma-surface interactions and physico-chemical pro­
perties of fusion reactor materials (e.g. first  wall).  Measurements have been performed 
on: 
. desorption spectra, adsorption  states,  surface coverages and sticking probabilities 
of the systems  Mo—N r and  Mo-EU  (refs. 1-3); 
. initial  thermal outgassing after different methods of surface preparation, thermal 
desorption spectra, equilibrium desorption  rates,  equilibrium surface coverages 
of different materials: Inconels, stainless steels, aluminium (refs.  4-10);· 
. diffusivity  of H and D in Inconel type-alloys (refs. 11,12). 
On-going  and planned activities are: 
. solubiity and  diffusivity  measurements  of hydrogen  (H,D),  in the first  wall  and 
structural candidate materials; 
. determination of outgassing  rates  of implanted H,D as a function of time; 
. thermodynamical properties and vapour  pressure  measurements  of Lil7Pb83  (ref. 13); 
. solubility  and  diffusivity  of H,D in  Li17PbS3. 
In order to obtain  diffusion  data for the evaluation of the tritium losses through 
the piping of the  cooling  systems in a  liquid  blanket composed of  Lii 7Pbs 3, an ex­
perimental  apparatus  has been set up. Several deuterium permeation  flux  measurements 
through stainless steel membranes have been performed at different  temperatures 
using  an ultra high vacuum installation. Some  measurements  have been performed on 
oxidized  membranes. 
References 
1. F. Reiter,  J. Camposüvan,  Desorption spectra and adsorption  states  of the system  Mo-N 2, 
EUR- 5404  (1975). 
2. F. Reiter,  J. Camposüvan,  Surface coverages and sticking probabilities of the system  Mo-N 2, 
EUR-5762  (1977). 
3. F. Reiter,  Desorption studies from Mo-surfaces, 9th  SOFT,  Garrnish-Partenkirchen  (FRG), 
14 -18  June  1976. 
4. F. Reiter,  J. Camposüvan,  Thermal desorption of Inconel,  EUR-5792  (1977). 
5. F. Reiter,  Thermal outgassing behaviour of Inconel after different methods of surface preparation 
Vacuum,  23 (1978), 559. 

I. 42 
6. F.  Reiter,  J. Camposilvan,  Thermal outgassing properties of stainless steel 316 L after different 
methods of surface preparation, 10th  SOFT,  Padova  (Italy), 4-8 September 1978. 
7. F.  Reiter,  J. Camposilvan,  Thermal outgassing properties of smooth and rough Inconel 600 
surfaces, 8th Int.  Vacuum  Congress, Cannes (France), 22-29 September 1980. 
8. J.  Camposilvan,  F. Reiter,  Thermal desorption spectra,  equilibrium  desorption  rates  and 
equilibrium  surface coverages of  mechanically  polished and of sand-and bead-blasted Inconel 600 
surfaces,  EUR-7114  (1980). 
9. F.  Reiter,  J. Camposilvan,  Initial  thermal outgassing properties of  aluminium  from  20°C  up to 
320°  C, 9th Symp.  on Engineering  Problems of  Fusion  Research,  Chicago  (USA),  26-29 October 
1981. 
10. F.  Reiter,  J. Camposilvan,  Thermal outgassing properties of  mechanically  polished and of 
sand- and bead-blasted Inconel 600 surfaces up to  500°C,  Vacuum,  32, (1982), 227. 
11. M.  Caodin,  J. Camposilvan,  F. Reiter,  Solubility  and  diffusivity  measurements  of hydrogen  in 
fusion  reactor candidate materials,  VUI  Congresso Italiano  sulla  Scienza  e Tecnologia  dei Vuoto, 
L'Aquila  (Italy), 11-15  Aprii  1983. 
12. G. Gervasmi,  J. Camposilvan,  F. Reiter,  Solubility  and  diffusivity  measurements of  protium  and 
deutenum in  Inconel 625, to be presented at the 9th int.  Vacuum  Congress,  Madrid  (Spain), 
16-30  September 1983. 
13. F. Reiter, R.  Rota,  J. Camposilvan,  Thermodynamic  properties of Li^Pb^-, 12th  SOFT,  Julich 
(FRO),  13-17 September 1982. 
14. O.  Gautscir.,  G. Hodapp,  EUR  report in press. The permeation of deuterium through austenitic 
stainless steel. 
Tritium  processing 
• Conceptual design studies 
Feasibility  studies have been carried, out starting in 1978 on a series of conceptual 
process schemes for the exhaust plasma processing, taking the Los  Alamos  TSTA 
process as a reference. The aim of  these  studies was to  identify  simpler and more 
compact processes  which  possibly might  allow  to increase the  overall  safety stan­
dard  of such a plant concept, besides savings in investment and operation costs. 
Two  scheme have been retained (refs.2,3), as indicated in  Figs.  5 and 6.. The second 
scheme offers a decisive advantage  with  respect to the first one for its lower T  inven­
tory by periorming the  final  separation of H  from  D and T by gas chromatography 
rather than by water  distillation. 
Both  schemes  rely  on electrolysis as key process steps. In the first  stage  this procedure 
is used for isotopic pre-enrichment (of H), in the second one for reconversion of the 
oxides  to elementary DT. 
The success of  these  key  steps  depends on  three  factors: 
. achieve a separation factor of at least 10 (in the  cell  1); 
. develop an electrolyzer  with  an as low as possible  liquid  inventor/ maintaining 
nevertheless a (relatively)  high  throughput; 
. render the design and material of  these  cells  compatible  with  high  T-oxide  con­
centrations. 
Answers  to these  questions are required for a  definitive  safety  assessment  of these 
schemes. 

I. 43 
STREAJ4  in «nutUd 
<S>w,wt <^ dintiUt» <S> oui put 
H.O < 2." H, 15 
0,0 120 t.4-10- 0, 120 
T,0 120 <10-' T, 120 Fig. 5 
STHEAU  «n moiud 
moût Ç> oui OUI 
M. 1 M, 15 
ojo 120 0, 1?0 
T,0 120 T, 120 Fig. 6 

I. 44 
Coolant  detntiarion 
Tritium  losses  from  the blanket to the He-primary  cooling  loop has been analysed 
(ref. 4) for a  FINTOR-D  type fusion reactor. Under the realistic assumtion  that  in 
such  a loop also some water  would  be present  due to in-leakage  from  the (secondary) 
steam generation loop HTO (tritiated water)  would  build  up with  time to non accep­
table levels by isotonic exchange of T  with  H. This  exchange offers on the other hand 
the opportunity for using an oxide-hydroxide  equilibrium  reaction for dehydration 
and hence detritiation of the He-stream. Studies of the- kinetics of the water absorption 
reaction:  Si O +  H, O ^  Sr. (OH)2  performed at the  temperature  of the  cold  side of 
the He-ioop-,  have  shr'vn  (ref. 5)  that  the tritiated water content in the loop  could  be 
kept at acceptable levels by this reaction without any losses in thermodynamic  efficiency. 
Electrolytic  cells development 
Experimental  work, was started for the  verification  of cathode potentials in relation 
to H-D isotopic separation factors of various cathode materials (refs.6,7). Ni or Fe 
electrodes  finally  coated, by NiS resulted, the. most  promising,  materials.. 
In parallel for the reconversion  step  of DTO to D-T, a new  cell  concept was designed, 
constructed and successfully tested. In order to eliminate the  liquid  content in the 
ceil compartments (= high T inventory),  only  the separator is soaked by the electro­
lyte,  sprayed into, the anode compartment in the  form  of an aerosol. The separator-
material  utilised  in these  experiments was built  from  asbestos  fibres, the electrodes 
from  Ni.  High  current densities have been achieved, so  that  a relatively  small  unit 
could  already satisfy the  daily  throughput requirements of an  INTOR-type  fusion 
device. 
Expending  the isotope enrichment  tests  to this new  cell  type, a substantial improve­
ment of the separation factors (above 10 for H/D) has been obtained, encouraging 
further research on  still  more efficient materials for suitable cathode and separator 
materials (wettability, radiation resistance besides enrichment) and on optimisation 
of operative conditions. 
References 
1. G.  Pierini,  H. Dworschak,  B. Spelta,  C. Rizzeilo,  S. Sansolini,  A. Tata,  Feasibility  on exhaust 
plasma  processes,  Proc.  Tritium  Technology  in Fission,  Fusion  and  Isotopic  Applications, 
Dayton,  Ohio,  April  29 -  May  1, 1980,  ANS  (1980)  119-124. 
2. G.  Pierini,  Reprocessing  of spent  plasma,  UK Patent  2.040.399B,  1982  published,  1981  application. 
3. Process for  removing  helium  and other  impurities  from  a mixture  containing  deuterium and 
tritium  especially  from  a fusion  reactor,  European  Patent  0032327,  1983  published,  1982 
appl ication. 
4. F.  Lanza,  P. Rocco,  F. Van Rutten,  A system for die  control  of tritium  losses in  primary  and 
steam  circuit  of a fusion  power  plant,  Proc.  of the  Fusion  Technology  1978,  Padova,  1978. 
5. F.  Van Rutten,  Helium  purification  by a  tritiated  water  absorption  system,  Proc.  of Fusion 
Technology  1980,  Culham,  1980. 
6. G.  Pierini  et al.. Experimental  investigation  on key equipments related to an alternative exhaust 
plasma  process, 9th  Symp.  on Engineering  Problems  of Fusion  Research,  Chicago,  HI., October 
26-29,  1981,  Vol.  H, pp.2080-33. 
7. G.  Pierini,  B. Spelta,  .Alternative  processing  modes to  cryogenic  hydrogen  isoiopes  separation in 
exhaust  fuel treatment,  Contribution  to a  special  issue of  Nuclear  Engineering  and  Design,  1983. 

I. 45 
Tritium  recovery from bjanket 
Theoretical  and experimental studies to investigate the  best  method for tritium recovery 
from  the  liquid  breeder material  Li17Pb83  are in progress at JRC-Ispra since some years. 
Measurements of hydrogen isotopes  solubility  and absorption kinetics on  Li17Pb83  at 
various  temperatures  (refs.  1,2,3)  have been carried out. Based on  these  data different 
models for tritium extraction from the blanket material have been considered. 
In the first two models, direct pumping was assumed on either a  stagnant  melt or 
an ideally agitated one. For an  LNTOR  type blanket, in the first  case  a steady  state 
Τ inventory of about  19 kg was calculated, whereas for the second case, less than 
160 g were obtained.  Taking  into account the impossibility of agitating the melt 
"in-situ",  direct pumping  could  be applied practically only on a side stream of the 
alloy  outside the blanket area, as indicated schematically in  Fig..  7, scheme A, the 
final Τ inventory depending in this  case  essentially on the  flow  rate  of the melt in 
the side stream. 
A third model is based, on countercurrent inert gas purging through the melt in a 
parked or a bubble column. In  Fig.  7", scheme B, a schematic representation is  .. 
shown. 
Provided  that  the Τ inventory shall be kept as low as.  200  g in a total blanket vo­
lume of  76 m 3, about  50 m 3 must be  treated  daily.  To achieve this, a  1.7 m high 
column  of 0.15  m diameter and. ~  520  m 3 He is required (ret.  4). 
For an engineering layout of such a  facility  it is however necessary to dispose of 
reliable  data on hydrodynamic characteristics of the  alloy  such as viscosity and 
surface tension and on the hydrogen desorption  rate.  For- the latter determination 
an experimental assembly has been- designed and is to be realized at JRC, in  which 
either direct pumping or countercurrent gas stripping operation conditions can be 
applied  (see chapter  6). 
References 
1. G. Pierini,  A.M.  Polcaro, P.F.  Ricci,  A. Viola,  Tritium  recovery from  liquid  Li 17Pb83  alloy,  blanket 
material.  Nuclear Engineering and Der.gn, in press,  1983. 
2. G. Pierini,  A.M.  Polcaro, P.F.  Ricci, A. Viola,  Tritium  recovery from  liquid  Li 17Pba3,  paper pre­
sented at the 12th  SOFTJulleh,  September  13-17, 1982  (in press). 
3. A_M.  Polcaro,  P.F. Ricci, A. Viola,  Kinetics  of hydrogen absorption on  liquid  Pb 83Li17  alloy, 
to be published. 
4. G. Pierini,  A.M.  Polcaro, P.F.  Ricci, A. Viola,  Perspectives for tritium recovery from  liquid 
Pb 83Li17  alloy,  EUR  report, in press,  1983. 

I. 46 
SCHEME  A 
Comp. A B C. 
T2 5G g/d 50 g/d' TRACES 
He 66.67  g/d 66.67  g/d 
SCHEME  B 
Comp. A Β C D E 
τ2 136 g 86 g 50 g/d / 50 g/d 
He / / 518 mVd 518 mVd / 
5.10 5Kg/d 5.10s Kg/d / / / 
Fig. 7 - Tritium  recovery schemes  from  blanket  (LinPrjB3  as a breeding material). 

I. 47 
3.5 ' Objectives of the JRC  Tritium  Laboratory 
3.5.1  Motivation  and orientation 
The problems posed by the utilisation and by the presence of tritium .in fusion de­
vices  came to the forefront only when the design of large T-burning devices was 
started. It is not so much a question of demonstrating their technical feasibility  like 
for the plasma  ignition  and control, but  rather  of verifying  urgently their practical 
applicability,  in particular for the specific requirements of the operational and  envi-' 
ronmental safety aspects. As seen in the  case  of a fusion power station, the tritium 
produced inside the reactor must be continuously recovered in situ; furthermore, 
the tritium content in the breeder has to be very low in order to  ensure  a low tritium 
inventory  for safety and economical reasons. These facts  will  possibly ask for 
tritium  recovery methods not so far considered. 
Another  example is represented by the permeation and trapping of tritium in the 
first wail  structures. As shown in chapter 3, the fact  that  tritium  will  be in. the. form 
of high energy particles (100 - 1000 eV) and in the presence of simultaneous neutron 
irradiation,  will  possibly strongly modify the implantation,  diffusion  and trapping 
mechanisms across the metallic structures. 
A third point is related to the: type of operations foreseen in the fusion reactor. 
As mentioned, one has to expect to handle large pieces of tritiated materials, in a 
very  difficult  environment and  with  stringent safety requirements, in particular 
concerning  the maximum allowable concentration of tritium. 
A final  aspect  is that  related to the conditioning of the used reactor  parts  (to be 
recycled  or disposed of as waste)  which  will  be at the  same  time tritiated and 
activated and  which  will  be heavy and complex in shape.  Again  the new problems 
to be solved  will  be numerous. 
In front of such a multi-facet situation, it is thought  that  a tritium laboratory at 
the JRC-Ispra, committed primarily to the safety problems of tritium technology 
in fusion, fits very  well  in the framework of the on-going and planned activities  within 
the member countries of the Community. There are common objectives to be 
achieved  in this  field,  to which  the laboratory  will  be devoted, such as: . 
. acquire experience on handling tritium in complex systems in connection  with 
the safety of operators during routine operation (general and  individual  protective 
means) as  well  as with  the protection of the environment  (waste  treatment, 
emergency clean-up, monitoring); 
. provide the conditions to experience procedures for handling tritiated  parts  of the 
reactor and to optimize process components in relation to their operational safety 
and tritium inventory. In particular: 
a) execute experiments on tritium loaded mock-ups reproducing renewable  parts 
the NET reactor in adverse operating conditions and by remote handling, 
aiming  to assess  decontamination, conditioning (boots, seals, weldings), assem­
bling,  and quality control procedures. 

I. 48 
b) test  new concepts and components developed in the European laboratories, 
such  as gas chromatography, electrolyzers and. catalyzers, tritium recovery 
systems  from  blanket. 
c) check the  validity  of physico-chemical data extrapolated to the tritium  case 
from  H/D behaviour. The  reliability  of such data is of primary importance 
for the proper design of containment and ciean-up systems. 
To copewith  these  requirements, a  flexible  multipurpose installation is required 
rather  than a  facility  devoted to a specific problem or system such as for example 
TSTA  (Los  Alamos)  'the time factor plays,, of course, an important role in such a 
decision  too. 
When  TSTA  was decided in the second  half  of the seventies, TNS (the next  step) 
and EPR (experimental power reactor) were scheduled, respectively, for 1984 and 
1990  with  0.25 and 4.6 kg tritium inventories, respectively. The time- span apparently 
available  before a. reliable tritium handling, system was needed for effectively opera­
tional  purposes, was very short. The  only possibility  therefore was to rely on  suffi­
ciently  weil  known techniques, such as water or cryogenic hydrogen  distillation, 
and to experience on a representative  pilot  plant scale one selected streamlined 
process scheme. 
In an experimental reactor such as  NET/INTOR,  the expected tritium inventory  will 
be of the order of several kilograms. However, it  appears  that  the major  part  of this 
inventory  will  be present  as hold-up in intermediate  storage  or at a stationary  state 
in the blanket. In  TSTA  the expected tritium inventory is 150 g, a significant  part 
of which  is taken by the  cryodistillation  column.  Given  the objectives of the JRC-
tritium  laboratory, tritium inventories even lower, of the order of 50 - 100 g,  would 
be sufficient for a significant operation. 
As far as the utilisation of the laboratory is concerned, it has a  twofold  goal: 
. to host research groups  from  laboratories in the European Community not dispo­
sing of facilities  for tritium handling for hot experimental verifications; 
. to perform tritium related experiments in the frame of the JRC  Fusion  Techno­
logy Programme. 
In the  following  an outline is given of  these  activities proposed by the JRC. 
3.5.2 Experimental  program proposed related to the  JRC-fusion  activity 
Tritium  permeation, inventory and outgassing in metallic structures 
An experimental investigation  will be carried out on the outgassing of deuterium 
and tritium into the plasma chamber during dwell-time, on the permeation and 
trapping of tritium through the first  wall  structures. 
The measurement program  will  deal  with: 

I. 49 
deterrniiiation of outgassing  rates  of implanted hydrogen isotopes from a  target 
of stainless steel, Inconel or  another  first  wall  or divertor candidate material 
as a function of time 
- at different  target  temperatures  (200 -  600°C), 
- at different particle energies (0.1 - 5 keV), 
- at different panicle beam densities (up to 1015 cm^.s"1); 
. determination of the recombination  constant  at the inner surface of the  target; 
determination of diffusivities and permeabilities of the  target  material; 
'. investigations Ion the mfluencs oflh-situ formed diffusion barriers on the per­
meability  of the  target; 
. determination of the hydrogen inventory as a function of the number of reaction 
cycles; 
. comparison of the experimental results on the behaviour of implanted hydrogen 
during  dwell-time with the results of existing, models  (TRIT,  PERI,  DIFFUSE), 
and possible new theoretical developments. 
As said, the activity  will  begin with measurements-using protium and deuterium in-a 
cold laboratory from 1984 - 1986 and  will  continue with tritium. The experimental 
installation  is schematically  represented  in Fig. 8. The experimental conditions  will 
simulate the  INTOR  burn and dwell-time cycles (100, 200 and 20 s, respectively). 
Fig. 8  - Schematic  view  of an  installation  for investigation  of the  behaviour  of implanted  hydrogen 
(H,D,T) in the  first  wall  of a  thermonuclear  fusion  reactor. 

I. 50 
Tritiated  components  trea tment 
In any situation where the  structurai  materials (metals) are in contact  with  tritium, 
this is adsorbed and  - especially at higher  temperatures  - dissolved in the metal, pro­
portionally  to the  square  root of its partial pressure. Decontamination can be 
achieved  by outgassing at a  pressure  and  temperature  dependent  tune  rate.  Appro­
priate heating and tritium trapping systems must be developed suitable aiso for 
remote operation  with  complex and bulky structures and components arising from 
maintenance and repair of torus subassemblies. Experimentation of such techniques 
on tritium loaded  mock-vps  prepared in close collaboration  with  the NET design 
team at JRC, are another item of research at the JRC. The mock-ups can be built 
in a reduced scale, but the dimensions must be such as to  allow  the integration of 
critical  items in them (e.g. pipe connections, welding joints) in  true  dimensions. 
The mock-ups should be built in the materials foreseen for the reactor  with  a pos­
sible  reduction in thickness and weight of the shielding layers.  All the remote main­
tenance  operations should be simulated, taking into account: 
. the radiation levels due to neutron-induced activation, 
. the contamination' due to tritium and erosion products, 
. the possible pyroforic effects of erosion dusts, (fire hazard). 
In particular the  following  topics must be  tested: 
. the outgassing properties of neutron-damaged materials, 
. the tritium recovery methods. 
Particular  attention  will  be paid  here  also to the further  treatment  and conditioning 
of these  tritiaied waste arisings and their characterisation. This problem, although 
dealt  with  already in the  fission  reactor scenario, requires reconsideration  because 
of the anticipated drastic increase in quantity of such- waste streams. One example 
of structures to be reproduced for maintenance simulation is shown in  Figs.  9-n. 
Testing  on processes 
Electrolysis 
In the alternative schemes for tritium recovery and  recycling,  assessed by the JRC 
(see chapter  5.2.2),  electrolysis of tritium oxide represented a key process  step. 
The recovery of tritium from highly tritiated waste waters, however, must be con­
sidered anyway an indispensable necessity in a fusion reactor system, whatever 
might be the process concept applied for the main stream and blanket  treatment. 
The activities proposed are the testing of: 
. suitable  separators  withstanding the adverse operating conditions due to the very 
high beta-radiation levels. A series of small cells  (separator  dimension of about 
40 x 40 mm) is planned, equipped  with  a catalytic  recombiner  from  which  the 
oxide  is refluxed into the cells (Fig.  12). 
. leak-tight cell to be assembled by  remote  handling techniques. No hands-on opera­
tions should be applied for maintenance of such a unit considering the high  toxicity 
level  of the tritium oxide. 
. procedures  for detritiation of  separator  and  recombiner  materials before removal. 

I. 51 
Column chromatography 
This  technique offers a very high  degree  of intrinsic safety as compared to the other 
isotope separation processes. Contrarily to cryogenic or water  distillation,  it works 
with  "zero" steady  state  hold  up of tritium, and has therefore no stationary inventory 
providing  a "clean"  facility  in shut-down conditions. Two concepts are applicable. 
The first one relying on the classical adsorption gas chromatographic technique, has 
been  tested  at the EPF Garching (ref. I) up to  batches  of 2 litres per cycle  with 
alumina  (plus 1% SiO^)  filing.  The isotopic mixture is separated into 6 streams, Le. 
FL>, HD, D-,, HT, DT, T2 in about 2 hours at  liquid  nitrogen temperature. To be 
applicable  its throughput must be increased by either shorter cycle periods or higher 
batch volumes, or parallel units, or fr/entuaily a combination of  these  possibilities. 
The parameters  to be studied in this respect are column diameter, working temperature, 
gas flow  rate,  and  finally  other support materials having a shorter retention time. 
Such  materials are described in the literature (ref. 2), but their compatibility  with 
tritium  has to be experimentally proven. A schematic, representation of a  test  loop is 
shown  in Fig.  13. 
The second system  uses  chemisorption on palladium supported on alumina grains. 
The three  isotope species are extracted in their pure form in the  sequence  T2, 
Di, R2- This technique has been proposed by CEA for the-JET  Tritium  Clean­
up System. The procedure is described in ref. 3. The volumes  treated  per cycle 
are larger than for the first system, but the time period required is much longer, 
so that  also  here  the daily  throughput  rates  must  still  be considerably increased 
for full  industrial scale application. Such scale-up  appears  feasible theoretically 
but has to be demonstrated experimentally. A  simplified  test  loop is represented 
in Fig.  1.4. 
Tritium  recovery  from blanket 
On the basis of the results from the on-going  cold  tests,  tritium extraction from 
liquid  metal blanket material  will  be performed on the most appropriate model. 
A similar  test  loop to the one applied for the  cold  experiments  will  be used (Fig. 15). 
It is composed of: 
a gas feeding and cleaning system; 
. a "saturator"  in which  the desired concentrations of tritium (or other hydrogen 
isotopes) in the  alloy  are prepared by bubbling a He stream containing tritium 
through the  alloy; 
. a "reactor", to  which  the pre-loaded  alloy  is fed as a uniform and continuous thread. 
The desorption kinetics of tritium can be measured  with  He flowing  in countercurrent 
through the reactor or keeping it  stagnant. 
The decision between the dynamic or the static operation mode is primarily related 
to the most convenient analytical technique available. In the  case  of testing  with 
tritium,  the static technique  appears  most suitable. In this way the He recycling sys­
tem to the reactor  (part  within  dashed line of Fig. 15) can be eliminated,  allowing  a 
considerable  simplification  of the  test  loop as  well  as a reduction of its overall 
tritium  inventory. 

I .  52
Operational safety
Experimental activities such as those described in the foregoing sections, imply the 
availability of adequate and reliable systems and procedures, for health protection 
and safety, which cannot therefore represent a research item ”a priori” . It is 
intended, however, to evaluate and to qualify by their day-by-day performance in 
practical conditions, systems and procedures such as:
. monitoring systems for area and surface contamination,
. protective clothing and equipment,
. dosimetry and operators surveillance,
. environmental dispersion and control,
. secondary containment ciean-up and gaseous waste treatment systems,
. procedures for intervention and maintenance, ·
. gove-box techniques,
. tritium compatible materials and equipment.
Tne results shall help to formulate recommendations for future facilities and in par­
ticular for the appropriate design of fusion reactor -systems.
References
1. H. Weichselgartner, H. Frischmuth, J. Perchermeier, A. Stimmelmayr, Garching EURATOM  
Association, Ootimisation of a large scale gas chromatograph for separating tritium and
DT from other H isotopes, presented at Knoxville Conf., 1983.
2. M. Bonivento, U. Croatto, R. Ganzeriaand G. Micheion, Isotopic separation of mix­
tures by gas-chromatography, Annuali di Chimica, 68, 1978, p.63-70.
3. Le Tritium, BIST No. 178, Février 1973, Pag.13-14.

Fig. 9 - Semi-permanent parts of the toms and internal renewable segments constituting one of the 12 sectors.

DIVERTOR CASSETTE
M A T E R IA L S : A1SI 316. W , C u W E IG H T  =  30 Ions
Fig. 10- INTOR-hjET plvase 2A.

Fig. 11  - INTOR-NET pfiase 2A
Kenuivubic blanke! scgmenls  
(lief. siJuliou willi  poIoiJal di verlor).
MATERIALS: AISI 316, BREEDER, SHIELD (Fe)
WEIGHTS:
SEGMENT A s 150 tona  
SEGMENT B  £ 100 Ions

I. 56
Fig. 12- Scheme o f a device for ( 3  radiation  
damage testing of electrolytic ceil  
separators (several cells electrically  
set in parallel).
Fig. 13 - Semi-scale chromatographic unitLEGENDA
V,. u VALVES, ELECTRO PNEUMATIC 
MP MEMBRANE PUMPS
F. M. MASS FLOW METER
M  MANOMETER
S.V. SAMPLE VOLUME
He I.D. He IONIZATION DETECTOR
G, . . SAES GETTERS
I.P.G. IMPURITY PURIFICATION GET 
B.V. BUFFER VOLUME

X . 57
Fig. 14- Palladium coated alumina  
chromatographic unit
Fig. 15 - Schematic design of the tritium recovery apparatus from liquid PbS3Li

I .  58
3.6 Description of the laboratory
3.6.1  Building
The building will be a reinforced, concrete structure, partially divided into two floors.
A layout of the laboratory is shown in Fig. 16. It includes:
A 'nail for process development (HPD),  which occupies approximately 3000 m3. It 
is equipped for housing glove-boxes with inert atmosphere to allow the installation 
and handling of equipment containing large amounts of tritium. The HPD will be 
air-tight, by means cn interlocked doors and be kept slightly below atmospheric 
pressure by an appropriate ’’hot ventilation” system. In case of abnormal tritium 
release from the secondary containment to the atmosphere of the HPD, an Emer­
gency Gean-up (ECU) system, automatically actuated, will be able to reduce the 
tritium concentration in the operational area and the tritium release to the environ­
ment to permissible levels..
The HPD will include an expanded area,  consisting of a 500 mJ steel box fully tight 
and separately ventilated. This will provide a containment for large components and 
will allow experimental tests on their detritiation and handling.
A set o î four laboratories  (about 450 m3  ) for experimental research on items requi­
ring sophisticated apparatus»:, but limited tritium inventories. These labs are there­
fore not connected to the ECU.
A separate laboratory  will be equipped for tritiated wastes treatment and condi­
tioning.  It is positioned under the area containing the different gaseous waste treat­
ment systems and is connected to the ECU.
The hot ventilation and tritium removal  apparatus from gaseous effluents will cover 
a volume of about 1400 m3  on the upper floor.
In order to comply with the existing regulations for nuclear facilities, the following 
systems will also be provided:
. dressing room, health physics room, health physics store and monitoring station, 
which cover about 250 m3  of the hot area;
. tritium storage and a counting area; this will cover about 250 m3  ;
. waste storage area; this will cover about 150 m3 .
3 .6 .2  Tritium containment systems
The concept of ’’multiple barrier containment system” will be adopted in the planning 
of the facility. The purpose of this concept is to limit the dispersion of tritium within 
each barrier system so that it can be recovered before it penetrates into the next con­
tainment area. This concept starts from a high-integrity primary containment system 
that is normally the processing equipment, connecting pipes, storage containers, etc.

I . 59
Large pieces of this equipment are enclosed in glove-boxes which form a secondary 
containment system and contain an inert gas atmosphere that is monitored for 
tritium.
The entire laboratory is enclosed within a hermetically sealed building that forms 
the tertiary containment, which can be hermetically sealed in case of accident.
Tritium monitoring and removal from each enclosure is assured by the following gas 
processing systems:
. gaseous waste treatment (GWT) for the effluents from the primary containment;
. inert atmosphere purification system (IAP) for the glove-boxes atmosphere and 
other secondary containment apparatus·;.
. emergency clean-up (ECU) for the HPD, the expanded area and the tritiated waste 
treatment laboratory.
Gaseous wastes treatment (GWT)
The reference flow sheet of this system is shown in Fig. 17. All gaseous effluents 
generated in the Tritium Laboratory are processed through the GWT system before 
being discharged into the atmosphere. The gaseous effluents will enter the GWT 
system through baffled traps and charcoal filters to collect organic vapours primarily 
due to vacuum pump oil. These steps are necessary to prevent possible poisoning 
of the catalyst bed. At the exit of the charcoal trap the gas is dumped into a large 
ballast tank kept at low pressure. This receiver tank is equipped with pressure sensing 
switches When the pressure in the tank reaches the high, sex value, a compressor is 
activated and transfers the gas from the tank to the treatment system. When the 
pressure in the tank is reduced to the low set value, the compressor stops and the 
gas flow ceases. The gas stream enters an electrical preheater and a catalytic reactor 
where the hydrogen isotopes and hydrocarbons are catalytically converted into 
water vapour and carbon dioxide by reacting with the oxygen in the gas stream at 
a temperature of about 500° C. The catalyst consists of a fraction percentage by 
weight of precious metal dispersed on a high surface-area of alumina carrier. On 
leaving the catalytic reactor, the gas stream enters a gas cooler where, by means of 
refrigerated water, its temperature is lowered to 25°C. The tritiated vapours are 
removed from the gaseous stream by a set of three fixed-bed driers one being in 
operation, another in regeneration and the third one in stand-by. The fixed-bed driers 
are filled with molecular sieves These driers operate in two steps with an intermediate 
moisture addition in order to obtain a further reduction of the amount of the tri­
tium oxide by isotopic dilution. The purified effluents either go to the exhaust stack 
or, if the tritium content is still excessively high, are recycled into the low-pressure 
receiver tank.
The driers are regenerated by heating and purging with a reverse flow gas stream. The 
purge gas is then cooled, separated from the condensate moisture and then recycled 
to the low pressure receiver tank.
inert atmosphere purification (IAP)
The reference flow sheet of this system is shown in Fig. 18. This system is aimed at 
keeping the inert gas atmosphere in the glove-boxes free from impurities. The inert

I .  60
atmosphere purification consists of a set of molecular sieve driers for water removal 
followed by copper oxide beds for oxygen removal. The gas is moved through the 
IAP by two blowers operating in parallel. These blowers are not hermetically sealed 
but are mounted in a sealed box provided with water cooling to prevent excessive 
temperature rises of the blowers.. A heated getter bed is included as part of the 
LA P system. This bed can getter tritium and nitrogen. Once the bed is saturated by 
nitrogen, it must be replaced. Nitrogen can be reduced to low ppm levels when 
necessary in this way; because of the limited capacity of the getter bed, however, 
this unit is not left on line continuously and is placed on stream only when it is 
necessary to reduce mtrogen or tritium levels in the glove-boxes. The drying towers 
are regenerated wnen either oxygen or water levels begin to increase. The regeneration 
is achieved by heating the towers and then flowing in countercurrent hydrogen, to 
regenerate the bed for oxygen removal and inert gas to regenerate the molecular sieve 
driers. On the exit of the towers the regeneration gas stream is cooled and then sent 
to the· GWT system to recover all. the tritiated water thus formed.
Emergency clean-up system  i  ECU)
The reference flow sheet of this system is shown in'Fig. 19. The ECU is a large capa­
city air detritiation system able to lower to permissible levels the tritium concentra­
tion in the operational area and the tritium release to the environment. The ECU is 
automatically actuated in case the room monitoring system detects an abnormal 
tritium concentration in the controlled area. At the same time the hot ventilation 
has to be stopped, thus isolating the atmosphere of the controlled area. A schematic 
representation of the gas processing systems both in normal and abnormal situations, .  
is shown in Figs. 20,21. Tritium is removed from the gaseous stream by a com­
bined effect of catalytic oxidation and adsorption on molecular sieve beds. The room 
is kept at a lower pressure than the other rooms, thus limiting the back-streaming 
of contamination to the latter. The effluents from the ECU are recycled to the 
controlled area, thus improving the air detritiation efficiency. The ECU is pan of 
the tertiary containment and has to be designed in function of the other characteris­
tics of the tertiary containment system. These are mainly:
. the size and the tightness of the tertiary containment enclosure;
. the sensitivity and the response time of the tritium room monitor'' in the controlled 
area. The second parameter is strongly depending on their location;
. the time required to stop the hot ventilation and to run in steady state with the ECU 
. the possible adsorption of tritium on the different surfaces in the laboratory. This 
may cause delay effects increasing the operation time required for the ECU.
3 .6 .3  Siting of the laboratory
In the preliminary design study of the laboratory as described in the foregoing 
points, the possible utilisation of already existing buildings or structures (ex­
cept the stack) has not been considered in order to allow an objective identifi­
cation of all requirements free from constraints deriving from such struc­
tures. However, it is known that, with the shut-down of the ESSOR-reactor, a 
number of facilities will become available which could, in principle, be exploited 
to host the tritium laboratory.

I. 6 1
In the following we will evaluate, in the light of the design requirements, which 
are the capabilities inside the ESSOR complex for doing this.
The utilisation of the ESSOR dome as tertiary containment implies some draw-backs 
that might easily off-set the advantage of having the building available with air- 
conditioning, ventilation and stack. The huge volume of the dome (~ 38,000 nr5) 
and the nature of construction materials would require the realisation of another 
containment shell inside the dome in order to keep the emergency clean-up within 
a reasonable size both for investment costs and operational efficiency.
The utilisation of the ESSOR bunkers on the contrary would appear practical only 
for the installation of an already fully developed and well defined pilot installation 
to be operated remotely (such as e.g. cryogenic or water distillation), due to the dif­
ficulties of access to the bunkers and their particular shape (unproportionally high 
with respect to their surface area).
A third potential area within the ESSOR complex is situated near the stack and the 
reactor dome. Two possibilities exist, one of which would allow the incorporation of 
the already existing building for the Super-Sara power supply system.
The second possibility is to-rationalize the layout by constructing a new building 
near the stack, as shown in Fig. 22.
3 .6 .4  Licensing procedures
According to. the Italian regulations for nuclear installations, the licensing procedure 
for a laboratory handling bulk tritium amounts beyond 100 Ci is normally es­
tablished by Article 55' of the LAW D.P.R. 13.2.1964, n. 185. A particular situation 
emerges for such a laboratory if built at the JRC-Ispra, in the area covered by the 
operation license of the ESSOR reactor, utilising facilities of the latter such as stack, 
gaseous effluents monitoring system, power supply (including emergency supply), 
and others.
In a meeting with representatives from the Direction ’’Energy Resources and Basic 
Industry” of the Ministry for Industry and Trade and from ENEA-DISP, respective­
ly , the legal and technical authorities for nuclear licences in Italy, the former 
considered a licensing procedure by Art. 55 for a T laboratory practially not com­
patible with the above mentioned particular situation, due to foreseeable interferences 
with the ESSOR license in effluents surveillance and safety control procedures both 
for normal operation and accidental situations.
A tritium laboratory should therefore be considered as a part of the ESSOR com­
plex and as such be approved by the procedure foreseen for ’’Modifications of 
existing nuclear installations” (starting from art. 42 of D.P.R. 185). In this case 
the authorisation for the construction could possibly be obtained within less than 
one year.
Another advantage would be that the release limits for ESSOR remain valid also 
for the T laboratory. For routine operation the amount of tritium oxide that can 
be released through the 80 m high stack is 2000 Ci/a.

I .  62
The dose commitment in case of the release of 10a Ci* of gaseous tritium from a 
100 m high, stack would be only 10 /a rem. For che very hypothetical release of the 
same amount in the form of initiated water, a factor of 10s must be applied, leading 
to 1  rem of committed dose. The hazard potential represented by this inventory 
of tritium even in its oxide form remains within the existing reference vaiue for the 
External Emergency Plan of the Centre, avoiding so any necessity for modification. 
This plan refers to a maximum credible accident of the ESSOR reactor resulting 
in an expected dose of 4 rem to adults living in the nearest village.
* ■ approximately 100 g of tritium.

PLAN AT 0.00 LEVEL  
(SCALE 1:200}
|) P H A S IN G
2) LAVAlOniES
3 ) H O T  A R E A
4| All) EOCK
PLAN AT +  4.00 LEVEL  
(SCALP 1:200)
P ig- ! < »

pig. J7 - Gaseoiis γ/aste treatment.

Fig. 18 - fner( gfmosplien: puri fica (ion.

Fig 19- Emergency clean-up system.

FLOW DIAGRAM IN NORMAL OPERATION MODE (TRITIUM RELEASE TO H.P.D.)
STACK
Fig. 20

FLOW DIAGRAM IN ABNORMAL OPERATION MODE (TRITIUM RELEASE TO H.P.D.)
STACK
Fig. 21

Fig. 22- Celierai lavo»! '>f (lie liSSOU conijiîex.

I. 70
3 .6 .5  Planning and resources
Starting from a positive Council decision the following steps can be envisaged for 
the realisation of the tritium laborator)':
A) preparation of the preliminary safety report to be transmitted with the request 
for ’’Modification of the ESSOR license” to the Ministry for Industry and Trade;
B) approval of the request by the safety authorities together with identification of 
safety relevant items (’’progetti particolareggiati”);
C) executive design of the laboratory (excluding experimental equipment);
D) construction and mounting of structures and components;
E) functional testing and commissioning.
It is intended to subcontract phases C) to E) to an external Architect-Engineer firm 
which will be responsible for the complete laboratory, excluded the scientific equipment. 
The JRC staffing required for the supervision of such a contract and for the imple­
mentation of phases A) and B) will be limited in the initial period to two prefessionals 
and an adequate number of technicians. With the completion of the construction of the 
facility and its cold commissioning, other technicians and operators will be added, up 
to a maximum of about 14 persons, including mechanics, electronics and ventilation 
specialists.
In parallel the design, preparation and testing of research specific equipment will be 
performed (point F) of Fig. 23). This task will be assigned to the groups already in­
volved at JRC in the measurements with H/D to evaluate processes for Tritium 
Handling (fusion Safety Programme).
On this basis the total operational and scientific staffing of the JRC related research 
activities (excluded the technical and administrative support) is evaluated to a 
maximum of 25 persons, 7 of whom will be professionals. This staff will be needed 
at the moment of the beginning of full operation. In the first 4 years, 10 persons 
will have to be allocated to this activity.
These figures do not take account, of course, of the scientific and technical groups 
from outside that might be interested to perform specific experiments in this laboratory. 
In Fig. 23 a break-down of the time schedule related to the various phases of prepa­
ration and commissioning is given together with that of the overall manpower 
accounted for the laboratory (in order to take account of the JRC’s technical and 
administrative overheads a factor of 2.2 has been applied).

I. 7 1
3 .6 .6  Cost evaluation
A preliminary design and cost evaluation are underway. The results of these studies 
will be supplied at the FTSC meeting of 16 September 1983.
Fig. 23- TIME AND MEN POWER RESOURCES FOR TRITIUM LABORATORY
1 2Years
3 4 5
A
B
C
D
E
FPrelim. Safety report A <  
P.S.R. approval
Executive design - contractor select.
- project elab.
Construct +  Mount - construct, selectis.
- construct +  mount
Funct. testing
Experim. equip. - projects
- preparation
- cold, tests
- install., hot testsB --------------
C  
—
F«-----------C  «------------
D 
—
D « -------------
E<
Men power ■  - A to E  
- F8
620
620
1022
1730
25
Total14 26 30 39 53

I. 72 
3.7 Conclusions 
The analysis performed shows  that  the problems raised by the presence of tritium 
in the fuei cycle of fusion reactors are complex. The data and information already 
available  or expected from classified work do not seem sufficient to solve  these 
problems. A vigorous experimental programme is needed to  meet  the requirements 
of NET and  those  of large scale power reactors. 
On this basis it has been shown  that  a laboratory for tritium handling at JRC  could 
give a significant contribution to the European effort in this area. The main ob­
jective  of such a laboratory  would  be that  of providing a basis to  test  components 
and operation procedures in support of the NET design and construction. This ob­
jective  could  be achieved in the  following  way: 
. by making experimental verifications of concepts related to  fuel  cycle and reactor 
operations, according to proposals  which  will  be made by the European Association 
and by the NET team; 
. by pursuing the tritium-oriented activity started at JRC in the framework of its 
fusion  technology programme. 
This  will  imply  a close collaboration  with  the European fusion research  centres  and 
the NET team starting  with  the design phase,  similar  to  fhat  being done in the 
other  areas  of the Fusion Technology activity, where JRC is  involved. 
In doing this, particular attention  will  be paid to acquiring experience in the opera­
tional  safety  aspects  related to tritium, a task  which  is in harmony  with  the general 
vocation  of the JRC, as it has been recently approved by the  Council  of Ministers. 

I. 73 
APPENDIX  1 - First round projects on tritium set up by the Fusion Technology 
Steering Committee 
The projects are presented in the form of  sheets  following  the  "call  for proposals" 
launched by the Commission in December 1982, namely: 
. fuel  clean-up system, 
. tritium recovery from waste streams, 
. tritium detector, 
. electrolytic  cell, 
. decontamination system, 
. industrial development of large components, 
. tritium recovery from blanket  (liquid  and  solid  breeder). 
Each  sheet  contains: 
. the description of the requested activity (doc  FTSC  418) 
. a summary of the actions proposed by the Associations (doc.  FTSC  5/3.3) 
. the  state  of the art of the related contracts  with  the Association (doc.  FTSC  5/6 and  6/9·)· 
Tritium  1 
FueJ_  clean-up system_ 
Paper study of alternative solutions for the  fuel  clean-up system. The input speci­
fication  is a gas  flow  rate  of 1.5 kg per day of a  fifty-fifty  mixture of deuterium-
tritium  with  an impurity content of less than \% except helium, corresponding to 
the INTOR  requirements. The main impurities are 0 2, N2, H70  CH4,  but the ex­
pected  level  of these  impurities is not known. An effort should be made to de­
termine  these  levels, and while precise data is  lacking,  model  cases  with  high frac­
tions of each of the species should be separately studied. The total impurities 
content in the output gas from the  fuel  clean-up system must be less than 1 ppm. 
Proposal  s_ from  Dj__F_ 
D : The proposal contains: 
a) experimental work in course, using metal  getters  for impurity removal; 
b) a paper study,  followed  by design and construction of a cryogenic dis­
tillation  test  unit (1982 - 1986). 
F : Proposed study (corresponding exactly to the  call  for proposals) (1983 - 1984); 
. accepted contracts/ceilings/time schedule: F - 0.4  MECU  - 1983/84 
D - 1.1  MECU  - 1982/84 
. German technical description has to be modified according to the reduced  ceiling, 
avoiding  duplication  with  French contract. 

I. 74 
Tritium  2 
Tritium  recover/ from waste  streams 
Study  of the recovery of tritium from impurities collected in the  fuel  clean-up 
system. This  will involve tritium both in the  gaseous  and chemically bound con­
dition.  This tritium must be recovered in order to  limit  the amount of effluents. 
The following  different processes may be considered: 
. decomposition of water and hydrocarbons on metallic beds,  followed  by the 
permeation of hydrogen isotopes through a palladium  alloy  membrane; 
. oxidation of all impurities on a catalytic bed  followed  by the trapping and de 
composition  of the tritiated water; 
. separation of the impurities by  means  of a chromatographic column. These im 
purities may be either burnt and collected as water, or decomposed and sepa­
rated on a second chromatographic column. 
All these  methods should be studied  firstly  on paper. However, problems such a; 
those  arising from the poisoning of the palladium membrane by water or hydro­
carbons should be indicated to enable the first  step  of experimentation to  start 
as soon as possible. The study of  these  processes should include an evaluation of 
the amount of tritiated waste produced. 
Proposals from B, D, F 
B proposes: 
a) exp. R&D of laser separation of T from water (1983 - 1985) 
b) a study  followed  by construction of a loop for T-recovery from the  fuel 
. clean-up waste (1983 - 1986). 
D proposes: 
a) exp.  R&D  of the recovery of T from tritiated water and  methane 
(1984 - 1986) 
b) exp. study of the poisoning of  Pd-Ag  membranes (to  start  in 1985). 
F. proposes: 
a) a paper study examining the  cases  for 
i) metal  beds 
ii) oxidation 
iii) chromatography (1983 - 1984) 
b) exp. study of the poisoning of  Pd-Ag  membranes, using existing loop 
(1983 - 1986) 
c) i) exp. study of some components  critically  determining the T-content 
in the waste  streams  (electrolytic  cell,  package in cryodistillation 
columns) 
h) an evaluation of the influence of the  T-level  on the extraction process. 

I. 75 
. accepted contracts/ceilings/time schedule: F - 1.2  MECU  - 1983 - 1986 
. technical description should be modified such as to exclude experimental studies 
other than poisoning palladium-silver membranes 
. the Belgian-laser proposal was referred to a later  stage. 
Tritium  3 
T ri tjum_ dj^ejrtor 
Development of a tritium detector capable of detecting low levels of tritium against 
a radiation background. The tritium detection system should be able to  measure 
under  these  conditions a  level  of tritium around 1 MPC  with  a response time less 
than a few minutes. 
Proposal  s_ from  JJ^ U  K_ 
B : would  develop prop, counter for air or id. (or  GM)  with  individual  gas 
flow;  prototype by 1986. • •·•- -
D : would  develop the sensor according to required  T-level  and environment 
for the monitor. Start 1984, prototype 1986. 
UK: would  develop method to distinguish between T-gas and water; and sensor 
to operate  continuously in a nuclear  facility  (1983 - 1984). 
. accepted contracts/ceilings/time schedule. 
It appeared  that  a monitor close to  that  specified in this task has been developed 
in Canada. Where B and D  still maintain their proposals, the Commission  will 
analyse the situation and report to the  FTSC. 
Tritium  4 
Ekcjrojiy_tic_cell_ 
Study  of an electrolytic  cell processing up to 100 ml per day  with  special conside­
ration  of maintenance and waste, for the reduction of small quantities of tritiated 
water. 
Proposal^  from  f  J^K,_JJRC_ 
B : proposes to scale up ongoing  e-ccll  development for 100 - 1000  Ci, to the 
levels  required in task T-4, including  building  of a prototype by 19S6. 
F : proposes to investigate  critical  problems such as the diaphragm material 
and remote maintenance; and to  build  (1985) and  test  an active prototype 
at levels 106 Ci/1 (1986 or later). 

I. 76 
UK : proposes stepwise development: 
i) 10 - 100 Ci 
ii) 100 - 1000 Ci 
iii) ending up in recommendations for the design of an e-cell system for 
1 MCi/day  (1986). 
JRC : is studying  low-liquid  inventory e-cells since 1981, aiming at hydrogen 
isotopic  separation.  Only  inactive expts. Project  ends  1984. 
. accepted contracts,fceilings/time schedule 
. no conclusions have been reached yet between B and F, but discussions continue 
and die project  will  be settled in the next future. 
Tritium  5 
Decontamination  system 
Paper study of the economic feasibility of processing.large volumes of air contami­
nated  with  tritium, including estimations of cost, release to the environment and 
maintenance problems. The decontamination unit should enable personnel to  enter 
the reactor  hall  without special protection 24 hours after shut-down.  Normal  or 
accidental  leaks to the  hall  atmosphere may be in the  range  1 to 1000  parts  per 
million  of the inventory per day. 
rhe atmosphere of glove-boxes and of secondary containments  will  need to be 
routinely  processed. The normal process consists of  oxidising  tritium and trapping 
:he water. Catalytic  beds  operating at very low tritium concentrations should be 
evaluated and, if necessary, research directed towards increasing the efficiency of 
:he best  catalysts. 
Proposals by B, D, F 
B : offers 
a) studying the economic feasibility of processing large volumes of air; 
b) exp. R&D to decontaminate and possibly recover T from inert gas 
flush  systems in glove-boxes (1983 - 1986). 
D : proposes to develop organometallic  getters  in which  T is irreversibly 
fixed  so that  final  disposal is easy (1983 - 1986). 
F : proposes 
a) a paper study including identification of T-source terms, concentration 
levels,  layout and economic optimisation of clean-up system (1983 -
1984); 
b) testing of commercial catalysts in a wide  range  of parameters  and 
developing  the most promising  ones  towards an optimal solution 
(1983 - 1986). 

I. 77 
. accepted contracts/ceilings/time schedule: D - 1.7  MECU  - 1983/86 
F - 1.6  MECU  - 1983/86 
. further discussions on the experimental  part  of the proposals were necessary 
. technical description of German proposal should be more detailed. 
Tritium  6 
Industrial  development of large components 
Industrial  development of  specific  large components,  fully  compatible for use  with 
tritium,  including  turbomolecular pumps  with  pumping speed between 500 and 
10,000  litres per second and valves of large dimensions  with  a diameter between 
1.5 and 2 metres. The different  steps  should'be as  follows: 
. Ask  from  laboratories  with  expertise in  tritium  handling proposals to make 
available  a unit in  which  such components  could  undergo  final  testing  with  tri­
tium.  This  facility  should be available at the time required by the development 
of the programme. 
. Make  enquiries in industry  with  the aim of assessing the  present  level  of develop­
ment of  specific  tritium-compatible components. 
. Start a study  with  industry of  tritium  compatible components not available to 
the specifications required for  fusion  needs. 
Propc^als_frojm 
D : proposes 
a) industrial survey 
b) feasibility  study of components (pumps and valves) 
c) planning of a  test  facility  (1985) (this  could  be followed  by fabrication 
of components and construction of  test  facility  by 1986/S7, as a pos­
sible  second step). 
F : proposes 
a) industrial survey 
b) study and development of components 
c) adaptation and equipment of existing  test  facility  (which  should be 
rented by the  fusion  programme) (1983 - 1986). 
. accepted contracts/ceilings/time schedule: D and F  Q'ointly)  - 2.85  MECU  - 1983/86 
. the industrial survey and  feasibility  study  will  be carried out  jointly  between D and F 
. the existing French test-facility  will  be accessible to all the Associations and the 
question  of rental is subject to further discussions. 

I. 78 
Pro posal  s_ by_  D^F,_JRC 
D : a) H  solubility  measurements  and other physico-chemical studies (1982/86) 
b) lab-scale testing of the gas bubbling method (unspecified). 
F : a) feasibility  test  of the method (1983 - 1986) 
b) H  solubility  and  diffusion  coefficient  measurements  (1982 - 1985) 
c) demonstration unit (1986). 
JRC : a)  solubility  and  diffusion  measurements  (1983 - 1984, University of 
Cagliari) 
b) experimental verification of the method (1983 - 1985). 
. accepted contracts/ceilings/time schedule: D - 0.4  MECU  - 1983/84 
F - 0.3  MECU  - 1983/84 
Blanket 9 
Proposals^  by_ B,_D,_F,  _I 
B : a) evaluate reaction kinetics  with  Li and  Li-Pb;  regeneration of  solid 
getters  (Y, ...) (1983 - 1984) 
b) demonstration of method in experimental loop (1984 - 1986). Blanket 8 
Tritium  extraction from  LinVbco:  inert "as bubbling 
Demonstration type laboratory  tests  should be performed to determine the potential 
of the method. First experiments should be carried out in static  Li17Pb83  with 
hydrogen. The eutectic used should be  well  defined and the impurity uptake during 
the tests  must be measured. In a second  stage  the validity  of the method should 
be confirmed in experiments based on tritium.  Additional  experiments  will  be 
required to determine the dependence of solubilities of hydrogen and its isotopes 
in Li17Pb83,  on temperature  and partial  pressure  (Sievert's law). 
Tritium  extraction based on the use of  solid  getters 
Thermodynamically  it is possible to extract tritium from  lithium  or Li17Pb83  with 
the use of  solid  getter  materials such as yttrium. However, the reaction kinetics, 
the getter  regeneration and the poisoning  rate  by impurities are  aspects  which 
will determine the potential of this method.  Also  the possible reaction of the 
getter  material  with  the liquid  breeder or a component of it such as Pb should 
be considered. In a first  stage,  experiments should be performed in static devices 
based on hydrogen. In a second  stage,  confirmation experiments based on tritium 
should  be included. 

I. 79 
D : a) batch experiments on  solubility  and reactions  with  Li-Pb;  poisoning 
of getters  (Y, ...) (1983 - 1984) 
b) loop experiments up to 10 mg  tritium  (1985 - 1986) 
c) idem up to 1 g  tritium  (1986 -
F : experimental study of reaction kinetics and extraction  from  various  getters 
(1984 - 1985). 
I : a) capsule (batch) e:.perimeiits on reactions  with  Li and relevant para­
meters; poisonin : of getier (Y,  Y-based  alloys) (1983 - 1984) 
b) loop extraction (! 985). 
. accepted contracts/ceilin's/timi schedule 
. following  discussions bet veen B and D, the repartition of founds  will  now be: 
B : 0.26  MECU 
D: 1.44  MECU 
Blanket  16 
Tritium  recover/ 
An experiment  will  be designed or proposed for an existing irradiation  facility  in 
which  the recovery of  tritium  from  the  solid  breeders  will  be directly demonstrated 
as a function of irradiation  temperature  and burn-up (or dose), using in appro­
priate method of T recovery: the samples  will  also be used to determine property 
changes  for comparison wi'h Task B-15  within  the limitations posed by the study 
of T extraction. 
Pro£Osals  by B.JD.J^J_ 
B : proposes on-line montoring and trapping of T released  from  vented capsules 
(see B-15). 
D : proposes a loop expe iment in  DIDO,  monitoring of T by an ionisation 
chamber in the carrie gas, after reduction of T 20 in a hot metal bed. 
F : proposes irradiation i i  SILOE  and systematic study of extraction methods 
covering  T2 and  tritu:  ted species. 
I : proposes to irradiate  LiA102  in a He-gas loop in  TRIGA  RC-1 and  trap  the 
tritium  in the gas by cryogenic and  getter  systems. 
. budget allocation not y r decided. 

COMMISSION OF THE EUROPEAN COMMUNITIES 
Joint Research Center 
II. A VIBRATING TABLE 
at 
JRC - ISPRA 
Second progress  report 

II. 2 
TABLE OF CONTENTS 
page 
1. INTRODUCTION  II. 3 
2. REVIEW OF RESEARCH  UNDERWAY  ON SEISMIC ANALYSIS AND 
IDENTIFICATION OF USE AND INTEREST FOR VIBRATING  TABLES  II. 3 
3. REVIEW OF EXISTING VIBRATING  TABLES  II. 4 
4. IDENTIFICATION OF POSSIBLE  COMMUNITY  NEEDS FOR A LARGE TABLE II. 4 
4.1 Conclusions  of the  Italian  ad hoc  Working  Group II. 4 
4.2 The  situation  in France II. 5 
4.3 The  situation  in the U.K. II. 6 
4.4 The  situation  in Greece II. 6 
4.5 The  situation  in the FRG II. 7 
5. EVALUATION  OF THE ADEQUACY  OF THE ISPRA SITE II. 8 
6. CONCLUDING  REMARKS  II. 9 
7. FUTURE DEVELOPMENTS  II. 9 

1. INTRODUCTION 
This  report is the second progress report on  vibrating  tables  pre­
pared  by the JRC ad hoc  Working  Group,  whose  mandate is to evaluate 
the interest for the  Community  to install at  JRC-Ispra  a vibrating 
table of intermediate  size  (100-300  t). 
The report includes: 
1. A review of research underway on seismic analysis and identification 
of use and interest for  vibrating  tables. 
2. A review of existing  vibrating  tables. 
3. Possible  Community  needs  for a large  vibrating  table. 
4. Evaluation  of the adequacy of the Ispra  site. 
5. Concluding  remarks. 
6. Future  developments. 
2i Review of research underway on seismic analysis and identification  
of use and interest for  vibrating  tables. 
Community 
- Qualification  and model validation  tests  are performed using 
medium-size  tables. 
- No large-scale  tests  are performed at present, but some interest 
has been expressed (see section 3). 
- Efforts  are underway for model and computer code development for 
seismic analysis. Amongst others, progress is hampered by a  lack 
of representative data on the reference  ground  motion to be consi­
dered. 

II. 4 
Japan 
- Extremely  vast  research activities covering all  aspects  of seismic 
analysis. 
- Series of large-scale demonstration  tests  (nuclear and petrochemical 
structures) planned on Tadotsu 15 m x 15 m, 1.000 t, shaking table. 
- Pseudo-static (reaction wall)  tests  on apartment buildings up to five 
stories  high: repeated reversed lateral forces increased in  steps 
(Tsukuba). 
U.S.A. 
- Wide spectrum of research activities sponsored mainly by National 
Science Foundation and National Regulatory Commission. 
- Active research on aseismic design and construction of buildings using 
both pseudo-static  tests  and a 45 t shaking table at University of  Cali­
fornia,  Berkeley. 
- Collaboration with Japan for testing of very large models. 
3. Review of existing vibrating tables. 
Four  tables  have been prepared,listing the major vibrating  tables  in 
the world and their main characteristics. They are reproduced in annex 
of the present note. The  biggest  tables  are in Japan with six  tables 
in the range  100-1.000  t. The largest table in the  U.S.A.  is the 45 t 
table at Berkeley (Ca.). The largest table in the Community appears to 
be the 14 ft x 14 ft biaxial table with 100 t capacity installed at 
6EC Power Engineering, Stafford  Mech.  Lab., U.K. 
4. Identification of possible Community  needs  for a large table. 
4.1 Conclusions of the Italian ad hoc  Working  Group 
An ad hoc  Working  Group  was set up by  ENEA  in March  1983 with the ob­
jective of 
a) collecting up-to-date information on dynamic testing of structures 
throughout the world-, 
b) analysing the  needs  in Italy for a vibrating table of large capacity; 
c) indicating the optimal characteristics of such a table. 

II. 5 
The W.G.  report was received in  July  and its main conclusions can be 
summarised  as follows: 
- there is, in principle, an interest for the possibilities that 
could  be offered by a large vibrating table with a capacity in the 
range  100-300  t: 
• Possibility to  test  complex components in the non-linear range. 
• Possibility to refine present  state  of knowledge on dynamic struc­
tural  behaviour and thereby reduce  safety  margins, both in nuclear 
and non-nuclear fields. 
- Perplexity is expressed as far as  cost  is concerned, both for the 
operation  of the table and for the construction of large-scale com­
plex models. 
- Access by  road  to inland  sites  imposes restrictions to the  size  and 
weight of the equipment that could be  tested.  The use of the 1.000 t 
Japanese table could be the ideal solution for very large models. 
- At present, the potential "market" in Italy is not judged sufficient 
to justify an adequate use of a large table. However, the situation 
could  change in  case  of a substantial modification of the present 
norms. 
- Final  advice can be given only when the  needs  from  the other Community 
countries have been fully appraised. 
- The large table should not be  seen  as an isolated tool, but rather 
as part of an adequate experimental context of seismic engineering 
which  clearly involves other experimental equipment than shaking 
tables. Attention should also be given to the problem of ensuring 
the availability of the necessary competences. 
4.2 The situation in France 
Two potentially interested organisations have been identified in 
France: 
a) according to  FRAMATOME  there is a lack of adequate experimental means 
in the Community for the seismic qualification of large equipment. 
The problem arises in particular for  sites  with a high seismic  risk 
in the  sense  that non-linear phenomena have to be taken into account 
and,  therefore, large-scale models should be built and  tested. 
Among  the problems to be considered, mention was made of: 
- the  primary  coolant loop of a PWR and its support; 
- steam generators (behaviour of tube bundles); 
- devices for storage in water pools of spent fuel elements; 
- behaviour of cranes. 

II. 6 
b) The "Centre Scientifique et Technique du  Bâtiment"  (CSTB)  in 
association  with  "Laboratoire  Central  des Ponts et  Chaussées" 
has in project the  construction  of a triaxial 6 m x 8 m  vibrating 
table  with  100 t maximum  equipment weight.  The construction  of 
this table has not yet been decided by the  French  authorities, 
but CSTB  is fully convinced that a large vibrating table is a 
unique  tool for: 
- the research on building structures constructed by assembly of 
prefabricated  components. Singularities are present at the junc­
tions  between  assemblies which are impossible to model analyti­
cally  and experimentation with a large table is the only way to in­
vestigate  the failure modes provoked by seismic loading; 
- the improvement of aseismic design concepts and construction 
techniques; 
- the development of techniques to increase the seismic resistance 
of existing buildings; 
- the development of models to represent actual structural behaviour 
and the validation of calculation methods; 
- the qualification, against a specified seismic excitation, of  full-
scale  equipment and housing structures. 
4.3 The situation in the U.K. 
From  contacts with  UKAEA, CEGB  and Nuclear Installations Inspectorate, 
it appears that the seismic qualification requirements as specified by 
the U.K. code are well met by the existing vibrating tables. The  pro­
blem  of qualification against earthquakes is not considered very rele­
vant in the U.K. 
The National Engineering  Laboratory  at Glasgow which operates three 
tables is open to collaborations. 
Contact  has also been taken by letter with the  Building  Research  Estab­
lishment, but no reply has been received yet. 
4.4 The situation in Greece 
Prof.  P.G.  CARYDIS,  chair for Earthquake Engineering, National  Technical 
University,  Athens, visited  JRC-Ispra  in July  and illustrated the 
Earthquake  Simulation  Facility  presently under construction in his de­
partment. 
According  to Prof.  Carydis,  an effort should be made at the Community 
level to 

II. 7 
- collect  and disseminate information on earthquakes 
- install standardised seismographs to record strong ground motions 
and create a corresponding data bank 
- create a  European  network of shaking  tables  with a centralised in­
formation  service on the particular characteristics 
- complement the above network with a table of large capacity 
- develop models describing dynamic structural behaviour under realistic 
ground  motion, and disseminate the resulting computer programmes. 
4.5 The situation in the FRG 
In the FRG research activities in the field of anti-earthquake design 
have been intensified in the  last  decade, in particular in view of  buil­
ding  nuclear reactors. At present a spin-off is observed in the field 
of civil  engineering. Dynamic testing of models and/or components is 
limited  to the use of smaller shaking  tables  and to the use of shakers 
on existing structures. Since six months, a  25 t shaking table is in 
operation at HRB  Jülich  (see table  1). 
The contacts which have been taken in FRG concern: 
(i) reactor vendors (KWU, HRB) 
(ii) licensing authorities (IRS,  Köln) 
(iii) research  institutes  (TH Darmstadt, HDR project of KfK). 
The vendors of reactor  systems  have a tendency to be in favor of  tests 
of prototypes  scale  up 1:1,  in particular in the field of  fluid  struc­
ture interaction (fuel  elements,  steam generators,  etc.).  The research 
institutes  prefer the smaller shaking  tables  in order to make detailed 
experimental verifications of the fundamental bricks of computer  pro­
grammes. The licensing authorities are inclined to follow the research 
institutes  i.e. to rely on the computer programmes to simulate huge 
structures provided the constitutive  laws  in the non-linear regime are 
validated by means of clean experiments. In general, in FRG like in 
Japan  and  U.S.A.,  the interest in huge shaking  tables  is decreasing 
and preference is given to the pseudo-dynamic method to verify experi­
mentally huge buildings. The main reasons are cost/benefit considera­
tions: a shaking table  costs  around  100 2 /  kg of model. 
The various organisations in  FRG  which have been contacted appreciate 
the initiative of the Commission.  They  will  formulate in the near future 
a more detailed advice on the need to coordinate in  Europe  the research 
in the field of anti-earthquake design and the need on thereby huge 
experimental  test  facilities. 

II. 8 
5. Evaluation of the adequacy of the Ispra  site. 
Study  contracts have been launched in order to  investigate  by static 
and dynamic geotechnical  tests  the adequacy of the preselected  site 
within  the  JRC-Ispra  to host  a large vibrating table. 
The tests  to be performed include: 
- the construction of a vertical strati graphic map of the  site; 
- the geomechanical characterisation of the soil (standard penetra­
tion  tests,  laboratory  tests  on soil specimens, control of under­
ground  water); 
- dynamic consolidation  tests; 
- use of special dynamic  exciters  to measure the attenuation of seismic 
waves  as a function of their frequency, and thereby to evaluate possibl 
effects  of the operation of the table on neighbouring installations 
and buildings.. 
Contacts with the local electricity company  (ENEL)  have ascertained 
that the necessary power supply can be ensured-without any difficulty. 
Concerning  access  to JRC-Ispra,  the present  road  connections have per­
mitted the transportation of components for three research reactors 
(Ispra-I, ECO,  ESSOR).  The construction of a new  road,  connecting the 
JRC to the nearby highway, could  also  be envisaged. Transportation of 
cumbersome and heavy equipment was  also  a concern for the USA  (space 
and nuclear programmes): a solution has been envisaged based on  diri­
gibles.  Similar  studies  have been performed in  Europe. 
Nevertheless, transportation of complete models could be avoided in 
many  cases.  On-site assembly is possible at  JRC-Ispra,  using existing 
facilities. 
As far as the availability of the necessary  competences  is concerned, 
although the JRC is not involved at present in seismic analysis, exper­
tise  exists  in the fields of 
- dynamic testing 
- instrumentation, data acquisition 
- numerical analysis of structures 
which  could be used for the present project. Obviously the recruitment 
of a few  specialists  would be necessary to  build  up a competent team. 

II. 9 
6. Concluding remarks. 
The analysis presented above, while still incomplete, allows to draw 
some  first indications: 
- Except for Greece,  all other  Member  States  consider that the 
present norms for seismic design may be adequately fulfilled 
using avaiblable experimental and analytical  tools. 
- A large number of small and  even  medium  size  shaking  tables  exists, 
some  of them having six controlled  degrees  of freedom which allow 
a more realistic simulation of expected ground motion. 
- A number of organisations in the Community have expressed a real 
interest for a shaking table of larger capacity.  This  indicates 
that not all phenomena until now can be modeled either by small-
scale  tests,  or analytically. 
- A co-operative European effort would be beneficial in the following 
areas 
• Standardisation of ground motion measurements and creation of a 
European  data  base. 
• Development of refined models and computer  codes  verified by ex­
perimental  equipment either available or to be built to enable a 
better understanding of phenomena and possibly reduce  safety  mar­
gins. 
7· Future developments. 
The following  studies  will be  executed  and included in a report to be 
issued by mid  1985: 
- evaluation of the study contracts to  assess  by static and dynamic 
geotechnical  tests  the adequacy of the Ispra  site  to host  a large 
vibrating  table; 
- further evaluation of the Community interest; 
- description of a possible research activity; 
- preliminary  estimate  of staff requirements and  costs. 
The results of  these  studies  should allow to decide whether or not to 
build  a vibrating table at Ispra and to launch a detailed design study 
to be ready by the end of  1986. 

II. 10 
Comments to Tables  1 to 4: 
1. Istituto SperimentalcModelli e  Strutture  S.p.a. 
2. National Engineering Laboratory 
3. 
4. Commissariat  à l'Energie Atomique 
5. Société  pour le Perfectionnement des  Matériels  et Equipements  Aérospatiaux 
6. Hochtemperatur Reaktorbau  GmbH 
7. Construction Engineering  Res.  Lab.  (US  Army) 
8. National Research Centre  for Disaster Prevention,  Science and Technology  Agency 
9. The Nuclear  Power Engineering Test Centre 
10. Institute  of Industrial  Science,  University of Tokyo 
11. Railway Technical Research  Institute,  Japan National Railway 
12. Public Works Research  Institute,  Ministry  of Construction 
13. Central Research  Institute  of Electric Power Industry 
14. Takasago Research  Institute,  Mitsubishi Heavy  Industries 
15. National Research  Institute  of Agricultural  Engineering,  Ministry  of Agriculture 
16. Port and Harbour Research  Institute,  Ministry  of Transport 
17. Shimizu  Construction Co. Ltd. 
18. Kajima  Construction Co. Ltd. 
19. Building  Research  Institute,  Ministry  of Construction 
20. Ohbayashhgumi Construction Co. Ltd. 
21. Disaster Prevention Research  Institute,  Kyoto  University 
Keywords: 
DOF  Degrees Of Freedom 
EHSS  Electro-Hydraulic Servo System 
EVS  Electrodynamic Vibration System 
H Horizontal 
V Vertical 

TA J 31 j 1 -  Major vibrating  tables  in Europe 
Table cnaractori sties ] -erf oriauce 
Facility Table size 
(m) Freer.  Panne J fax rodel 
weig!it  (t) Controlled 
D.O. P. i'ax disp. 
(nun) !;ax vel. "a: 
iv.y.'/soc  (g : accel  . 
TO 
1.1 Aar  also I r,mianti 
Genova-Ttaly 3 · 5  x 3 · 5 0-60 2 H + 70 
V ± 70 CGO 
550 2.4 
1.2 ic:'.::r. 1 
Bcrga.no-Italy 2.5 x 4 0.1-50 2 ± 35 400 
1 .3 il!-2L 2 
Glasgow-UK 3x3 0-33 20 3 ?I± 125 
V± 125 
1 ,/| GIX 3 
•Jhetstond-UK 
StaPford-UK 3.85  x 3.2 
14ft x 14ft 1-33 
3-1000 25 
100 2 
o 
1.5 C.'iA 4 "VESUVS" 
Saclay-Fran.ee 3.1 x 3.1 1st 
resonance 
freq.200Hz 20 1 i 100 1000 1.8 
1 .6  CiJA "TOlJliTIAGOL" 
Sa c1ay- Fr anc e 2 x 2 1st 
resonance 
freo.200Mz. 10 2 H ± 125 
V± 100 2200 
1200 
1.7 soi'ünc/w 
V <: 1 i: 'V-  V i 11 ac on b  1 a y 
France 3 x 3 0-50 10 2 ff I 110 
V± 110 600 
600 3 
LH ilkU 6 CfcibH  "SAMSOM"  . 
Julnc-i-Fi'JG 5 x 5 0.5-100 25 3 1" 200 1000 1.5 
1 .S' national  Tech. 
University 
Athens-Greece 4 x 4 0-60 10 6 ;i ± loo 
V ± 100 cso 
625 1.5 
·."> n 

TABLE  2 - Major  vibrating  tables  in USA 
Facility Table  size 
(m) Freq.  range 
(Hz) Max model 
weight  (t) Controlled 
D.O.I 7. Max displ. 
(mm) Max vel. 
nun/sec Max aceel. 
Is'*) 
2.1 University of Illinois,  Urbana, 
Illinois (Uniaxial, 1968) 3.65 x 3.65 0- 50 4.5 1 H±51 381 5 
2.2 University of California, Berkeley, 
California (Biaxial, 1971) 6.1 x 6.1 0- 50 45 5 II ± 1 52 
V± 51 635 
254 0.67 
0.22 
2.3 CERL7,  Champaign, 
Illinois (Biaxial, 1973) 3.65 x 3.65 0- 200.· 6 5 H±73 
V ± 35 813 
686 15 
30 
2.4 Union Carbide, Oak Ridge, 
Tennessee  (Biaxial, 1980) 1.83 x 1.83 0-20 7 2 H± 193 
V± 193 305 
305 0.25 
0.25 
2.5 SUNY,  Buffalo, N.Y. 
(Biaxial,  1983) 3.65 x 3.65 0.1-60 20 5 II ± 1 50 · 
V± 150 635 
318 1.0 
1.0 
2.6 NASA,  Marshall  Space  Flight 
Center,  AL (Biaxial, 1969) 3x 4.5 — 0.9 6 — 2438 100 
2.7 Cornell University,  Ithaca, 
N.Y.  (Biaxial, 1982) 2.1x2.1  . 0-50 0.27 2 H±76 
V + 76 800 
800 6 
2.8 Westinghouse,  Pittsburg, 
Pa (Biaxial, 1978) 4.9 x 4.9 0- 100 17.8 3 H ± 305 
V± 305 762 
762 3.5 Table  characteristics l'ci Tonmnce 

TABLE  3 - Major  vibrating  tables  in Japan 
Table  characteristics  Performance 
Facility Table  size 
(m). Freq.  range 
(Uz) Max mode! 
weight (t) Con [rolled 
D.O.F. Max displ. 
(mm) Max vel. 
mm/sec Max accel. 
(S's) 
3.1 NRCDP STAB 
Tsukubo-Ibanigi 15x15 EHSS  0-50 500 2 H ± 30 
V ± 30 370 
370 0.55 
1 
3.2 NUPEC9 
Tadotsu-Takamatsu 15 x 15 EHSS  · 1000 2 II ± 200 
V ± 100 750 
375 1.8 
0.9 
3.3 IIS,  UT 10 .Chibeu- 10 x 2 EHSS 170 1 ± 100 0.4 
3.4 RTRI,  JNR11 
Kusutsdii,  Tokyo 10x 2 100 1 ± 30 0.4 
3.5 PWRI, MC12,  Tsusubct/ 6x8 EHSS  0-30 100 1 ± 75 600 0.8 
3.6 CRIEPI13, Abiko,  Chiba, 6.5 x 6 EHSS 125 I ± 50 600 1.2 
3.7 NRCDP STA8 
Tsuku  beu-6x6 EHSS  0-50 75 3 H± 100 
V±50 SOO 
600 1.2 
1 
3.8 TRI MM14 
Wmcji,  Hyogo 6x6 EHSS 50 2 II ± 50 
V± 50 420 
420 1.2 
1.2 
3.9 NRIAEMA 15, Tsukuba 6x3.2 EVS  . 0- 50 30 1 ± 50 320 0.4 

Table 3 -  cont. 
Table characferislics Performance 
Facility Table  size 
(m) Frcq.  raiige Max model 
weight  (t) Controlled 
D.O.F. Max dLspl. 
(mm) Max vol. 
mm/see Max accol. 
(H's) 
3.10IM-IRI,  MT16 
Yokosuka,  Fanaga  me 5.5x2 EVS 17 1 ± 50 150 0.5 
3.11 SHIMIZU'7,  Tokyo 5x4 EHSS 12 1 ± 100 720 1 
3.12  KAJIMA" 
Tokyo 4x4 EHSS 20 2 II ± 150 
V ±75 1140 
455 1.2 
2 
3.I3PHRI,MT,Ä, Yokosiika 4 x 3.5 EVS 30 1 ± 50 250 0.45 
3.14  BRIMC",  Tsukube 4x3 EHSS  0-50 20 1 ±75 600 1 
3.15  0IIBAYASHI, Tokyo20 4x3 EHSS 10 1 ± 100 · 1 
3.16  DPR211, Kyoto 3x3 EHSS 12 1 ± 100 0.5 
3.17  DPRI11 
Kyoto 2.5 x 2.5 EVS 8 2 H ± 50 
V±50 500 
500 0.5 
0.5 

TABLE  4 - Major vibrating  tables  around  (lie world  (besides  EEC,  USA and  Japan) 
Table  characteristics Performance 
Facility Tabic size 
(m) Freq.  range 
(Hz) Max model 
weight (t) Controlled 
D.O.F. Max displ. 
(mm) Max vcl. 
mm/sec Max accel. 
la'*i 
4.1 University of  Kiril  and 
Mctocljj,  Scopie,  Yugoslavia 
(Kiaxial,  1980) 5x5 0-30 40 5 II ± 125 
V± 50 635 
380 0.67 
0.4 
4.2 Arya  Mchz University, 
Teheran,  Iran 
(Uniaxial) 5x5 0-50 50 1 II ± 50 — 0.6 
4.3 University of  Palliavi, 
Siraz,  Iran 
(Uniaxial) 4x4 0-50 20 1 H± 50 — 1.1 
4.4 University of Mexico, 
Mexico  City,  Mexico 
(Uniaxial,  1975) 4.5x2.5 0-50 20 1 H±51 381 1-2 
4.5 Toni'.  Ji University, 
Shanghai, PRC 
• (Biaxial,  1982) 4x4 0-60 10 6 II ± 100 
V± 100 890 
625 1.5 
2.7 

COMMISSION OF THE EUROPEAN COMMUNITIES 
Joint Research Center 
III. PERSPECTIVES  OF HOSTING AT THE  JRC-ISPRA  ESTABLISHMENT 
A FUSION MACHINE  LIKE IGNITOR 
Second progress  report 

III.  2 
TABLE OF CONTENTS 
page 
1. INTRODUCTION 
2. POWER SUPPLY 
3. MAIN BUILDING  AND AUXILIARY  SYSTEMS 
4. LICENSING 
5. COSTS EVALUATION III. 3 
III. 3 
III. 4 
III. 4 
III. 5 

III.  3 
1. Introduction 
In the preceding  progress  report  dated May 1383 (COM  (83)393  final) 
the results  of a preliminary  assessment  of three main items  related  to 
the siting  of IGNITOR  at Ispra were given : electrical  supply  system, 
main building  and auxiliary  systems,  licensing. 
During  this period  these  items  have been  further investigated, 
considering especially  the utilisation  of the  facilities  offered  by 
the ESSCR  installations.  The  consideration  of many  detailed  and 
practical  aspects  has been made possible  on one  side by a  better 
definition  of the  various nuclear  activities  planned  in the  ES30R 
area, in the  frame of the  proposed  pluriannual  programme  of the JRC 
(CO." (83)327  final),  and on the  other side  by the  availability  of 
updated  detailed information  on the  features  and' requirements  of the 
IGNITOR  experiment. 
2. .Power  supply 
We have received  the detailed  answer  of ENZL to the  questionnaire 
specifying  the supply  requirements.  The  feasibility  of the  supply 
directly  from the 380 kV  line at the  border of the  Ispra Centre  has 
been officially  confirmed.  The cost for the  connection  to the  line has 
been given together with  the contractual conditions  for the  supply in 
the period  1907-1991. 
In the reference:  cose of 3 pulson/day,  the  offer am be  quoted  as 
follows : 
. Installation  costs : 820.000.000  Lit 
. Construction  time : 24 months 
..Annual  charge  for electrical  power : 565.000.000 Lit  (supposing  a 
consumption  of 120.000 kv/h/y 
, For energy  dissipated  in the  transformer  a supplementary  charge  of 
75 Lit/kwh  is requested. 

III. 4 
The or. ly condition put on the  power  delivery is that the shots be 
pre-.i;xou. n.C8d. It is further  suggested  to operate during the  working 
time of working  days when the chort  circuit  power  is at maximum  level 
and the pulses  provoke  less disturbance to the  power  grid. 
The site of the transformer  station  within the Ispra  Centre  has been 
identified,  and the layout of the  electrical  line up to the  medium 
voltage substation  near the IGNITOR  main building has  been defined. 
3. Main building and  auxilitiry  nyntcms 
Considering that no other use is  proposed  for the  ESSCR  containment 
building  and on the basis of the  updated  information on the design of 
IGDITOR  and operation and  licensing  requirements, the  installation  of 
IGN'ITOR  in this building has  been reconsidered. In  fact in view of the 
new information obtained, the previous  judgement  about  difficulties  in 
locating  the machine  inside the  containment  was premature. 
The solution envisaged now is its  installation  at working  floor  level 
(Tig. 1) which seems  to offer many practical  advantages.  The shielding 
can be provided by  removable  blocks  handled  by the  existing  50 T 
crane. Many  auxiliaries,  like  liquid  nitrogen  distribution  station, 
exist already in the building.  Other  auxiliary  services, including 
activated  materials interim storage, can be  easily  located in various 
existing  rooms,  with  direct  access to hot  cells  (Fig,  2). The large 
containment  building  connected  with an  80 m. high  ventilation  stack 
ensure ample  protection against any  maximum  hypothetical release of 
tritium  or activated dust. 
The relative  disposition  of IGHITOR  and the other laboratories and 
programmes  within the  ESSOR  installation  is shown on Fig.  3. 
Δ, Licensing 
An explorative contact har.  been made  with the  Italian  licensing 
authorities  in order to assess the possible  approach  to the  licensing 
of various possible new nuclear  activities  in the  ESSOR  complex 
including  IGMITOR. 
The suj'j-ested  approach  is to put all these  activities  under  the 
existing  licence and to deal with  tho.-.i  an "modification of the 
installation".  In this hypothesis, that  most probably  will be followed 
os a central guideline in the  licensing  process, it  will not even be 
necessary for  IGTJITOP.  to follow the  procedure  required by the art. 
DP?. 12.2, 196£ n. 1S5,  mentioned  in the previous progress report. In 
this case the simpler  procedure  of the art. 12 DPR 155  would  be 
followed. 

III. 5 
5. Cor.t3  evaluation 
The evaluation of the  cost.j  for the  site preparation (including the 
pcver supply  system  and buildings), the adaptation of various 
auxiliaries,  the supply of  power  and technical services, considering a 
construction  period of  about  three years  (1SS4-87)  and an operation 
period of four years  (19BB-1991)  is in progress. This evaluation  will 
be available together with the technical report at the end of 
September. 

III. 6 
Fig. 1 : Installation  of IGNITQR 
in ESSOR containment 
building. 
Vertical  section 

III.  7 
Fig. 2 : Installation  of IGNITOR  in 
ESSOR containment  building 
Layout of  auxiliaries 

III.  8 
Fig. 3 : Disposition  of IGNJTQR and 
other laboratories  and 
programmes  within E5SCR 
complex