Jet Joint Undertaking : annual report 1994.
Date: 2017-10-16
Institutions: European Commission
Subjects: Joint European Torus, activity report, nuclear energy, nuclear fusion

J O I NT  EUROPEAN  TORUS 

JET 

JOINT 

UNDERTAKING 

ANNUAL 

REPORT 
1994 

EUR  16475-EN-C 

EUR-JET-AR17 

JET 

JOINT 

UNDERTAKING 

ANNUAL 

REPORT 
1994 

MAY  1995 

LEGAL NOTICE: 

Neither the Commission of the European Communities nor any person acting on behalf of the 

Commission is responsible for the use which might be made of the following information. 

Catalogue  number:  CD-NA-16475-EN-C 

for  the  report  EUR  16475  (EUR-JET-AR17) 

This  document is intended for information  only and should not be used as a  technical 

reference. 

Editorial  work  on this  report  was  carried  out  by 

B.E.  Keen  and  G.W.  O'Hara 

Prepared  and  produced  by JET Publications  Group. 

©Copyright  ECSC/EEC/EURATOM,  Luxembourg  1995 

Enquiries  about  copyright  and  reproduction  should be addressed to: 

The  Publications  Officer,  JET Joint  Undertaking,  Abingdon,  Oxon, 0X14 3EA, U.K. 

Printed  in England 

Preface 

Introduction,  Summary  and  Background 

Introduction 

Report  Summary 

Background 

Objectives  of  JET 

JET,  Euratom  and  other  Fusion  Programmes 

The Joint  European  Torus 

The  Community  Fusion  Programme 

Large  International  Tokamaks 

Technical  Status  of  JET 

Introduction 

Technical  Achievements 

Technical  Developments  for  Future  Operations 

Scientific  Advances  during  1994 

Introduction 

Experimental  Programme 

Main  Scientific  Results 

Progress towards  a Reactor 

Future  Programme 

Introduction 

JET Strategy 

Future  Plans 

Members  and  Organisation 

Members 

Host  Organisation 

Project  Team  Structure 

Administration 

Introduction 

Finance 

Contracts  Service 

Personnel  Service 

Health  Physics and  Safety 

Press and  Public  Relations 

Publications  Group 

Appendices 

I  The  JET  Council 

II  The  JET  Executive  Committee 

III The  JET  Scientific  Council 

1 

1 

2 

3 

7 

10 

16 

21 

23 

59 

69 

70 

72 

99 

103 

106 

107 

113 

116 

116 

123 

123 

127 

128 

132 

135 

136 

139 

140 

141 

j£ 

1994 was a year of significant achievement for 

Preface 

the JET Project. The longest and most complex 

shutdown since the machine was first 

assembled was successfully completed at the 

start of the year. By the end of January, the 

components for the new pumped divertor 

phase had been installed, which entirely 

transformed the interior  of the torus. The 

experimental  programme began, as planned, 

early in  1994 with virtually a new machine. 

In  1991, JET had been extended to the end of  1996 to enable the Project to 

establish reliable methods of  plasma purity control and plasma exhaust in operational 

conditions  relevant for the Next Step Tokamak, and to prepare for the final phase of 

JET with deuterium - tritium  plasmas. Plasma impurities  have always been a major 

obstacle to steady state operation which is desirable in a fusion reactor. The pumped 

divertor  has been installed in the JET machine in order to control power exhaust and to 

screen plasmas from  impurities which cool the plasma and reduce the number of 

energy producing reactions. 

In February  1994, the first  plasmas were produced successfully  in the new 

machine configuration, ushering  in the experimental  programme.  Initial experiments 

were focused on establishing and characterising  plasma behaviour  and on  confirming 

reliable operation  in the  new configuration. The careful design of the divertor  target 

plates and sweeping of  the plasma over the target tiles permitted the  high power 

handling capability  of the divertor to  be demonstrated. As a result,  "blooms"  of 

carbon  impurities which  had previously terminated  high performance,  could  now be 

eliminated.  Furthermore, the Torus cryopump  has proven to  be effective for the 

improvement  of  impurity and density control  of the plasma. 

The scientific  results  obtained  during  the  year  under  review  showed  improving 

plasma  performance.  Energy  confinement  times  of  more than  one second  have 

been  consistently  achieved.  In the  high  mode of  plasma  energy  confinement (H-

mode),  this  containment  has been  maintained  throughout  a pulse  length  of  20 

seconds. The triple  fusion  product  of  plasma  density,  plasma  temperature  and 

confinement  time,  obtained  simultaneously,  has reachec  "near  break-even 

conditions",  which  compares  with  the  best  values  reached  in JET before  the 

machine  was  modified.  Higher  values are expected to  be achieved  during  the 

continuing  campaign  in  1995. 

The JET machine,  as the  most  powerful  fusion  experiment  in the  world, is 

designed to operate with  large plasma currents and high  power  heating systems. In 

the new configuration, the  plasma current  - a major factor  in the performance  of a 

tokamak  - was gradually  increased again to  5 Megamperes. This value greatly exceeds 

the current  employed  in any other fusion experiment  presently operating  in the world. 

During the past year, experiments were also conducted with  total  power from the 

Neutral  Beam Injectors and Radio Frequency  Heating systems of  over 24  Megawatts in 

the plasma. This will  be further  increased during  1995. The experiments showed that 

reliable operations at these high currents, similar to those  required  in a fusion  reactor, 

can be sustained in a divertor  configuration  without  major  problems  in operation. 

JET's present experimental  programme concentrates on the study of  reactor 

relevant  problems and the  provision of further  important  information for the design 

of the International Thermonuclear  Experimental  Reactor (ITER) Project. 

ITER is a collaborative  agreement  between  Euratom  and the  Governments  of 

Japan, the  Russian  Federation  and the  United  States of  America.  Its  basic  objective 

is to  demonstrate  the  scientific  and technological  feasibility  of  fusion  energy  for 

peaceful  purposes.  ITER would  accomplish  this  objective  by  demonstrating 

controlled  ignition  and extended  burn  of  a deuterium  - tritium  plasma,  with 

steady-state  as an  ultimate  goal. 

The present  Engineering  Design Activities  (EDA) of  ITER have  identified  further 

research  and  design  issues which  need to  be addressed, and these can be 

investigated  in JET. A three-year  ITER-EDA  Support  Phase of  JET would  provide 

further  data  of  direct  relevance  to  ITER before  JET entered  into  its final  phase of 

D-T operation. The JET Council  has therefore  decided  to  seek  a further 

prolongation  of  the  Project  until  the  end of  1999. The  prolongation  would  enable 

JET to  make  essential  contributions  to  the  development  and  demonstration  of a 

viable  divertor  concept  for  ITER by carrying  out  experiments  using  D-T  plasmas in 

an  ITER-like  configuration.  This would  broaden  the  basis for  the  D-T  operation  of 

ITER. JET's proposed  experimental  programme  would  allow  a co-ordinated,  timely 

and  cost  effective  investigation  of  the  various  options  for  an  ITER divertor. 

Hopefully,  the  Council  of  Ministers will  approve  the  JET extension  later  this  year. 

Due to temporary  blocking  by the European  Parliament of  part of the  Community 

Fusion Programme  Budget for  1994, the original  budget  could  not be adopted  until 

August. As a consequence, JET had to operatein the meantime within  a constrained 

budget  and rigorous expenditure  restrictions were enforced within  the  Project. 

It was with  great  sadness that  we  learned  of  the  death  of  Jean Teillac  in Paris 

on  10  March  1994.  Professor  Teillac  chained the  Interim  JET Council  established  in 

1977  and was  Chairman  of  the JET Council  from  1978 to  1984. The JET Project  is 

deeply  appreciative  of  the  role  that  he played and the  devotion  to JET that  he 

demonstrated  during  the  crucial  formative  years of  the JET Joint  Undertaking. 

Those of  us who  knew  him will  also fondly  recall  his  humanity,  courtesy  and 

dignity.  His immense  contribution  will  also  be remembered  when  the  history  of 

JET is recorded. 

I should  like to  express  my sincere  gratitude  to  the  members  of  the  JET Council 

for  their  unfailing  support  during  my term  as Chairman, which  ended  in  March 

1995.  Professor  Francis Troyon, who  has succeeded  me, is a long  standing 

member  of  the  JET Council  and was  Chairman  of  the  JET Scientific  Council  from 

1988  to  1992.  I wish  him  every success  in  his new  role. 

I would also like to  record  my belated congratulations to  Dr Robert Aymar, who 

resigned from the JET Council and was appointed  Director of  ITER last autumn, and to 

thank  him for  his valuable contribution  to the  Project. 

The JET Executive Committee and the JET Scientific Council continued to  provide 

excellent support and advice to the JET Council throughout the year, for which we on 

the JET Council are deeply grateful. 

In conclusion, I congratulate  the JET Director,  Dr Martin  Keilhacker and all the staff 

of  the  Project for  their  outstanding  success  in completing  a demanding  and 

complex  programme.  I wish  the  Project  every success. 

(ji^^^O^w^ 

H von  Bulow 

Chairman  of  the JET Council 

March  1995. 

I N T R O D U C T I O N, 

S U M M A RY 

A ND 

B A C K G R O U ND 

Introduction,  Summary  and 
Background 

Introduction 

The Joint  European  Torus  (JET) is the  largest  project  in the 

coordinated  fusion  programme  of  the  European  Atomic  Energy 

Community  (EURATOM), whose  long  term  objective  is the  joint 

creation  of  safe environmentally  sound  prototype  fusion  'eactors. 

The  Statutes  setting  up the JET Project  include  a requirement  for 

an Annual  Report to  be produced  which: 

"...  shall show  the  current  status of  the  Project, in  particular 

with  regard  to  timetables,  cost, performance  of  the  scientific 

programme  and  its position  in  the  Euratom Fusion Programme 

and in  the  world-wide  development  of  fusion  research. ' 

This Report  is designed to  meet this requirement.  It provides an 

overview of the scientific, technical and administrative  status of the JET 

programme, which  is intended to  be comprehensible to the average 

member of  the public. Where appropriate, descriptive sections (in italics 

and boxed) are included to aid the  reader's understanding  of  particular 

technical terms used throughout  the Report. 

A more detailed and comprehensive description  of the technical 

aspects of  the JET Project can be found  in the JET Progress Report. 

Report  Summary 

The  Report  is essentially  divided  into  two  main  parts: 

• 

• 

the  scientific  and technical  programme  of  the  Project; 

the  administration  and organization  of  the  Project. 

The first  part  of  the  Report  includes  a brief  introduction,  provides 

an overview  of  the  planning  of  the  Report  and  sets the  background 

I N T R O D U C T I O N, 

S U M M A RY 

A ND 

B A C K G R O U ND 

Nuclear  Fusion 

Energy  is  released  when  the  nuclei  of 

light  elements  fuse  or  join  together  to 

form heavier ones. The easiest reaction  to 

achieve is between the two heavy isotopes 

of  hydrogen  (deuterium  and  tritium). 

Most  of  the energy  released  in  this 

reaction  is carried away by a high  speed 

neutron.  The remaining  energy goes  to 

the alpha-particle  (helium  nucleus,  *He) 

which  is also produced  in  the  reaction. 

In a  fusion  reactor,  a jacket  or  blanket 

around  the  reactor  region  would  slow 

down 

the  neutrons,  converting 

their 

energy  into  heat.  This  heat  could  be 

extracted 

to 

raise 

steam 

for 

conventional  electricity  generation. 

to  the  Project.  This is followed  by a description  of  JET and  its 

experimental  programme  and  explains  its  position  in the  overall 

Euratom  and  International  Fusion  Programmes.  In addition,  it  relates 

and compares JET to  other  large fusion  devices throughout  the  world 

and confirms  its  pre-eminent  position  in fusion  research. 

The following  section  reports the technical  status of JET including: 

the completion  of technical changes during the  longest shutdown  in 

JET's history to  install the  new divertor  configuration; preparations for 

future shutdowns and progress on systems for future operation. This is 

Fuels 

followed  by a section on scientific achievements during  1994. It sets out 

As  deuterium  is a  common  and  readily 

separated  component  of  water,  there is 

progress towards  reactor  conditions and compares the  performance 

a  virtually  inexhaustible  supply  in  the 

between JET and other  tokamaks.  It shows the substantial  achievements 

oceans of  the world.  Incontrasi 

tritium 

does not occur na tur ally in any significan t 

made by JET since the start of operations  in  1983. The scientific  part of 

quantities  and  must  be  manufactured. 

this  Report concludes with  a description of the proposed future 

programme of JET until  its planned conclusion. 

The second part of the  Report explains the organisation and 

This can be achieved  by using  reactions 

that  occur  between  neutrons  formed  in 

the  fusion  reactions  and  the light  metal 

lithium. 

Therefore,  although 

the 

fusion 

management  of the  Project. It describes the administration  of JET, in 

reactions  occurring  in a reactor  will  be 

which  it details the  budget  situation; contractual  arrangements; and sets 

between  deuterium  and  tritium, 

the 

consumables  will  be  deuterium  and 

out staffing  arrangements  and complement. 

lithium. 

Background 

In the early  1970's, discussions were taking  place within  the  European 

fusion  research  programme on a proposal to  build a large tokamak 

fusion device to extend the plasma parameters closer to those  required 

Fusion Reaction  D +  T ->  "He + n 

Tritium  Breeding 
Reactions 

6Li + n  —>  T + 4He 
'Li + n  —> Γ + 4He +  η 

There are  sufficient  reserves of 

lithium 

available 

to  enable  world  electricity 

generation  using  fusion  reactors,  to  be 

maintained  at present  levels, for several 

in a reactor.  In  1973, an international  design team started work  in the 

hundreds  of years. 

I N T R O D U C T I O N, 

S U M M A RY 

A ND 

B A C K G R O U ND 

UK, and by mid-1975, the team  had completed  its design for  a very 

large tokamak  device. 

On 30th May  1978, the Council of  Ministers of the European 

Conditions  for  Fusion 

Fusion  reactions  can  only  take  place  if 

the  nuclei  are  brought  close  to  one 

Communities decided to build the Joint  European Torus (JET) as a Joint 

another.  However,  all  nuclei  carry  a 

Undertaking of the European Fusion Programme. To implement the 

positive  charge and therefore  repel each 

other.  By heating  the  gaseous  fuels  to 

Project, the Joint Undertaking was originally established for a period of 

very high temperatures, sufficient  energy 

12 years, beginning on  1 st June  1978. The device would  be built on a site 

can  be  given  to  the  nuclei  that 

the 

repulsive  force  can  be  overcome  and 

adjacent to Culham  Laboratory, the nuclear fusion research laboratory of 

they  to fuse together.  In the  deuterium-

the United  Kingdom Atomic  Energy Authority  (UKAEA), and that the 

tritium  reaction,  temperatures  in excess 

of  WO million  degrees Kelvin are required 

UKAEA would act as Host Organisation to the Project. Figure  1 shows the 

- several times hotter  than  the centre  of 

site of the JET Joint Undertaking at Culham in the United Kingdom. 

the sun.  Below  100 million  degrees,  the 

deuterium-tritium 

reaction  rate  falls  off 

The Members of the Joint  Undertaking  are Euratom, its Associated 

very rapidly:  to  one-tenth  at  50  million 

Partners in the framework  of  the Fusion Programme, including  Sweden 

degrees,  and  20,000  times  lower  at  10 

million  degrees. 

and Switzerland, together with  Greece, Ireland and Luxembourg, who 

A reactor  must  obtain  more  energy 

have no Contracts of Association with  Euratom. 

from  the  fusion  reactions  than  is put  in 

to  heat  the  fuels  and  run  the  system. 

Eighty per cent of the expenditure of the Joint  Undertaking  is borne 

Reactor power  output  depends  on  the 

by Euratom.  As the host organisation, UKAEA pays ten  per cent, with 

square  of  the  number  (n)  of  nuclei  per 

unit  volume  (density)  and  the  volume 

the  remainder  shared between  Members  having Contracts of 

of  gas. 

Association with  Euratom  in proportion  to the  Euratom financial 

participation  in the total  costs of the Associations. 

The Project Team is formed  mainly by personnel from the Associated 

Power  losses must  also  be  kept  to 

a minimum  acceptable  level by  holding 

the hot  gases in  thermal  isolation 

from 

their  surroundings.  The  effectiveness 

of this isola tion  can be measured  by  the 

Institutions, although  some staff  are assigned on a secondment  basis 

energy  confinement 

time  (τϋ - the  time 

from the Institutions  and the  Directorate  General of the  Commission 

taken  for  the system  to cool  down  once 

all  external 

forms  of  heating 

are 

resoonsible for  Science Research and Development (DGXII). 

switched  off. 

In July  1988, the Council of  Ministers agreed the  prolongation  of  the 

In  a  fusion  reactor 

the  values  of 

temperature, 

density  and 

energy 

JET Joint  Undertaking  to  31st  December  1992. A further  proposal to 

confinement 

time  must  be  such 

that 

prolong JET to  31st  December  1996 was approved  by the Council of 

their product  (n .rj),  exceeds the  figure 
of5x102'm3skeV.  Typical values  forthe 

Ministers  in December  1991. The extension  is to allow JET to  implement 

parameters 

that  must  be  attained 

the new  Pumped Divertor  Phase of  operation, the objective of which is 

to establish effective  control  of  plasma  impurities  in operating 

conditions close to those of the  Next Step. An extension of the JET 

simultaneously 

for  a reactor  are: 

Central  ion  temperature,  Γ 

W-20keV 

Central  ion  density,  ni 

2.5x102°mJ 

programme to  1999 is currently  being proposed  in support o~ the ITER 

Energy  confinement 

time,  rf 

divertor while satisfying the  requirements  of JET D-T operations. 

Objectives  of  JET 

The original  decision  of  the  Council  of  Ministers  in  1978  states  that 

the  JET Joint  Undertaking's  mandate  is to: 

I-2s 

The  temperature 

is  expressed  as 

the average  energy  of  the nuclei  (IkeV 

is  approximately  equal  to  10  million 

degrees  K). 

N T R O D U C T I O N, S U M M A RY A ND B A C K G R O U ND 

Fig.1 : Aerial v i ew of t he JET Joint U n d e r t a k i n g, situated near O x f o rd in t he United Kingdom 

GAS 

PLASMA 

· *. 

.Ψ 

> ι 

■·©■► 

«L 

Plasma 
of 

temperature 

As 

the 

the 

fuel 

is 

increased,  theatomsin 

thegasbecome 

ionised,  losing  their  electrons,  which 

normally  orbit  around  the  nuclei.  The 

mixture  of  positively  charged  ions  and 

negatively  charged  electrons 

is  very 

different 

from  a  normal  gas  and  is 

given  a special  name  -  PLASMA. 

The fact  that  a plasma  is a  mixture 

of  charged  particles  means  it  can  be 

controlled  and influenced  by  magnetic 

fields.  With  a suitably  shaped  field  it 

should  be  possible 

to  confine 

the 

plasma  with  a  high  enough  density 

and  a  sufficiently 

long 

energy 

confinement 

time  to obtain  net  energy 

gain. 

The configuration 

that  has so  far 

advanced  furthest 

towards  achieving 

reactor  conditions  and  on  which  most 

data 

is  available 

is  the  TOKAMAK, 

originally  developed 

in  the USSR. 

I N T R O D U C T I O N, 

S U M M A RY 

A ND 

B A C K G R O U ND 

Fusion  Reactor 

In a fusion  reactor  a lithium  compound 

would  be incorporated  within  a  blanket 

surrounding 

the  reactor  core  so  that 

some  neutrons  can  be  utilised 

for 

manufacturing 

tritium.  The 

tritium 

produced  would  then  be  extracted  for 

use in  the  reactor. 

The  blanket  would  also  provide 

the means of utilising  the energy  carried 

away 

from 

the 

reactions 

by 

the 

neutrons.  As  the  neutrons  are  slowed 

down 

within 

the 

blanket, 

its 

temperature  would  rise  thus  enabling 

steam  to  be  raised  so  that  electricity 

could  be generated  in the  conventional 

manner. 

Ultimately, 

it  is  hoped 

that 

the 

conditions  would  be reached  to  enable 

a  reactor 

to  be  built  utilising 

the 

deuterium-deuterium 

reactions  below: 

D  +  D  ->  3He  + n 

D  +  D  -H>  T +  p 

In  this  case  there  would  be  no 

need  to  manufacture 

tritium  and  a 

virtually 

inexhaustible 

reserve  of 

energy  would  become  available. 

Schematic  of a Fusion  Reactor 

Superconducting  Magnet 

Plasma 

Blanket 

Shielding 
Structure 

Heat 
Exchanger 

Deuterium  Fuel 

Tritium 

Tritium and 
Helium 

Helium 

Steam  Boiler 

Electric  Power 

Turbine and  Generator 

'... construct, operate and exploit as part of the Euratom  fusion 

programme and for the benefit of its participants in this 

programme, a large torus facility of  tokamak-type and its auxiliary 

facilities in order to extend the parameter range applicable  to 

controlled thermonuclear fusion experiments up to conditions 

close to those needed in a thermonuclear reactor. ' 

The principal objective of JET is to enable the essential  requirements 

of a tokamak  reactor to  be defined. To implement this, it was necessary 

to create and study plasma in near-reactor  conditions. 

There are four  main  areas of  work: 

1.  the  study  of  scaling  of  plasma  behaviour  as parameters 

approach  the  reactor  range; 

2.  the  study  of  plasma-wall  interaction  in these  conditions; 

3.  the  study  of  plasma  heating; 

4.  the  study  of  alpha-particle  production, confinement  and 

consequent  plasma  heating. 

In addition, JET is pioneering  two  key technologies  required  in 

fusion  reactors:  the  use of  tritium  and  remote  handling  techniques. 

%&~?JU  ¡MÄEU-M-

J E T,  E U R A T OM 

A ND  O T H ER 

F U S I ON 

P R O G R A M M ES 

JET, Euratom and  other 
Fusion  Programmes 

The  Joint  European  Torus 
JET uses the  tokamak  magnetic  field  configuration  to  maintain 

isolation  between  the  hot  plasma  and the walls  of  the  surrounding 

vacuum  vessel.  A diagram  of  the JET apparatus  is shown  in  Fig.2  and 

the  principal  design  parameters  are presented  in Table I. 

Fig.2:  Illustration  of  the  JET  Apparatus 

J E T, 

E U R A T OM 

A ND 

O T H ER 

F U S I ON 

P R O G R A M M ES 

PARAMETER 

PLASMA  MINOR  RADIUS: 

HORIZONTAL 

VERTICAL 

PLASMA  MAJOR  RADIUS 

FLAT-TOP  PULSE LENGTH 

WEIGHT  OF THE IRON CORE 

SIZE 

1.25m 

2.10m 

2.96m 

20s 

2800t 

TOROIDAL  FIELD COIL  POWER (PEAK  ON  13s RISE) 

380MW 

TOROIDAL  MAGNETIC  FIELD AT  PLASMA CENTRE 

3.45T 

PLASMA CURRENT: 

CIRCULAR  PLASMA 

D-SHAPE  PLASMA 

VOLT-SECONDS  TO DRIVE PLASMA CURRENT 

ADDITIONAL  HEATING POWER 

Table  I: Original  Design  Parameters  of JET 

3.2MA 

4.8MA 

34Vs 

25MW 

The  toroidal  component  of  the  magnetic  field  on  JET  is  generated 

by  32  large  D-shaped  coils  w i th  copper  windings,  which  are  equally 

spaced  around  the  machine.  The  primary  winding  (inner  poloidal 

field  coils)  of  the  transformer,  used  to  induce  the  plasma  current 

which  generates  the  poloidal  component  of  the  field,  is  situated  at 

the  centre  of  the  machine.  Coupling  between  the  primary  winding 

and  the  toroidal  plasma,  acting  as  the  single  turn  secondary,  is 

provided  by  the  massive  eight  limbed  transformer  core.  Around  the 

outside  of  the  machine,  but  within  the  confines  of  the  transformer 

limbs,  is  the  set  of  six  field  coils  (outer  poloidal  field  coils)  used  for 

Magnetic  Circuit 
(iron  transformer  core) 

Inner  Poloidal  Field  Coils 
(primary  transformer  circuit) 

Toroidal 
Field 
Coils 

O u t er  Poloidal 
Field  Coils 
(for  plasma 
positioning 
and shaping) 

Plasma  w i th  Plasma  Current,  Ip 
(secondary  transformer  circuit) 

Resultant  Helical  Magnetic  Field 
(exaggerated) 

Magnetic  Field  Configuration 
The 
field 
tokamak  magnetic 
configuration  is  built  up  from  three 
components.  The  first  of  these is 
produced  by a set of coils  around  the 
minor circumference. These coils produce 
the toroidal magnetic  field around  the 
major axis of the machine. The second 
component (poloidal field) is produced 
by a large current caused to flow through 
the plasma  by transformer action.  The 
combination of these produces a helical 
magnetic field which  keeps the plasma 
away from  the vessel  walls. The  final 
component is generated by  a  set of 
hoop coils,  which is used to shape and 
stabilise the position of the plasma. 

J E T, 

E U R A T OM 

A ND  O T H ER 

F U S I ON 

P R O G R A M M ES 

postioning,  shaping  and stabilising  the  position  of  the  plasma  inside 

the vessel. 

During  operation  large forces  are  produced  due to  interactions 

between  the  currents  and  magnetic  fields. These forces are 

constrained  by the  mechanical  structure  which  encloses the  central 

components  of  the  machine. 

The  use of  transformer  action  for  producing  the  large  plasma 

current  means that  the JET machine  operates  in a pulsed  mode. 

Pulses can be  produced  at  a maximum  rate  of  about  one  every ten  to 

twenty  minutes, and  each one  can  last for  up to  60  seconds  in 

duration.  The  plasma  is enclosed within  the  doughnut  shaped 

vacuum  vessel which  has a major  radius  of  2.96m  and  a  D-shaped 

cross-section  of  4.2m  by 2.5m.  The amount  of  gas  introduced  into 

the  vessel for  an experimental  pulse amounts  to  less than  one  tenth 

of  a gramme. 

The construction  phase of  the  Project,  from  1978  to  1983, was 

completed  successfully  within  the  scheduled  period  and within  8% 

of  projected  cost  of  184.6  MioECU  at January  1977  values. 

The first  plasma  pulse was  achieved  on  25 June  1983  with a 

plasma  current  of  17000A  lasting  for  about  one  tenth  of  a second. 

This first  phase of  operation  was  carried  out  using only  the  large 

plasma  current  to  heat the  gas.  In  1985, the first  additional  heating 

system,  employing  radio-frequency  heating, came  into  operation  and 

during  1991  reached  22MW  of  power  into  the  plasma. The  neutral 

beam  heating  system was  brought  into  operation  in  1986, and 

exceeded  its design  capability  in  1988, with  21.6MW  of  power 

injected  into  the  torus. 

Experiments  have  been  carried  out  mainly  using  hydrogen  or 

deuterium  plasmas, although  during  1991, experiments  were 

performed  in  helium-3  and  helium-4  and  a preliminary  experiment 

was  performed  using  10% tritium  in  deuterium.  In the  final  stage  of 

the  programme,  it  is planned  to  operate  with  deuterium-tritium 

plasmas  so that  abundant  fusion  reactions  occur.  The  alpha-particles 

liberated  from  the  reactions  should  produce  significant  heating  of 

the  plasma.  During  this  phase, the  machine  structure  will  become 

radioactive  to  the  extent  that  any  repairs and  maintenance  would 

have to  be carried  out  using  remote  handling  systems. 

J E T,  E U R A T OM  A ND  O T H ER  F U S I ON  P R O G R A M M ES 

The  Community  Fusion  Programme 
Objective  and  Near-term  Programme 

The long-term  objective of the  programme, embracing all activities 

undertaken  in Member  States (plus Switzerland)  in the field of 

controlled thermonuclear  fusion  by magnetic  confinement,  is "the joint 

creation of  safe, environmentally  sound prototype  reactors, which 

should result  in the construction  of economically viable power stations, 

which will  meet the needs of  potential  users.  In this context,  particular 

Heating 

Initial  production  and  heating  of  the 

attention will  be given to the constraints  imposed  by the  requirements 

plasma  is produced  by the large  electric 

of  power  utilities"  (Council  Decision 94/268  Euratom of  26 April  1994 

current 

flowing 

in  the  plasma 

itself 

(ohmic  heating)  used  to  generate 

the 

concerning the  Euratom  Framework  Programme  1994 to  1998, OJ No L 

poloidal  magnetic  field. 

115, 6.5.94).  The long timespan and the  large human and financial 

efforts  needed to  attain this objective call for  a concentration  of 

Community  action on the objective, the complete cohesion of  the 

The heating  effect  of  this current  is 

reduced  as  the  plasma  gets  hotter  as 

the  electrical  resistance  of  the  plasma 

decreases with  increasing 

temperature. 

Therefore, 

it  is  necessary  to  provide 

network  of organizations  associated in the Community  action and the 

additional  means  of  heating 

if 

the 

full exploitation  of  cooperation with  the  major fusion  programmes 

outside the  Community. 

Safety  and environmental  issues will  play a central  role  in the 

realisation  of  the  large  devices, which, after  JET, are included  in the 

strategy  leading  towards  a prototype  reactor.  This strategy  includes, 

in  particular: 

-  an experimental  reactor,  Next  Step, the  overall  objective  of  which 

is to  demonstrate  the  scientific  and technological  feasibility  of 

fusion  energy  for  peaceful  purposes.: 

-  a demonstration  reactor,  DEMO, capable  of  producing  significant 

quantities  of  electricity. 

temperatures  needed  for  a reactor  are 

to  be  reached. 

Two  main  additional 

heating 

methods  are  in  general  use: 

(1)  Neutral  Beam  Heating: 

In 

this 

method, 

a  beam  of 

charged 

hydrogen  or  deuterium 

ions  is 

accelerated 

to  high  energies  and 

directed  towards  the  plasma.  As 

charged  particles  cannot  cross  the 

magnetic 

field  confining 

the 

plasma, 

the  beam  must 

be 

neutralised.  The resulting  neutral 

atoms  cross the magnetic  field  and 

give  up 

their  energy 

through 

collisions  to  the  plasma, 

thereby 

raising  its 

temperature. 

For the  period  1994-98,  the  priority  objective  is to  establish  the 

(2)  Radio  Frequency  Heating:  Energy 

engineering  design  of  the  Next  Step within  the  framework  of  the 

quadripartite  cooperation  between  Euratom, Japan,  Russia and  the 

USA on the  Engineering  Design Activities  for  the  International 

can be absorbed by the plasma  from 

high power  radio-frequency  waves. 

The frequency of operation  is chosen 

to be close to that at which  the ions 

or  electrons  orbit  or  gyrate  in  the 

Thermonuclear  Experimental  Reactor  (ITER-EDA).  Specialized  studies 

magnetic  field. 

are also  needed  to  look  at  possible  improvements  to  concepts  in 

plasma  physics and  plasma  engineering,  as well  as to  carry  out  the 

long  term  technology  developments  recuired  for  progressing  towards 

the  exploitation  of  fusion  as an energy  source.  The  results  of  such 

studies  will  be of  benefit  both  in the  operation  of  ITER and, in the 

longer  term,  in the  conceptual  definition  of  DEMO. 

10 

J E T, E U R A T OM A ND O T H ER F U S I ON P R O G R A M M ES 

Concept Improvements^) 
^ ­— — ^ Í Ni 

TL \1^- ^N 

i f Next­Step \ / \ 
W ITER W DEMO 

^

-^ 
/ S p e c i a l i s e d^ 
/ Devices ^ 

© 

\ In j 

\Associations/ 

^ ^T 

ir rí^—^ 

Long Term Technology J) 

JG93.604/7 

Fig.3: Interactions of European and International Activities. 

The proposed strategy calls for the simultaneous development of 

three areas of activity (see Fig.3), on which efforts will be 

corcentrated mainly by means of shared­cost actions: 

Next Step Activities: design proper, and R&D supporting design, 

construction and operation of the Next Step; 

Concept Improvements: Next Step and DEMO oriented R&D on 

plasma physics and engineering; 

Long Term Technology: DEMO and reactor­oriented R&D on 

technology. 

1994 Achievements 

In the frame of the 1990­1994 Fusion Programme (Council Decision 91/ 

678/Euratom of 19 December 1991, OJ No. L 375, 31.12.1991), a large 

fraction of the activities in 1994, including those on JET and within the 

Associated Laboratories, was in support of the Next Step. In particular, 

the major upgrading of the JET device with the installation of a pumped 

divertor to establish reliable methods of plasma purity control and 

essential for supporting the Next Step design. 

The Next Step design has progressed in the framework of the 

quadripartite Agreement (Euratom, Japan, Russia and USA) on 

cooperation in the ITER­EDA. The overall programme objective of ITER is 

"to demonstrate the scientific and technological feasibility of fusion 

ι ι 

J E T,  E U R A T OM  A ND  O T H ER  F U S I ON 

P R O G R A M M ES 

Principal  Para 

meters  of  ITER 

Major  Radius 

Minor  Radius 

Elongation 

Plasma  Current 

Toroidal  Field 

Fusion  Power  . 

Burn  Time 

8.0m 

3.0m 

1.6 

24MA 

5.7T 

1.5GW 

1000s 

Active 
Control  Coils 

THE ITER DEVICE 

Cryostat 

Central Solenoid 

Shield/Blanket 

Plasma 

Vacuum 
Vessel-Shield 

Plasma Exhaust 

Divertor  Plates 

energy for  peaceful  purposes."  ITER would accomplish this objective by 

"demonstrating  controlled  ignition and extended  burn of  deuterium-

tritium  plasmas, with  steady-state  as an ultimate goal, by demonstrating 

technologies  essential to a reactor  in an integrated system, and by 

performing  integrated testing of  high  heat-flux  and nuclear  components 

required to  utilize fusion energy for  practical purposes".  The 

Agreement further  specifies that the  ITER-EDA - to  be conducted  by the 

four  Parties under the auspices of the IAEA (International  Atomic  Energy 

Agency) and carried out  by a Joint Central Team (JCT) located in three 

internationally  staffed joint work  sites in San Diego (USA), Naka (Japan) 

and Garching  (EU) and by four  Home Teams - shall be implemented by 

two  or  more Protocols.  The ITER Council, in January  1994, considered 

an Outline  Design report  produced  by the ITER JCT as "an  acceptable 

basis for  consideration  by the  Parties"  to  proceed, Protocol  1  was 

completed  and Protocol 2 (covering the period  up to the scheduled 

completion  of  the  EDA in July  1998) was signed on 21  March  1994. 

The  Next  European  Torus (NET) Team  pursued  its  Next Step 

related  activities,  in  particular  in assisting  the  Euratom  Home Team 

12 

J E T,  E U R A T OM  A ND  O T H ER 

F U S I ON 

P R O G R A M M ES 

(HT) Leader  in the performance  of  his duties and responsibilities in 

the  ITER frame.  An industrial  grouping  was selected to contribute to 

the  overall  design  activities for the Next  Step.  Lists of qualified  firms 

have  been established  for the R&D and supply  of  prototypes  for the 

Next  Step in fifteen  technologies  specific  to fusion. 

The Safety and Environmental  Assessment of  Fusion  Power 

(SEAFP) was performed  as a collaborative  effort  by the NET Team, the 

Euratom-UKAEA  Association  and a temporary  grouping  of  industries, 

and with  the participation  of other  Associations  and the Joint 

Research  Centre  (JRC).  The report  will  be issued in early 1995. 

The conceptual  design of a powerful  source of  high-energy 

neutrons  for testing  fusion-relevant  materials  has started  in  the 

mu tilateral  framework  (presently  Euratom, Japan, USA) of an 

Implementing  Agreement  in the frame of the IEA (International 

Energy  Agency,  Paris): the development  and characterisation of 

radiation-resistant  and low-activation  materials  is underway.  R&D on 

tritium  breeding  blankets was pursued, with  the aim of  building 

DEMO-relevant  blanket  modules to be tested  in ITER. 

In the Associated  Laboratories, specialised devices have made 

contributions to the development  of the physics base for ITER and 

DEMO.  They explore, in support of the main objectives of the 

programme, the accessible parameter  range and the possible  modes of 

operation for the different  confinement  concepts and configurations. 

Impurities 

Impurities  released  from 

interactions 

between 

the  plasma  and  material 

surfaces  can  have  major  effects  on 

plasma  behaviour  by causing: 

(a)  increased  radiation  losses; 

(b)  dilution  of  the  number  of  ions 

available  in  the  plasma  between 

which  fusion  reactions  can  occur. 

A  measure  of the overall 

impurity 

level is given  byZell  which  is defined  as 

the  average  charge  carried  by  the 

nuclei  in the plasma.  A pure  hydrogen 

plasma  would  have  Ζ H  =1  and any 

impurities  in  the plasma  would  cause 

this  value  to be increased. 

In JET,  Zefl 

is generally  in the range  from  1.2-3. 

Major  energy 

losses  can  result 

from  two  radiation  processes: 

•  Bremsstrahlung 

Radiation 

-

radiation  is emitted  when  electrons 

are decelerated  in the electric  field 

of an ion.  The amount  of  radiation 

emitted 

increases  with  Zell. 

Bremsstrahlung  radiation  imposes 

a fundamental  limit  to the  minimum 

plasma  temperature  that  must  be 

attained  in a fusion  reactor; 

•  Line  Radiation  -  heavy  impurities 

will  not be fully  ionised  even in the 

centre  of  the plasma  and  energy 

can  therefore  be lost  through  line 

Furthermore, these devices provide testbeds for the development of 

radiation. 

new concepts and techniques for plasma engineering and wall 

tecinology.  These also serve for fundamental  fusion  physics studies, for 

the development  of diagnostics, for the preparation of collaboration on 

larger  devices, for innovative studies and for the training of young 

professionals, as well as being the link which  allow the incorporation of 

university  research into theoretical, numerical or diagnostics activities. 

Progress in the understanding  and control of confinement, plasma-

wall  interaction, fuelling and exhaust, as well as heating and current 

drive was achieved on the specialised Tokamaks (TORE SUPRA, ASDEX-

Upgrade, TEXTOR, FTU, TCV, COMPASS, RTP, ISTTOK, etc) and in  the 

accompanying  programmes. 

Considerable  effort 

is  made  to 

keep  the level  of impurities  in the JET 

plasma  to  a  minimum.  The  vacuum 

vessel is baked  at 300 "C to remove  gas 

\  particles  trapped  on  the  vessel  walls 

which  might  be  released  by  plasma 

bombardment. 

Interactions  between  the  plasma 

and  vacuum  vessel walls  would  result 

in the release of heavy metal  impurities. 

To reduce  this possibility,  the edge  of 

'  the  plasma  is  defined  by  upper  and 

lower  belt  limiters.  These are  cooled 

structures  circling  the outboard 

torus 

wall  with  carbon  or  beryllium 

tiles 

attached.  Carbon  and beryllium  have 

a  relatively  low  electric  charge  on the 

Studies on alternative  lines  in toroidal  magnetic  confinement 

nucleus. 

were  pursued: absence of disruptive  density  limitation  on the W7-AS 

1  3 

J E T,  E U R A T OM 

A ND  O T H ER 

F U S I ON 

P R O G R A M M ES 

Stellarator;  operation  of  the  RFX Reversed-Field  Pinch to  higher 

currents  and further  diagnostics  developments  (e.g.,  polarimetry); 

start  of  operation  of  the  EXTRAP-T2  Reversed-Field  Pinch and  of  the 

TJ-1U  Stellarator;  and  engineering  design  and  prototype 

development  for  a possible  large  Stellarator  (W7-X). 

In technology,  activities  include,  in  particular:  superconducting 

coils  development,  plasma  facing  components,  fuel  cycle,  remote 

handling,  and  operational  and  environmental  safety. 

The scientific  and technical  achievements  of  the  Community 

Fusion  Programme  place  Europe  in the  forefront  of  world  fusion 

research.  Concentration  on the  most  successful  toroidal  magnetic 

confinement  line, the  tokamak,  and  on  a few  promising  lines  akin  to 

it  - while  keeping  a watching  brief  on  inertial  confinement  fusion 

and  on other  approaches  - continues  to  be justified  by the  results. 

Management  of  the  Programme 

All magnetic fusion research is integrated  into one Community Fusion 

Programme which  presents itself as a single body in its relations with 

other world fusion programmes. The European Commission, assisted by a 

Consultative Committee for the Fusion Programme (CCFP) composed of 

national  representatives, is responsible for the implementation of the 

Fusion Programme. The Programme operates principally through: 

contracts of Association with  Member States (plus Switzerland) or 

organisations within  Member States, the JET Joint Undertaking, the NET 

Agreement which takes into account Euratom participation  in the ITER-

EDA, the JRC, contracts of  limited duration (in particular  with 

organisations  in Member States without Association) and industrial 

contracts.  Through the multipartite Agreement for  "Promotion of  Staff 

Breakeven 

This  condition 

is  reached  when 

the 

power  produced  from  fusion  reactions 

is equal to that necessary for  maintaining 

the  required  temperature  and  density 

in  the plasma  volume. 

Ignition 
Ignition  of  a mixture  of  deuterium  and 

tritium  would  be reached  if  the  power 

Mobility", the mobility of scientists and engineers was developed.  In 

produced  by  the  alpha-particles  (20% 

coordination with the  "Human  Capital and Mobility"  programme, 

of  the  total 

thermo-nuclear 

power) 

released  from  the  fusion  reactions  is 

fellowships were awarded. Dissemination of  knowledge and exploitation 

sufficient  to  maintain  the  temperature 

of  results was performed through  laboratory  reports, publications in 

of  the  plasma. 

scientific journals, workshops and conferences.  An itinerant  exhibition 

was further  developed and run by the  "Fusion  EXPO" consortium. 

Community  financial  participation  continued  to  be about  25%  of 

the  running  expenditure  of  the  Association,  45%  of  capital  cost  of 

projects  having  been  awarded  priority  status,  and  80%  of JET 

1  4 

J E T,  E U R A T OM 

A ND  O T H ER 

F U S I ON 

P R O G R A M M ES 

expenditure.  Currently,  Community  funding  of  fusion  research is 

about  200  MioECU  per  annum: when  funding  by  national 

administrations  and/or  bodies  is taken  into  account,  the  expenditure 

on  fusion  from  all  sources  in  Europe amounts  to  about  450  MioECU 

per annum.  About  1,750  professional  scientists  and engineers  are 

currently  engaged  in fusion  research  in  Europe. 

The  1994  -  1998  Fusion  Programme 

The  March  1994, assessment  by the  CCFP of  the  evolution  of  the 

Programme  in the  medium-term  was  instrumental  in the  preparation 

of  the  1994-1998  Fusion  Programme  Decision  (Council  Decision  94/ 

799/Euratom  of  8  December  1994, OJ No.  L 331, 31.12.94). 

Following  this  decision, the  envelope  for  fusion  activities  amounted 

to  840  MioECU  (of  which  46  MioECU was  for  fusion  activit es at  the 

JRC).  Such a figure  could  possibly  be increased  if  the  overall  budget 

for  the  two  Framework  Programmes  (European  Union  and  Euratom), 

presently  established  at  12,300  MioECU, were  to  be  increased  by up 

to  700  MioECU  in  1996. 

In the frame of this  Programme, the following  actions  have been 

undertaken or  initiated: a proposal to extend JET until  1999 to  perform 

specific tasks in support  of  ITER, is being considered; proposals for 

upgrading  existing devices (ASDEX-Upgrade, TORE-SUPRA) and for  the 

construction  of  new ones (Wendelstein  7-X,  MAST) are being assessed; 

a proposal for joining the activities of  three Associations  into a 

"Trilateral  Euregio Cluster"  is under consideration. 

International  Collaboration 

The Community  approach  has led to  an extensive  collaboration 

between  the  fusion  laboratories.  For example,  most  Associations 

undertake  work  for  other  Associations.  The Associations  are 

partners  in JET, NET and  ITER and  carry  out  work  for  them  through 

various  contracts  and  agreements.  The  Programme  has built  across 

Eu'ope  a genuine  scientific  and technical  community  of  large  and 

small  laboratories,  readily  able to  welcome  newcomers  (as it  is the 

case with  imminent  signature  of  a contract  of  Association  between 

Euratom  and TEKES-Finland  and with  the  exploratory  discussions 

concerning  a new  Association  in Austria), and  directed  towards a 

common  goal.  The  leading  position  of  the  Community  Fusion 

1  5 

J E T,  E U R A T OM  A ND  O T H ER  F U S I ON 

P R O G R A M M ES 

Programme  has also  made  Europe  an attractive  partner  for 

international  collaboration.  Apart  from  the  most  far-reaching 

collaboration  illustrated  by the  ITER project,  bilateral  Framework 

Agreements  have  been  concluded  (with  USA and Japan)  or  are to  be 

concluded  (with  the  Russian  Federation).  Also, a Memorandum  of 

Understanding  with  the  Government  of  Canada, which  is ready  for 

signature, will  cover  inter  alia the  involvement  of  this  country  in the 

Euratom  contribution  to  the  ITER-EDA.  Finally, there  are  nine 

Implementing  Agreements  in the  frame  of  the  IEA (the  latest  one, on 

"Nuclear  Technology  of  Fusion  Reactors", was signed  in  1994). 

Large  International  Tokamaks 

In  1994, JET, the  largest  tokamak  in the  world,  restarted  operation  in 

its  new  divertor  configuration.  It  soon  demonstrated  the  high  power 

handling  capabilities  of  its  new  divertor  targets  and that  the  severe 

impurity  influxes  ("carbon  blooms"),  which  previously  terminated 

high  performance  plasmas, were  eliminated. 

Further, the  large TFTR tokamak  at PPPL, USA, has made a major 

contribution  to the fusion effort with  its ex:ensive tritium  programme. 

This has now  lasted well over a year and has recently produced  in excess 

of  10MW of  peak fusion  power. 

In addition, the large Japanese tokamak JT-60U continues to 

improve  its performance  by using boronisation  of  its vacuum vessel. It 

Disruptions 

TViere is a maximum  value  of density 
which can be contained  with a  given 
plasma curren t. If this value is exceeded 
a disruption occurs  when the  plasma 
confinemen t is suddenly destroyed and 
the plasma  current falls to zero in a 
short  period  of  time.  Under  these 
conditions high mechanical and thermal 
stresses are produced on the machine 
structure. Disruptions  are thought to 
be  caused  by  certain  instabilities 
developing  on  specific  magnetic 
surfaces. 

TFTR 

JET 

ITER 

i Η D 

3.0m 

'8.0m 

1.6 

5.7T 

30-200MW 

30  - IGNITION 

24MA 

^ 

« 

MINOR  RADIUS 

MAJOR  RADIUS 

ELONGATION 

TOROIDAL  FIELD 

INPUT POWER 

a 

R 

κ 

Β 

Ρ 

FUSION  FACTOR 

QDT 

PLASMA  CURRENT 

I 

0.85m 

2.5m 

1.0 

5.2T 

32MW 

0.3 

3MA 

1.25m 

2.96m 

1.8 

3.45T 

36MW 

1.1 

7MA 

Fig.4: Operating  parameters  of  three  large  tokamak  designs 

16 

J E T, 

E U R A T OM  A ND  O T H ER 

F U S I ON 

P R O G R A M M ES 

MINOR  ELONGATION  MAJOR 

FLASMA 

TOROIDAL 

INPUT 

MACHINE  COUNTRY 

RADIUS 

κ 

RADIUS  CURRENT 

FIELD 

POWER  START DATE 

a(m) 

R(m) 

l(MA) 

B(T) 

P(MW) 

Jfcl 

tc 

JT-60U 

JAPAN 

TFTR 

TORE-

USA 

SUPRA 

FRANCE 

T-15 

Dlll-D 

CIS 

USA 

1.00 

0.85 

0.85 

0.80 

0.70 

0.67 

ASDEX-U 

GERMANY 

0.5 

FT-U 

ITALY 

0.31 

1  8 

1.6 

I.O 

l.O 

1.0 

2.5 

1.6 

1.0 

2.96 

3.2 

2.50 

2.4 

2.4 

1.67 

1.65 

0.92 

7.0 

4.0 

2.7 

2.0 

2.0 

3.0 

1.2 

1.2 

3.5 

4.2 

5.6 

4.2 

4.0 

2.1 

2.7 

7.5 

42 

40 

40 

22 

-

22 

7 

-

1983 

1991 

1982 

1988 

1989 

1986 

1991 

1988 

Table  2:  Large Tokamaks  operating  around  the World 

has achieved the highest fusion  product  (ητΕΤ) of  12.3x1020m"3skeV at 

an ion temperature of  41 keV, with an equivalent  QDT value of  -0.6. It 

has also reached quasi-steady-state conditions  in an ELMy Η-mode  in 

deuterium with  a high fusion product  of 4.5x1020nrr3skeV and an 

equivalent  QDT of  0.25-0.36 

Now that  the  ITER Engineering  Design Activity  (EDA) is  underway, 

achievements  in tokamak  research  and, particularly,  for  the  largest 

tokamaks  (Fig.4),  have  become  even  more  relevant.  Table  2 sets  out 

an overview  of  the  larger  tokamaks  worldwide, together  with  their 

main  parameters  and  starting  dates. Considerable  progress  has been 

made  by these tokamaks,  and these  are detailed  below. 

MACHINE 

JET 

JT-60U 

TFTR 

Dlll-D 

ELECTRON TEMPERATURE 

T.(keV) 

11 

10 

11.5 

6 

ION TEMPERATURE 

T(keV) 

DURATION (s) 

FUSION PRODUCT 

18 

1.5 

41 

0.6 

44 

0.5 

5.5 

0.5 

nTTE(X1020m-3keVs) 

10 

12.3 

6.6 

5 

Table  3:  Fusion  Products  in Large  Tokamaks 

1  7 

Density  Control 

Increasing the density can be achieved 

by introducing additional gas into the 

vacuum vessel,  by the injection of 

energetic neutral atoms (neutral beam 

heating) and by solid pellet injection. 

Increasing the input power to the 

plasma  through  additional  heating 

raises  the  electron  density  limit. 

However,  problems can occur when 

this heating power is switched off, if 

the  electron density is too high.  To 

overcome this problem, the outer layer 

of  the plasma  is moved,  prior  to  the 

switch-off  point, so that it  bears  on 

the  carbon  tiles covering the  inner 

wall.  The tiles have been found, to 

provide  a pumping  mechanism  for 

removing particles so that the density 

can be reduced below the critical limit. 

J E T, 

E U R A T OM 

A ND  O T H ER 

F U S I ON 

P R O G R A M M ES 

Peak  Fusion  Performance 

The highest peak  performance  as measured with the triple  product of 

deuterium  density, energy confinement  time and central ion 

temperature  ηΒτΕΤ is now taken over by JT-60U with  a value of 

12.3x1020nr3keVs  at a central  ion temperature  of 41 keV. This is 

calculated to yield a fusion  efficiency  QDT of  0.6. This is still surpassed by 

the  QDT of JET which was close to  unity in  1991. During  1994, JET has 

also reached high values in the divertor  configuration, with  triple 

products of  7.4x1020nr3keVs.  Dlll-D and TFTR in the  USA have both 

maintained  considerable  progress and have doubled their  previous 

records this year. 

Peak values  of  the  fusion triple  product  obtained  in the  four 

largest  tokamaks  are shown  in Table 3. 

Technological  Developments 

The heat  load to the divertor  and the  particle  removal  rate both pose 

crucial problems that  must be solved before a tokamak  reactor  can be 

built.  Several forms of advanced divertors are emerging, generally 

involving  closed divertors with  large active cooled areas. This allows for 

impurity  retention as well as a high divertor  density, so that  most of  the 

power  entering the divertor  can be radiated before it strikes the target 

plates. This power  can then  be absorbed over a much  larger area. 

In JET, divertor  plasmas heated with  up to 26MW  have been 

thermally stable and maintained at high density with  more than 70% of 

the power  radiated. However, the discharge returned to the low 

confinement  (L-mode) regime. ASDEX-U has achieved similar  results with 

high radiation and detached divertor  in the Η-mode regime. 

In the JET and Dlll-D machines, divertor  cryopumps  have proven to 

be highly successful  in maintaining  quasi-s:eady-state  density  regimes at 

high fuelling  and pumping  rates. In addition, TORE-SUPRA in France has 

demonstrated  that an ergodic  divertor  with  impurity  puffing can 

efficiently  radiate away incoming  power  and retain those impurities. 

Tritium  Experiments 

After  the successful first tritium  operation  carried out  in JET during  1991 

with  a D-T mixture containing  11 % T, tritium  experiments  have been 

carried out  in TFTR, USA, using a 50:50  D-T mixture. JET has plans to 

carry out further  tritium  experiments during  its  1996 campaign. 

ι s 

J E T, 

E U R A T OM 

A ND  O T H ER 

F U S I ON 

P R O G R A M M ES 

TFTR operation  has been  producing  fusion  power  exceeding 

10MW.  It  has carried  out  many  D-T  experiments  over  a period  of 

about  a year  and  has processed  380kCi  of  tritium  in that  period.  In 

total,  680MJ  of  fusion  energy  has been  produced.  In addition, TFTR 

has  detected  the  first  signs  of  alpha-particle  heating.  Plasmas of 

2.7MA  current  in a 5.5T  toroidal  field,  in the  so-called  supershot 

regime,  have  been  heated  by 39MW  of  neutral  beam  power  of  50% 

deuterium  and  50%  tritium  neutrals.  Even though  the  Toroidal 

Alfven  Eigenmode  (TAE) activity  increased with  the  obtained  fusion 

power,  no  major  effect  on  the  alpha-particles  was observed. The 

helium  (cold alpha-particles)  transport  was found  to  be  roughly  the 

same as for  the  hydrogenic  species. 

An extension  of the JET Programme to the end of  1999 is currently 

being  proposed. Two  periods of  D-T operation are foreseen for the JET 

prog'amme  to the end of  1999, subject to the  necessary approvals. JET 

will  make important  contributions to  D-T physics for  ITER (including 

Η-mode threshold, ELM and confinement  behaviour,  and some radio 

frequency  heating studies). The first  period (DTE-1), scheduled for the 

second half  of  1996, would check whether the  more favourable 

confinement  found  during  D-T operation  in TFTR extends also to the 

ITER-relevant divertor  and operating conditions  in JET.  In addition, it 

should demonstrate  long  pulse fusion  power  production. JET alone can 

investigate these topics in its divertor  configuration. The second period 

(DTE-2), scheduled for  1999, should  permit a thorough  study of  D-T 

plasmas with  enhanced  levels of alpha-particle  heating, capitalising on 

the performance  improvements  achieved in the preceding  experimental 

campaigns with  deuterium. 

1  9 

T E C H N I C AL 

S T A T US 

OF 

J ET 

Technical  Status  of JET 

Introduction 

JET entered  1994 nearing the end of  the longest and most extensive 

modification  since the  initial assembly of  the device. Two faulty  toroidal 

magnetic  field (TF) coils had been replaced and the  new  pumped 

divertor  configuration  had been installed. The interior  of  the torus had 

been completely  rebuilt to  include four  new divertor  coils, an inertially-

cooled divertor  target  structure and a cryogenic vacuum pump. 

The main objective  had been to  install the components  of the 

pumped  divertor  and undertake  its associated system  modifications, 

which were aimed at establishing  in deuterium  plasmas  "reliable 

methods of  plasma  purity  control  under  conditions  relevant for the Next 

Step Tokamak".  The main components  of the divertor  are, as follows: 

• 

the four  poloidal coils can create various magnetic  configurations, 

with  the X-point far away from  the target  plates, and includes the 

capability  of X-point  sweeping to spread the thermal  load. The coils 

are contained  in a 1.2mm  thick  inconel casing, on which  the rest of 

the divertor  components  are installed; 

• 

the target  plates, arranged  in a U-shaped  contour,  collect  power 

released from the  plasma. These consist of an inertially  water-cooled 

structure, which support  carbon fibre composite  (CFC) tiles, shaped 

to  maximise the  power wetted surface. The CFC will  be replaced in 

1995 with  beryllium tiles, for  comparison  of  plasma  performance; 

• 

the toroidal  cryopump  controls  plasma density in the divertor  region. 

It is anchored to the outer  coil casing and its main components  are a 

water-cooled  baffle, a liquid nitrogen  cooled copper  backed panel, a 

set of  helium cooled  pipes and a chevron structure. 

2  1 

T E C H N I C AL 

S T A T US 

OF 

J ET 

Fig.5:  Inside of vessel after  completion of installation of components 

The heating and current  drive systems were also modified and now 

comprise an upgraded  neutral  beam (NB) system to  inject  7.8MW  of 

power at  140keV and  13.6MW at 80keV, new  Ion Cyclotron  Resonance 

Frequency (ICRF) antennae (designed to couple 20MW) and a new 

Lower  Hybrid (LH) launcher (designed to couple  10MW). Other  new 

installations  included a system for  plasma current  and shape control, a 

Fast Radial Field Amplifier  for vertical  position control, eight  Saddle Coils 

inside the vacuum vessel for  disruption control and for the study of 

Toroidal Alfvén  Eigenmodes (TAEs) and an extensive array of  new 

diagnostics, especially for  divertor  measurements. 

The shutdown  was successfully  completed  with  pumpdown  of  the 

torus  in  mid-January  1994. The  inside of  the  vessel on  completion  of 

the  shutdown  is shown  in  Fig.5. The first  plasma  in the  new  pumped 

divertor  configuration  was  produced  in  m d-February  and  by mid-

March  successful  2MA  divertor  plasmas  had  been  established.  During 

1994, the  plasma  current  was  increased to  5MA,  the  total  heating 

power  to  26MW,  the  stored  energy  to  11.3MJ  and the  neutron  rate 

to  4x1016 neutrons  per second. 

22 

T E C H N I C AL 

S T A T US 

OF 

J ET 

The high power  handling  capability of  the Mark  I divertor  target has 

been demonstrated  and the severe impurity  influxes (carbon  "blooms"), 

which  previously terminated  high  performance  plasmas, have been 

eliminated. The cryopump  reduces recycling, eliminates the effects of 

wall saturation  (observed in previous long pulse operation), allows 

effective  particle control, and generally  allows  higher  performance. 

The two  NB injectors  have routinely  provided  up to  19MW and 

140MJ has been injected during  a 20s ELMy H-mode.  13MW of ICRF 

power  has also been coupled, but this  is limited  by a combination  of 

unsatisfactory  control  electronics, unequal coupling of the straps of the 

Power  Supplies 

antennae array and low  power transfer to the plasma under some phase 

The  electric  power 

to  the  JET device 

conditions. New control electronics systems are currently  being  installed 

and will  be tested with  plasmas in early  1995. Further antenna 

modifications  are scheduled for  the beryllium tile exchange  shutdown 

during  an experimental  pulse is  counted 

in  hundreds  of  megawatts. 

An agreement  with  the  Generating 

Boards  allows  up  to  575MW  of  pulse 

power 

to  be  taken  directly 

from 

the 

and the Mark  II divertor  shutdown  planned to start  in March and June 

400kVgrid,  which  after 

transformation 

1995, respectively.  Combined  NB and ICRF heating  powers  up to  26MW 

down  to  33kV  is  fed  to  the  JET loads 

through  a system  of  circuit  breakers. 

have been injected and close to 3MA  has been driven  non-inductively 

Two  flywheel  generators  are  used 

with  up to  6MW  of  LH power. 

to  provide 

the  peak  power 

for 

the 

toroidal  magnetic  field  coils and  ohmic 

The saddle  coils  have  been  used for  initial  experiments  on TAE 

heating  circuit.  Each of  the  generators 

modes and the  disruption  feedback  stabilisation  system  is in the  final 

stages  of  commissioning.  Only the  lower  saddle  coils are  now 

available  for  experiments,  since the  upper  saddle  coils were  disabled 

in  September  1994  after  being  damaged. 

During  1994, three  manned in-vessel interventions were carried out 

has  a  rotor  9m  in  diameter  weighing 

775  tonnes.  Between  pulses,  8.8MW 

pony  motors  are  used  to  increase  the 

speed  of  rotation.When 

power 

is 

required 

for  a  JET  pulse, 

the 

rotor 

windings 

are  energised 

and 

the 

rotational  energy  of 

the  flywheel 

is 

converted 

into  electrical  energy.  O n 

in full  air-suits (since beryllium evaporation  had been used). In late 

slowing  down  from  the maximum  speed 

March  1994, following  first  experiences with  the restart of  operations, 

of??5rpm 

to half speed, 

thegenerators 

can reach  deliver  2.6GJ  of  energy  with 

an inspection  showed that  no significant  in-vessel damage had 

a peak  power  output  of  400MW. 

occurred.  In September/October  1994, damage to the  upper  saddle coils 

was stabilised and a further  diagnostic  system was installed. In early 

November  1994, the beryllium  evaporator  heads were removed, 

strengthened  and reinstalled. 

The following  sections  detail  the  technical  achievements  made 

during  1994. 

Technical  Achievements 
Commissioning  of  Divertor  Coils 

The purpose of the JET coil system is to  establish, maintain  and control 

the tokamak  plasma configuration  (Fig.6). It  includes: the toroidal coils, 

23 

T E C H N I C AL 

S T A T US 

OF 

J ET 

Divertor coils 
(D1,  D2,  D3andD4) 

Fig.6: Cross-section of JET showing toroidal, poloidal and divertor coils 

which establish the toroidal  magnetic field; the poloidal coils P1, acting as 

primary windings of the tokamak transformer; and coils P2, P3 and P4 to 

control  plasma radial position, vertical position and shape. The divertor 

coils D1,  D2, D3 and D4 were installed inside the vacuum vessel, and 

now form an integral part of the poloidal system. In conjunction  with 

existing coils, these coils establish and control the new divertor  magnetic 

configuration. 

In early  1994, the divertor  coils were progressively commissioned  in 

line with  experimental  requirements, which  was most activity  centred 

around the thermal  performance  of the coils and cooling system. The 

coils are heated from two  sources: the ohmically generated  heat due to 

the coil current;  and the  heat transfer from the surrounding vacuum 

vessel and in-vessel components. The coils are cooled by freon  pumped 

through the conductor  cooling  channels. The cooling system is designed 

24 

T E C H N I C AL S T A T US OF J ET 

8 

7 

6 

5 5 

S 4 
o 

r
e
w
o
P

o
c

2 

1 

• Estimated power (kW) 

■ Measured power (kW) 

■ 

■ / 

■ · 

n/ 

/^ 

/^ 

S\ 
) 50 100 150 200 

°C 

Vessel temperature (°C) 

1 I 

è 

r 

o 
fi 

σι 
O 

250 

Fig.7: Heating power to Coil D4 with target plates uncooled 

to maintain the measured temperature difference across the coils at 

1 5'C. This limits the thermally developed shear stress in the inter-turn 

insulation to 10MPa. 

The thermal performance of the coils has been monitored. The aim 

is to estimate whether any part of the epoxy-glass insulation mignt be 

subjected to temperatures which could degrade its properties as an 

insulating or structural material. Since direct measurements of the epoxy 

tempe'ature are not possible, the heat power load to the coils, and the 

temperatures on the inconel casing have been measured. The present 

Mark I divertor target plate assembly is supported on the divertor coils, 

and its alignment ultimately relies on the mechanical stability of the 

epoxy. Calculations were also made to evaluate faults such as loss of 

vacuum or failure of the coil cooling system. From the thermal point of 

view, the epoxy is a matter of concern. The temperature of the casings 

is monitored, at four points per coil, by thermocouples welded to the 

casing. To assess the power scaling and the main heat transfer 

mechanism to the coils, measurements were performed with the 

vacuum vessel at several temperatures in steady state conditions. Tests 

were performed, to assess the effect of a vacuum loss in the vessel. Η 

and N, gases were injected up to 10mbar pressure, with the vessel at 

25 

 
 
T E C H N I C AL 

S T A T US 

OF 

J ET 

y' 

/ 

V  - ΤΊ 
frt\  ΛΓ\Ί 
t x p.  (α? 1 U'  by 

fi  r 

/ 

/ 

/ 

I 

/ 

20 

l

O

ra 
o 

i

)
n
s
e
r

o 
S 

1  10 

T3 
o 
CL 

— 

5 

/ 

I 
ƒ 

y 

^^ 

°, 

D 

I

I

2 

4 

II 

6 

8 

Total dose  (x106Gy) 

P5 
fi 
T. 
σι 
O 

Fig.8:  Gas  released  under  irradiation 

150'C  and 250°C, respectively,  and the  increased  heat  power  measured 

(see Fig.7). 

Molecular  conduction  appeared  to  be the  main  means of  heat 

transfer  across the  inter-space,  between  casing  and  epoxy  insulation. 

This  interspace, which  is pumped  down  to  about  1 millibar  was 

formed  during  resin  polymerisation. The epoxy  layer  under  coils D2 

and  D3 was  most  critically  affected  by the  vessel temperature,  due  to 

the  larger  than  expected,  heat  conduction  from  the  target  plate 

graphite  tiles to  the  support  structure.  As a consequence,  baking 

and operation  with  the  target  plate  uncooled  is now  restricted  to a 

maximum  vessel temperature  of  200°C  per  hour,  and with  target 

plates  cooled  to  320°C. 

Irradiation  Tests  on  Divertor  Coil  Insulation 

In the  D-T  phase, the  divertor  coils will  operate  under  neutron  and 

gamma  fluxes  which  may affect  the  characteristics  of  the  electrical 

insulation.  To assess the  effect  of  radiation  damage, test  samples 

were  irradiated  in conditions  close to  those  during  the  active  phase. 

Irradiation  conditions  achieved were  fast  neutron  fluence  rate 

corresponding  to  a dose rate of  6Gys ' and a gamma  rate of  5Gys '. 

26 

 
 
T E C H N I C AL 

S T A T US 

OF 

J ET 

Three  irradiations  were  performed  at total  doses of  105,  106 and 

10'Gy,  respectively. 

Under  irradiation,  covalent  bonds  of  the  epoxy  resin  atoms  were 

broken  forming  free  radicals, which  recombined  in  new  compounds 

with  a teration  of  physical,  mechanical  and chemical  properties. The 

material  lost  tensile,  shear  and  impact  strength  and  became  brittle. 

Moreover,  freed  protons  and  methilic  groups  recombined  producing 

hydrogen  (mostly),  methane  and  ethane.  Up to  4x105Gy,  gas was 

released  during  the  three  exposures, with  a slight  regular  increase 

with  increasing  doses. At  higher  doses, the  production  rate  strongly 

increased with  increasing  doses, sometimes  with  abrupt  changes, as 

showr  in  Fig.8. This behaviour  can  be explained  by the  presence  of 

cavities  in the  resin, either  formed  during  impregnation  and/or 

formed  due to  material  degradation  under  irradiation. The gas 

produced  filled  the  cavities, then  it  was  slowly  released. 

It was  concluded  that: 

• 

loss of  the  insulation  mechanical  properties  preceded  loss  of 

electrical  properties.  The  resin started  to  become  brittle  at doses 

higher  than  106Gy,  but  at  a dose of  107Gy,  the  electrical 

resistivity  value was still  acceptable  for  insulation  purposes; 

• 

radiolysis  was  the  actual  insulation  life  limiting  phenomenon; 

•  gas  release was  not  significant  up to  105Gy,  but  at  higher  doses, 

it  caused  swelling  and  a pumping  system would  be  required  to 

evacuate  the  gas from  the  divertor  coils cases. 

New  Coil  Protection  System 

Due to  complexity  of  the  new  magnetic  coil  system, a new  Coil 

Protection  System  (CPS) has been  introduced  to  protect  the  coils 

against:  electrical  faults; and  electrical,  mechanical  or  thermal  over-

stressing  due  to  operation  outside  safe  limits.  Electrical  faults  are 

detected  by comparing  the  actual  electrical  performance  of  the  coils 

with  a circuit  model  run  in  real time.  Electrical,  mechanical  and 

thermal  stresses are computed  from  the  currents  flowing  and 

voltages  applied  to  the  coils.  If  predetermined  values for  any of  these 

parameters  are exceeded, the  pulse  is terminated. The  protection 

system  is implemented  fully  digitally  using a high  performance  digital 

signal  processor. 

27 

T E C H N I C AL 

S T A T US 

OF 

J ET 

The  radial  and vertical forces  on  the  outer  poloidal  coils  P2, P3 

and  P4 and the  vertical  forces  on the  divertor  coils  are  computed 

using fluxes  measured  by coil  mounted  flux  loops and  measured 

ampere-turns.  It was  not  possible to  fit  vertcal  flux  loops on  the 

divertor  coils, so that  radial forces  acting  on these  coils are  computed 

analytically.  The tensile  and  shear  stress of  each coil  is computed  as a 

linear  combination  of  vertical  force,  radial force  and temperature. A 

simple  model  is used to  estimate  the  temperatures  of  the  epoxy 

insulation  and copper  windings  of  the  divertor  coil. The  inputs  to  the 

model  are the  vessel temperature,  coil  case temperature,  coils 

currents,  coolant  flow  and coolant  inlet  and  outlet  temperatures. 

The system became operational  in May  1994 and was progressively 

commissioned and upgraded to  include  new requirements. Statistics for 

the four-months  of July to October  1994 have shown that the 

protective actions were mainly caused by parameters  being outside 

limits, due to  induced currents and voltages during  disruptions. Early 

experience indicates that the system will  be useful and reliable. Due to 

its flexibility,  protection functions  not  included in the original  design 

have already  been implemented  and others are planned. 

Plasma  Control 

Due to  the  enhanced  asymmetry  of  the  plasma  in the  new  divertor 

configuration  and to  the  strong  coupling  with  the  divertor  coils, 

reconsideration  of  essential  plasma  and  machine  control  and 

protection  was  necessary.  Disruptive  instabilities  frequently  lead  to 

loss of  vertical  position  and, consequently,  to  larger vertical  plasma 

displacements  (which  can  be  up to  1m). The associated  vertical  force 

acting  on  the  vessel  is large  at  high  plasma  current  and when  larger 

shaping  currents  are applied, such as in single  or  double-null 

configurations.  The vertical  instability  produces  potentially  dangerous 

forces  on  in-vessel  elements  such as protection  tiles. 

The  previous  analogue vertical  stabilisation  system  has been 

replaced  by one which  employs  a Fast  Radiei  Field Amplifier  (FRFA) 

and  a fast  digital  signal  processing  and controller  unit. The 

stabilisation  is based on  proportional  feedback  to  the vertical  speed 

of  the  current  centroid, weighed  with  the  plasma  current. The 

average  FRFA current  is kept  near  zero  on  average  by a slower 

28 

T E C H N I C AL S T A T US OF J ET 

„ -1 a) 

„ 1.0 
È Ζ, 

0.5 

„ ­ 0 .2 R , ­ R0 
E. 

­0.6 

~ 4 I 
< „ 'FRFA 
ÍS­

0 — 

> 51 
Ä 0 

­5 

rW 

TrrrViVVrrT^V-'JjW 

1 ¿LL 

15.80 15.85 15.90 15.95 

W Time (s) 

ΡΓ 

15.78 15.79 

15.80 

Time (s) 

Fig.9: (a) Plasma and FRFA signals before and during a disruption; (b) D. plasma position and FRFA signals 

during ELMs 

proportional­integral current feedback. The system is designed to 

stabilise plasmas with instability growth rates up to 1000s'. 

After first operation with plasma, a number of modifications had 

to be made. The switching frequency was unexpectedly high and the 

highest switching level +1 OkV was often invoked at a high rate, causing 

"sparkling" at in­vessel components and poor plasma behaviour. This 

problem was due to larger interference at 600Hz caused by the divertor 

coil power supplies on the plasma feedback signal, and by the high 

noise sensitivity of the original FRFA hysteretic controller. 

The interference was suppressed by abandoning compensation of the 

feedback signal for the unwanted contribution from the divertor currents 

and also by omission of the signal contributions from the lower set of 

poloidal field pick up coils. The FRFA hysteretic controller algorithm was 

altered to avoid multiple switching by noise and to achieve a closer 

similarity with a linear voltage control. Hysteresis is now set to one switch 

level (2.5kV), independent of the instantaneous output level. Practically 

all disruptions and also giant ELMs caused a voltage saturation and a 

current limit trip of the FRFA, leading to a vertical instability and large 

vertical forces on the vessel. Figure 9(a) illustrates a typical case cf sudden 

saturation, probably caused by a giant ELM (at t=15.95s). Stabilisation is 

not regained and the plasma becomes unstable and disrupts. Figure 9(b) 

indicates the response on smaller ELMs. 

29 

T E C H N I C AL 

S T A T US 

OF 

J ET 

Amplifier 4 

Amplifier 3 

y  Signals from pickup coils  \ Ω* 
VP 1 UVP 3U  VP5UVP7U  VP 1 LVP 3L  VP5L  VP7L 
I  1  1  i 

i  1  1  I 

Signal conditioning 

/\ 

REF1 

REF 2 

V or I references  Controller  V or I references 
to amplifiers 

|  to amplifiers 

| 

ι 

ι 

Fig.10: Schematic of disruption feedback stabilization system 

The feedback  signal VREF shows synchronous  perturbations even 

though the vertical current centroid  location zt remains almost  constant 

as opposed to the  radial  position  Rr  This behaviour  suggests that  the 

resaturation of the originally  intended  plasma current  derivative  method 

could be advantageous. As an alternative, the  implementation  of  soft 

X-ray cameras is proposed for feedback  stabilisation. This measurement 

should be free from  magnetic  noise and insensitive to ELMs. 

Disruption  Feedback  Stabilization  System 

The aim of  the  disruption  stabilization  system, shown  schematically 

in  Fig.10,  is the  detection  and  suppression  of  the  m=2,  n=1  MHD 

modes, which  are commonly  observed  to  precede,  and  ultimately 

cause,  major  disruptions  in tokamaks.  While  extension  of  the 

tokamak  operating  regime  by suppression  of  disruptions  would  be a 

major  achievement,  the  capability  of  working  more  closely  to  the 

existing  operational  limits  with  a greater  margin  of  safety would, in 

itself,  represent  a significant  advance. Thus, the  system  is  potentially 

30 

T E C H N I C AL 

S T A T US  OF 

J ET 

-Xpoint dump plates 

, Active saddle coils 

Fig.11 : O v e r v i ew  of  saddle  coils  in  one  Octant  of  t he  vessel 

of  great  value to  tokamak  operations,  in  particular,  in the  context  of 

steady-state  operation  of  a reactor,  where  a major  disruption  may 

have  both  technical  and economic  consequences. 

The disruption  feedback  stabilization  system  consists  of: 

• 

four  disruption  feedback  amplifiers  (DFAS) in the range 0-10kHz; 

•  eight  saddle coils in the vacuum vessel; 

•  saddle coil protection  system, including saddle coil crowbars (SCC); 

•  disruption feedback  controller  (DFC) with  associated  magnetic 

detection system. 

Installation of  the eight  saddle coils was completed  during the 

shutdown.  Each saddle coil covers one quadrant of  the vacuum vessel in 

toroidal  direction  and approximately  60° in poloidal  direction, as shown 

in Fig. 11,  which shows a schematic  overview  of  part of the system. 

The disruption  feedback  controller,  which  is the  key  component 

in the  system,  receives  input  from  four  magnetic  pick-uo  coils  inside 

the  torus,  from  which  the  amplitude,  phase  and frequency  of  the 

precursor  are determined. The  necessary  magnetic  correction  signals 

3  1 

T E C H N I C AL 

S T A T US  OF 

J ET 

Corrugated He pipe 

: 

ι.  , 

/ 

He pipe support wire 

Support strap 

I  Support clamp 

Divertor coil 

LN pipes 

Water cooled baffles 

Back panel 

Fig.12:  Divertor  cryopump 

are then  calculated  and control  signals  sent to  the  DFAS, which 

operates  in current  control. This produces  an appropriate  magnetic 

correction  field,  via the  saddle coils, to  stabilize the  mode  growth. 

Commissioning  of  the  system  is underway  and  initial  open  loop 

power  experiments  have  been  performed  in the  presence of  plasma. 

Although  the  principal  aim of  the  system  is stabilization  of  disruption 

precursors, a wide  range of  further  experiments  is possible. Non-

rotating  modes  due to  error  fields  are of  major  concern  in the  design 

of  ITER and  did  limit  the  operation  regime  of  JET in the  past. The 

DFSS system will  allow  the  physics of  such  modes to  be  investigated 

in detail. Several aspects of  the  physics  of  the  m=2,  n=1  mode can 

also  be addressed.  In addition,  the  influence  of  error  fields  and  MHD 

activity  on the  core and  edge  plasmas can  be investigated. 

Cryopumps  and  Cryoplant 

An  integral  part  of  the  divertor  is a large  cryocondensation  pump 

shown  schematically  in  Fig.12, which  extends  over  the  full  toroidal 

length  of  the  outermost  divertor  coil,  and which  also  acts as the 

mechanical  support. 

The supercritical  helium  loop  used to  supply  the  pumped  divertor 

cryopump  in shown  schematically  in  Fig.13.  Prior  to  connection  to 

the  in-vessel  pump,  the  loop  and the  new  refrigerator  were 

commissioned  "off-line"  using  a spare quadrant  of  the  cryopump 

32 

T E C H N I C AL 

S T A T US  OF 

J ET 

Cryoplant 
T C F 5 0 o r T C F 2 00 

x)  Sep. pot  Y 

5K/12bar 

~5K 
1.5 bar 

Valvebox 

Plant  cold 
ejector 

0.9 bar 

He  tank 

Separation 
pot 
-5K/2.3 bar 

Æ-
Æ 

4K/0.9 bar 

Subcooler 
- 4K/2.7 bar 

5-10g/s 

Cryolines 

* 

Coaxial, LN2 -  shielded"' 

<5.5K/2.5 bad 

25W 

Pump  divertor  cryopump 

Safety 
devices 
balloon 

25W 

25W 

Pump  divertor  cryopump 

îî  25W 

Fig.13:  Schematic  of  supercritical  helium  loop  for  divertor  cryopump 

installed  adjacent  to  the  cryoplant.  The  in-vessel  pump  was  connected 

to  the  cryoloop  in July  and,  following  a  relatively  short  period  of 

commissioning,  entered  routine  operation,  with  an  effective  pumping 

speed  of  2x1 OVs"1 for  the  divertor  structure  plus  the  cryopump.  In 

operation,  the  pump  has  proved  to  be  resilient  to  all  plasma  operation 

scenarios  and  is unaffected  by  plasma  disruption. 

Although  helium  gas  is not  pumped  by  cryocondensation,  it  can  be 

pumped  using  cryosorption  onto  a pre-deposited  layer  of  argon  frost. 

This  technique  has  previously  been  used  successfully  in  the  neutral  oeam 

injectors  and  will  be  applied  to  the  torus  cryopump  using  an  argon  spray 

system  integrated  into  the  liquid  nitrogen  cooled  chevron  structure  of 

the  pump.  Commissioning  of  the  dedicated  argon  gas  introduction 

system  is well  advanced. 

Following  preliminary  operation  to  confirm  satisfactory  performance, 

a  Lower  Hybrid  Current  Drive  (LHCD)  pump  was  installed  on  the  torus  at 

the  beginning  of  the  year  and  brought  into  routine  operation  by  mid-

33 

T E C H N I C AL 

S T A T US 

OF 

J ET 

1994. The pump  has proved to  be a necessary prerequisite to  obtaining 

high  power operation of the  LHCD system. The performance  of  the 

pump,  measured during the  limited time available for acceptance tests, 

conforms to the design values, with  a pumping speed near  105fs~'. The 

pump  has demonstrated  continuous and stable operation for  periods of 

several weeks, including  overnight  helium glow discharge cleaning. 

Neutral  Beam  Heating 

During  the  major  shutdown, the  neutral  beam systems  had  been 

modified  and  upgraded. Some injectors  had  been  converted  to  the 

new  high  current  80kV  configuration  originally  developed  for  the 

Preliminary  Tritium  Experiment.  Eight  of  these were  installed  on  the 

beam system  at  Octant  No. 4. The Octant  No. 8  beam  system 

Neutral  Beam  Heating 

operates  at  140kV.  The injectors  were  recommiss oned  during  the 

The two JET neutral beam systems have 

early  part  of  the  year  and  brought  up to  full  power.  The  high  voltage 

been designed for  long  (~10s)  beam 

pulses. They  have the  unique  feature 

injector  routinely  delivered  up to  7.8MW  to  the  plasma  (-5%  above 

that each injector consists of eight beam 

the  original  design value) whereas the  injected  power  level  of  the 

sources in a single integrated beamline 

system connected to the torus.  The first 

high  current  injector  was  progressively  increased  up to  11.5-12MW 

beam sources were designed to operate 

(50%  above the  original  design value). 

at accelerating voltages up to 80kVand 

in 1990 one system was substituted with 

Commissioning  also took  place  of  the  new  remotely  operated 

units capable of operating up to  140kV. 

vertical  and  horizontal  steering  systems for  each of  the  positive  ion 

In addition, this box was also converted 

to  operate with  helium CHe  and 'He) 

neutral  injectors  (PINIs) which  provide  the  eight  beams  integrated 

beams during 1990. In the D-T phase, 

within  each  injector.  The new  vertical  steering  facility  allowed  the 

one unit will be converted for operation 

with tritium at  160kV. 

power  deposition  profiles  to  be optimised  for  the  upwardly  shifted 

Each  system  is connected to  the 

plasma  equilibria,  characteristic  of  the  new  pumped  divertor. 

torus by a long narrow duct through 

which up  to  12MW of power  can  be 

In addition, the new extended protection assemblies for  both the 

directed. 

neutral  beam drift  injection  ducts were commissioned together  with 

their  extensive thermocouple  instrumentation. This extended  protection 

allowed  safe injection of  neutral  beams down to  plasma currents of 

-1MA,  whereas in previous campaigns, injection  had been limited to 

>1.5MA. This proved invaluable  in high β  experiments. 

These tasks were accomplished  successfully and the neutral  beam 

system operated on  nearly every day of the  1994 campaign. The overall 

statistics for  power  injected into JET with the  neutral beam system are 

shown  in Fig. 14. The system was more heavily used in  1994 than any 

previous year and the proportion  of  pulses at high  power (one quarter 

of  all pulses) exceeded those in previous campaigns.  Figure  14 shows 

that the spectrum of  injected  power was nearly  uniform  up to  16MW, 

34 

T E C H N I C AL S T A T US OF J ET 

Γ~\ 

Π 

/ 

100 

fã 
o 
n 
m 
o 
\_ 
Xi 

E 

ë 50 

, 1 

| 1 

ι —I 

ι —I 

I 1 

ι —I 

— 

/ 

Γι 

0 5 10 15 20 

Power (MW) 

0, 

CO 

σι 
O 

Fig.14: Histogram of injected neutral beam power during 1994 

reflecting the wide variety of experiments, in which the system was 

used. Above 16MW, the number of pulses was reduced as the system 

only attained these power levels reliably from October onwards. 

Radio Frequency Heating 

The ion cyclotron resonance frequency (ICRF) heating system is used for 

high power centralised heating of the plasma, with increased emphasis 

on Fast Wave Current Drive (FWCD) studies with the new antennae. The 

localisation depends mainly on the magnetic field and is insensitive to 

parameters such as density and temperature. Wide­band operation (23­

57MHz) allows variation in both the choice of minority ion species 

heated (H or Hej at present, D in the future D­T phase) and the localised 

position of the heating. With the previous antennae, up to 3.5MW on 

one antenna and 22.7MW total coupled power for 2s were achieved. 

New ICRF antennae have been installed and optimised to the 

geometry of the divertor plasmas. Their location in the torus has been 

revised to give four arrays of two adjacent antenna. Each array has four 

RF radiating conductors (see Fig. 15) or straps, which provide an 

enhanced radiated spectrum. Variation in the relative phase of the RF 

currents in the straps allows this spectrum to be varied for both heating 

35 

T E C H N I C AL 

S T A T US 

OF 

J ET 

Fig.15:  The  n ew  ICRF  a n t e n na 

and current  drive experiments. In addition, the cortrol  electronics have 

Radio  Frequency  Heating 

Ion  Cyclotron  Resonance  Frequency 

(ICRF)  heating  has been  chosen  for JET 

and  the wide  operating  frequency  band 

(23-57MHz)  allows  the  system  to  be 

operated  with  the  various  mixes of  ion 

been completely  rebuilt to allow operation with four  straps closely 

species required  in  the different  phases 

coupled and to  improve the  reliability of the ICRF plant by reducing the 

of  the  scientific  programme  and 

to 

choose  the  location  where  the  heating 

cross-talk  between  modules. The start of  operations was made  difficult 

in  the plasma  occurs. 

by the fact that  both the control electronics and the antennae are new 

The ICRF heating  system  has  been 

designed 

in  eight 

identical  modular 

systems and have required considerable  commissioning  time. Several 

units.  Each  unit 

is  composed  of  a 

problems  have been identified and remedial actions has been taken. 

Two  major  improvements  have been made to the  power 

transmission. Longer  matching trombones  have been added to  many of 

tandem  amplifier  chain,  a  network  of 

coaxial transmission  linesand  matching 

elements and finally on antenna  loca ted 

in  the  vacuum  vessel on  the outer  wall. 

The  eight  RF  generators  produce  a 

the transmission  lines to cope with  the  requirements of tuning  four 

maximum  output  power  of32MW.  The 

antenna straps, and conjugate  box decouplers are installed on three 

net  power  coupled  to  the  plasma  has 

reached  22.7MW, 

compared 

with 

modules to compensate for  power  coupled across straps of  an antenna 

theoretical  limit  of  24MW. 

when  phased for  current  drive operation. In spite of  early difficulties, a 

record power of  13.5MW  had been launched  in a divertor  plasma, 

compared to  12MW  in X-point  plasmas in the preceding campaign. 

Lower  Hybrid  Current  Drive 

The Lower  Hybrid Current  Drive (LHCD) system, operating  at 3.7GHz, is 

capable of  driving a significant fraction of  the toroidal  current  in the 

plasma. This is achieved  by launching an RF wave predominantly  in one 

toroidal  direction. This wave accelerates the high energy electrons  in the 

plasma and so drives a current.  It may be used to stabilise sawtooth 

36 

T E C H N I C AL 

S T A T US 

OF 

J ET 

Fig.16:  T he  LHCD  l a u n c h er  (  l e f t)  w i th  t he  ICRF  a n t e n na  ( r i g h t) 

oscillations, thereby  increasing  central  electron  temperatures.  It  is this 

system  for  controlling  the  plasma  current  profile  which  is considered  to 

be the  main  tool  to  stabilise  high  beta  poloidal  plasmas with  a large 

bootstrap  current  (the so-called  advanced tokamak  scenarios). 

With  a prototype  launcher,  up to  2.3MW  of  lower  hybrid waves  have 

been coupled  to JET plasmas. The  Lower  Hybrid  generator  consists of  24 

klystron  amplifiers. The total  generator  power  is  15MW  in pulses up  to 

10s and  12MW  up to  20s. The power  is transmitted  from  the  generator 

through  waveguides  to  a new  upgraded  LHCD launcher,  which  is 

installed on the  main  horizontal  port  of  Octant  No:3. In the  launcher,  the 

power  is progressively  split  to  feed  a phased array of  384 waveguides  at 

the  grill  mouth  facing  the  plasma. The front  end of  the  launcher  as seen 

from  inside the vessel is shown  in Fig.16. Three vertically  superposed 

areas of  the waveguide  array can  be phased  independently.  This allows 

for  high  flexibility  in the  wave spectrum  composition. 

The  control  system  for  the  generator  includes  a number  of 

automatic  software  routines  to  simplify  operation  of  the  plant.  One 

of  these  is  used  to  condition  the  launcher  vacuum  waveguides, 

without  plasma  load.  This  routine  pulses  the  generator  modules  at 

the  maximum  rate  allowed  and  automatically  adjusts  the  power  to 

the  maximum  that  the  launcher  can  handle.  A  second  software 

routine  optimises  the  power  within  an  LH pulse  into  plasma  to  the 

37 

T E C H N I C AL 

S T A T US  OF 

J ET 

Neutron Activation 
System 

Lidar Thomson 
Scattering 

X.U.V.and V.U.V. Broadband 
Spectroscopy 

Active Phase 
Spectroscopy -

V.U.V Spectroscopy" 
Spatial Scan 

X-Ray Pulse Height-
Spectrometer 

Surface Probe 
Fast Transfer System 

Far Infrared Interferometer  ' 
and Polarimeter 

Microwave  Reflectometer, 

Charge Exchange 
Recombination 
Spectroscopy 

\ 

X-Ray Crystal 
Spatial Scan 

- Plasma Boundary Probe 

Neutron Yield Profile 
Measuring System 

-Cassette 

2mm Microwave 
Interferometer 

Bolometer Array 

^ Soft X-Ray 
Diode Array 

Neutron Yield Profile 
Measuring System 

High Resolution X-Ray 
Crystal Spectrometer 

Time Resolved 
Neutron Yield Monitor 

Fig.17: General layout of diagnostics in the machine 

maximum  possible  by ramping  up the  power  until  there  is an arc  in 

the  launcher.  The  power  is then  removed  for  a pre-set  time  and 

reapplied  at  a level  20kW  below  that  which  caused the  arc. 

Full current  drive conditions  at  2MA  plasma  current  were 

achieved  in the  first  dedicated  LH experiments  in  mid-year,  and at 

3MA  with  about  6MW  LH power  coupled  towards  the  end of  the 

year.  The variation  of  LH power  deposition  profiles  was studied  with 

extensive  code  modelling  of  the  experimental  data  in  preparation  for 

profile  control  experiments.  Strong  broadening  of  the  current  profile 

was obtained  with  off-axis  LH current  drive  (LHCD).  In addition, 

sawteeth  were  stabilised  with  near-axis  LHCD.  Performance  can  still 

be  increased  up to,  the  maximum  design  launched  power  of  10MW. 

Diagnostics  Systems 

The status  of  JET's diagnostic  systems  at the  end  of  1994 is 

summarized  in Tables  4 and  5 and their  general  layout  in  the 

machine  is shown  in  Fig.1 7. The staged  introduction  of  the 

diagnostic  systems  onto  JET has proceeded  from  the  start  of 

T E C H N I C AL S T A T US OF J ET 

SYSTEM 

DIAGNOSTIC 

PURPOSE 

ASSOCIATION 

KB1 

Bolometer array 

Time and space resolved total radiated power 

KC1 

Magnetic diagnostics 

Plasma current, loop volts, plasma position, shape of flux surface, 
diamagnetic loop, fast MHD 

KE3 

Lidar Thomson scattering 

T„ and ne profiles at 4Hz and 8Hz for short periods 

KF1 

High energy neutral particle analyser 

Ion energy distribution up to 3.5MeV 

IPP Garching 

JET 

JET and Stuttgart 
University 

Purchased from loffe 
St Petersburg 

KG1 

Multichannel far infrared interferometer 

fneds on four vertical chords and four horizontal chords 

CEA Fontenay­aux­Roses 

KG3 

O­mode microwave reflectometer 

ne profiles and lluctuations 

KG4 

Polarimeter 

JncBpds on eight chords 

KH1 

Hard X­ray monitors 

Runaway electrons and disruptions 

KU? 

X­ray pulse height spectrometer 

Monitor of TL„ impurities. LH fast electrons 

KK1 

Electron cyclotron emission spatial scan 

Te(r,t) with scan time of a few milliseconds 

KK2 

Electron cyclotron emission fast system 

Te{r,t) on microsecond time scale 

K K 3 

Electron cyclotron emission heterodyne 

Te(r,t) with high spatial resolution 

K L I* 

Plasma viewing 

Monitor plasma in visible light 

K L :Ì 

Surface temperature 

Surface temperature of target tiles 

KM1 

2.4MeV neutron spectrometer 

KM3 

2.4MeV time­of­flight neutron spectrometer 

Neutron spectra in D­D discharges, ion temperatures and energy 

distributions 

JET and FOM Rijnhuizen 

JET and CEA 

Fontenay­aux­Roses 

JET 

JET 

NPL, UKAEA 

Culham and JET 

FOM Rijnhuizen 

JET 

JET 

JET 

UKAEA Harwell 

NFR Studsvik 

KM7 

Time­resolved neutron yield monitor 

Triton burnup studies 

JET and UKAEA Harwell 

KN1 

Time­resolved neutron yield monitor 

Time resolved neutron flux 

KN2 

Neutron activation 

Absolute fluxes of neutrons 

KN3 

Neutron yield profile measuring system 

Space and time resolved profile of neutron flux 

KN4 

Delayed neutron activation 

Absolute fluxes of neutrons 

KR2 

Active phase neutral particle analyser 

Ion distribution function, T,(r) 

KS1 

Active phase spectroscopy 

Impurity behaviour in active corditions 

KS3 

Η­alpha and visible light monitors 

Ionisation rate. Ze f fl impurity fluxes from wall and limiter 

KS4 

Charge exchange recombination spectroscopy 
[using heating beam) 

Fully ionized light impurity concentration, T.(r), rotation velocities 

KS5 

Active Balmer α spectroscopy 

TD, ND and Z^ir) 

UKAEA Harwell 

UKAEA Harwell 

UKAEA Harwell 

Mol 

ENEA Frascati 

IPP Garching 

JET 

JET 

JET 

KSS­

3ragg rotor X­ray spectrometer 

Monitor of low and medium Ζ impurity radiation 

UKAEA Culham 

KS?· 

Poloidal rotation 

Multichannel spectroscopic measurement of poloidal rotation 

UKAEA Culham 

KT2' 

VUV broadband spectroscopy 

Impurity survey 

UKAEA Culham 

KT3 

Active phase CX spectroscopy 

Full ionized light impurity conceitration, T,(r), rotation velocities 

JET 

KT4­

Grazing incidence+visible spectroscopy 

Impurity survey 

UKAEA Culham 

KX1 

High resolution X­ray crystal spectroscopy 

Central ion temperature, rotation and Ni concentration 

ENEA Frascati 

ΚΥ3· 

Plasma boundary probes 

Vertical probe drives for reciprocating Langmuir and surface 
collector probes 

JET, UKAEA Culham 

KZ3' 

Laser injected trace elements 

Particle transport. Tp. tmpunty behaviour 

Κγ1 

Gamma rays 

Fast ion distribution 

JET 

JET 

• Not compatiDle with Tritium 

Table 4: Status of JET Diagnostic Systems, December 1994 ­ Existing Diagnostics 

39 

T E C H N I C AL 

S T A T US 

OF 

J ET 

SYSTEM 

DIAGNOSTIC 

PURPOSE 

ASSOCIATION 

STATUS 

KB3D 

Bolometry  ol  divertor  region 

Power  balance  of  divertir  plasma 

KB4 

In-vessel  bolometer  array 

Time  and  space  resolved  radiated  power 

KC1D 

Magnetic  pickup  coils 

Plasma  geometry  in divertor  region 

KD1D 

Calorimetry  of  Mark  I divertor  targets 

Power  balance  of divertor  plasma 

KE4 

Fast  ion  and alpha-particle  diagnostic 

Space  and time  resolved  velocity  distribution 

KE9D 

Lidar  Thomson  scattering 

Te  and  ne profiles  in divertor  plasma 

KG6D 

Microwave  interferometer 

jnedl  along  many  chords  in divertor  plasma 

JET 

JET 

JET 

JET 

JET 

JET 

JET 

Operational. 

Partially  operational. 

Operational. 

Operational. 

In installation. 

Commissioning. 

Installed. 

KG7D 

Microwave  comb  reflectometer 

Peak  ne along  many  chords  in divertor  plasma 

JET 

In installation. 

KG8 

Ε-mode  reflectometer 

Measurement  of density  fluctuations  and  profiles 

in edge  and  SOL 

JET  and 
CFN/IST  Lisbon 

In installation. 

KJ3 

Compact  soft  X-ray  cameras 

MHD  instabilities,  plasma  shape 

JET 

Operational. 

KJ4 

Compact  soft  X-ray  camera 

Toroidal  mode  number  determination 

JET 

Operational. 

KK4D 

Electron  cyclotron  absorption 

neTe profile  along  many  chords  in divertor  plasma 

JET 

Commissioning. 

KM2 

14MeV  neutron  spectrometer 

KM5 

14MeV  time-of-flight  neutron  spectrometer 

Neutron  spectra  in  D-T  discharges, 

ion temperatures  and  energy  distributions 

UKAEA  Harwell 

In installation. 

NFR  Gothenberg 

In installation. 

KT1D 

VUV  spatial  scan of  divertor 

Time  and  space  resolved  impurity  densities 

JET 

Not  operational, vacuum  system  fault. 

KT5D 

KT6D 

Toroidal  view visible  spectroscopy  of  divertor 

Tz  and  Vz, ion  temperature  and  toroidal  velocity 

plasma  from Octant  No: 7  mid-plane 

of  impurities 

Waiting  for  installation. 

Poloidal view  visible  spectroscopy  of 

divertor  plasma  using a  periscope 

mpurity  influx,  2-D  emissivity  profile  of  lines 

JET 

Operational. 

KT7D 

VUV  and  XUV  spectroscopy  of  divertor  plasma 

Impurity  influx,  ionization  dynamics 

KY4D 

Langmuir  probes  in divertor  target  tiles 
and  limiters 

ne  and Te  in the divertor  plasma  and  limiters 

KY5D 

Fast  pressure  gauges 

Neutral flow  in divertor  region 

KY6 

50kV  lithium atom  beam 

Parameters  of  the  scrape-off-layer  plasma 

KY7D 

Thermal  helium  beams 

ne  and Te  in the divertor  plasma 

JET 

JET 

JET 

JET 

JET 

Operational. 

Operational. 

Operational. 

Commissioning. 

Installed. 

Ka1 

Thin-foil  charge  collectors 

Lost  alpha-particle  detection 

JET 

Waiting  for  installation. 

Table  5: Status  of  JET Diagnostic  Systems,  December  1994  -  New  Diagnostics 

operation  in  June  1983.  The  present  status  is that  48  systems  are  in 

existence.  35  of  these  systems  were  diagnostics  which  had  existed 

from  the  1992  campaign  but  had  been  modified  or  upgraded  for 

operation  in  the  new  phase  of  JET or  in  the  active  D-T  phase.  Table  5 

sets  out  the  list  of  23  additional  diagnostics,  which  were  specifically 

prepared  for  divertor  operation,  and  shows  their  present  status. 

Operational  experience  on  the  existing  diagnostics  has  been  good 

and  most  of  the  systems  have  operated  automatically  with  minimal 

manual  supervision.  The  resulting  measurements  have  been  of  high 

quality  in  terms  of  accuracy  and  reliability,  and  have  provided 

essential  information  on  plasma  behaviour  in  JET.  Further  details  on 

specific  diagnostics  systems  are  given  below. 

40 

T E C H N I C AL 

S T A T US  OF 

J ET 

Radiometer-

Polychromator-

Michelson-

Fig.18:  Poloidal  cross-section  of  JET showing  location  of  the  horizontal 

ECE sightlines  of  the  conventional  antennae,  and  the  near-horizontal 

sightline  of  the  quasi-optical  collection  system. The  sightlines 

employed  by  the  different  ECE measurement  instruments  are  marked 

Temperature  and  Density  Measurements 

Major  modifications  to  the  electron  cyclotron  emission  (ECE) system 

were  made  during  the  shutdown  to  adapt  the  diagnostics  to  the  new 

configuration  and to  further  enhance  the  measurement  capability.  In 

1994,  the  three  different  instruments  in the  ECE system  (Michelson 

interferometers,  the  grating  polychromator  and the  heterodyne 

radiometer)  again  provided  electron  temperature  data  on  most 

pulses. At  the  interface  with  the  machine,  new  antennae  to  adapt  to 

the  new  plasma  configuration  and  a quasi-optical  collection  system 

for  the  heterodyne  radiometer  were  installed.  All  systems  now  view 

the  plasma  through  double  vacuum  windows. 

A new antenna  array which  provides three horizontal  sightlines 

passing close to the  new  plasma centre (typically  ~0.3m  above the 

vessel micplane)  replaces the old fan-shaped array which  had sightlines 

41 

T E C H N I C AL 

S T A T US 

OF 

J ET 

Short pulse ruby lasers 

Vacuum beamline 

Digitizers 

Polychromatoi 

Safety shutter 

Camac and detection 
electronics 

Scattered light 

Collection mirrors 

Main & divertor 
LIDAR light paths 

Fig.19:  Cutaway  section  of  the  LIDAR  Thomson  Scattering  System 

spread  across  the  whole  poloidal  cross-section.  There  is  also  one 

obliquely  viewing  antenna  directed  towards  the  plasma  centre.  The 

original  array  of  four  horizontally  viewing  antennae  has  been 

retained.  To  realise  the  high  resolution  capability  of  the  heterodyne 

radiometer,  a  Gaussian  beam  (or  quasi-optical)  collection  system, 

employing  t wo  mirrors  inside  the  vacuum  vessel  and  scalar  feed 

horns  outside,  was  also  installed.  This  give  a  small  antenna  spot  size 

in  the  plasma  (~0.08m  diameter)  which  minimises  the  effect  of  flux 

surface  averaging  of  temperature  measurements.  Figure  18  shows 

the  locations  of  the  most  important  sightlines  now  installed. 

Installation  and  testing  of  all  the  new  components  for  the  main 

LIDAR  system  were  completed  on  time  ready  for  integrated 

commissioning  (see  Fig. 19). The first  successful  LIDAR  Thomson 

scattering  profile  measurements  at  the  new  standard  rate  of  4Hz 

throughout  a  pulse  were  obtained  in  March.  Figure  20  shows  an 

42 

T E C H N I C AL 

S T A T US 

OF 

J ET 

Pulse  No: 33643 

3 
Major  radius  (m) 

Pulse  No: 33643 

-20 

3 
Major  radius  (m) 

Fig.20: Temperature  and density  profiles obtained  by the  LIDAR  System 

example  of  LIDAR  electron  temperature  and  density  profiles  during  a 

recent  pulse. 

During  the  year,  the  diagnostic  has  also  demonstrated  an  8Hz 

repetition  rate  capability  of  the  system  for  a  7s  burst  during  a  pulse.  To 

do  this,  the  layout  of  the  laser  beam  paths  were  reorganized  in  the  Roof 

Laboratory.  By using  a dielectric  polarizer  as a  beam  combiner  and a 

pulsed  magneto-optic  polarization  rotator  to  correct  the  polarization, 

the  beams  of  both  the  available  4Hz  lasers  were  combined  along  the 

main  LIDAR  path.  Better  matching  of  the  pulse  energy  between  the  t wo 

lasers  is required  to  make  this  scheme  a  regular  option. 

A  new  LIDAR  Thomson  Scattering  system  was  installed  during  the 

shutdown  for  density  and  temperature  measurements  in  the  divertor 

region.  The  system  employs  in-vessel  mirrors  to  deflect  the  laser  beam 

and  the  collected  light  from  the  divertor  region  out  through  the  main 

horizontal  port  of  Octant  No.5.  The  silver  coated  mirrors  failed  during 

43 

T E C H N I C AL 

S T A T US 

OF 

J ET 

the final thermal cycling tests and new aluminium  coatings  had to be 

applied to the substrates. The reduced reflectivity  of aluminium  relative 

to silver combined with the fitting  of  double windows  on the vacuum 

vessel unfortunately  meant  halving of  expected signal. 

Alignment  of  the  system  has proved  a problem. The  in-vessel 

mirror  system  does  not  allow  a direct  view  of  :he  divertor  structure 

outside  the  very  narrow  field  of  view  of  the  actual  collection area. 

When  the  laser  strikes the  divertor  target,  not  only  a large  stray  light 

pulse  but  an  intense  white  light  signal  is observed. The  combination 

of  poor  transmission, alignment  difficulties  and  "wall  pulse"  has 

prevented  plasma  measurements,  so far. 

Three  microwave  diagnostics  share a single  set of  waveguides 

and  antennae  for  divertor  plasma  measurements.  These are a dual 

frequency  interferometer  for  line-integrated  electron  density 

measurements,  a  "comb"  reflectometer  for  estimating  the  peak 

density  along  a sightline  and an electron  cyclotron  absorption 

diagnostic  for  the  determination  of  the  local  electron  temperature-

density  product  (the electron  pressure). 

An  Electron  Cyclotron  Absorption  (ECA) diagnostic  is being 

developed,  as it  is not  possible to  determine  the  electron 

temperature  from  electron  cyclotron  emission  measurements  in the 

optically  thin  divertor  plasma. The  plasma  is not  sufficiently  opaque 

to  re-absorb the  intense  ECE from the core plasma, which  radiates at 

the same frequencies.  By measuring the absorpt on of  radiation  from an 

external source, the optical  depth will  be determined directly.  From this 

measurement, the electron  density-temperature  product (the electron 

pressure) will  be deduced. The diagnostic will  measure the  attenuation 

by the divertor  plasma of  radiation from  swept frequency  microwave 

sources.  Commissioning the system has been delayed  by problems  with 

the supply of the  Backward Wave Oscillator  source, which  is a key 

component  of the system.  It is expected that the diagnostic will  be able 

to  measure the spatial  profile of the neTe product  in the range 2.5x1020 

to  5x102lm~3eV at  most values of toroidal  field, with  an accuracy of 

±20%,  on a timescale of ~1ms. 

Boundary  Measurements 

Langmuir  probes are one of  the oldest  plasma diagnostics and still play 

a crucial  role in diagnosing the  plasma  boundary.  In JET, there are 39 

44 

T E C H N I C AL S T A T US OF J ET 

104 

cñ~~ 
E 
> 
cu 
CO 
o 103 

ω 
3 
CO 
ω φ 
1_ 
CL 

S 102 
o 
o 
ω 
LU 

\ 

\ 

: \ 

\ 

\ 

\J 

\ V |A 

Target 

ƒ 

j 

°  o o ^ i ïL  Upstream 

t f f i %î 

^ •^ 
^o 

V e s s=  1-19cm 

"SD 
\ 

101 

I 
-2 

) 
C 

I 
2 

I 
4 

vlidplane distance (cm) 

σι 
ρ 

3 

\ 

Ni 
6 

Fig.21: Comparison  of  electron  pressure  profile  from  the  reciprocating 

probe  at  the  machine  top  with  electron  pressure  profile  measured  with  a 

triple  probe  in the  outer  divertor  leg, during  a sweep  of  the  strike  point 

triple and 32 single Langmuir  Probes in the divertor  and a further 22 

single  Langmuir  probes in the  limiters. This system is fully  operational 

and due to  its self testing and configuring, it  is highly autonomous. The 

triple  probe system can measure electron  density and temperature  at up 

to  5kHz, which  is essential for the study of  ELMs and discharges where 

the strike zones are swept across the  divertor. 

In the scrape-off  layer (SOL) at the top  of the  machine, two 

reciprocating  Langmuir  probe systems provide measurements of the 

electron  density, electron temperature, electron  pressure, plasma flow 

velocity end floating  potential. This data  provides the best indication  of 

separatrix  parameters, which are crucial boundary  conditions for 

modelling the SOL plasma.  Figure 21  is an example of the plasma 

pressure  recorded with  the  reciprocating  probe.  Also shown  is the 

electron  pressure profile  in the divertor  from a triple  Langmuir  probe, 

which  had the outer  strike zone swept  across it. This has been  mapped 

around the magnetic  surfaces to the same coordinate  system as the 

reciprocating  probe.  One of the  reciprocating  probe systems has an 

exchange chamber, which  allows collector  probes to  be mounted. 

These have been extensively  used to  monitor  beryllium  evaporations and 

metallisation  during glow  discharge cleaning. 

Since the  new plasma configurations  are further  away from  the 

original set of  magnetic  sensors, a new set of  magnetic  probes, saddle 

loops and full  flux  loops were  installed  inside the vessel to  maintain the 

45 

T E C H N I C AL 

S T A T US 

OF 

J ET 

Pulse  No:  32838 

_  lp =  4.8MA 

2.0 

1.0 

D) 
CD 
Χ 

-1.0 

-2.0 

1.0 

2.0 

3.0 

Radius (m) 

Fig.22: Real-time  display  of  the  plasma  boundary  of  a 5MA  discharge. 

Agreement  with  a full  calculation  by  EFITJ is indicated  by the  dotted  line 

boundary  reconstruction  accuracy. The new sensor configuration was 

designed using simulated  plasmas to assess the reconstruction  accuracy. 

Operation  has demonstrated  boundary  reconstruction  in the new 

configuration, accurate to within  a few centimetres  using the  EFIT and 

XLOC codes (see Fig.22). The main diagnostics  used in verifying the 

calculated  boundary  position are the divertor  Langmuir  probes, 

reciprocating  probe and plasma viewing cameras. 

The XLOC code is also used in the  real-time  plasma position and 

current  control system (PPCC) to control the current  in the coils, which 

shape and position JET plasmas. Reliable magne:ic  data  is essential for 

safe operation  of JET. Therefore, data quality  is monitored  by post-pulse 

validation software, which warns of  potential errors  in the  magnetics 

data. A minimal set of  analogue signals is also connected to the PPCC 

system for  safe termination  of  the pulse in the event of  loss of  digital 

communication  through which the magnetic  data  is received. 

Neutron  Measurements 

The most  important  diagnostic  for  the  assessment  of  the  fusion 

performance  of  the  tokamak  is the  time-resolved  neutron  yield 

monitor  comprising three pairs of fission chambers arrayed around the 

46 

T E C H N I C AL 

S T A T US  OF 

J ET 

(a)  Pulse  No: 26087 

14MeV 
neutrons 
(x50) 

12.0 

13.0 

14.0 

15.0 

Time  (s) 

Fig.23: Comparison  of  2.5MeV  neutron  emission  with  the  resulting 

14MeV  emission  from  triton  burn-up  in  D-D  discharges. The  2.5MeV 

neutron  time-traces  are  broadly  similar  during  their  ramp-up  phases, 

although  their  terminations  are  accompanied  by  different  phenomena, a 

carbon  bloom  for  Pulse  No:26087  and  ELMs for  Pulse  No:32919. The  effect 

of  carbon  influx  is noticeable  on the  14MeV  neutron  time-trace  for  Pulse 

No:26087. The  14MeV  neutron  signals  have  been  multiplied  by  50 

tokamak. This is a highly  reliable diagnostic  for  relative  measurements 

but  its absolute calibration  must  be determined  anew when changes are 

made to  hardware within  and surround the  machine. Experience of 

changes has been acquired over the years that  new calibrations can 

now  be predicted with  some confidence. For the  new  divertor 

configuration,  many changes were made. However, the divertor  itself 

was not  expected to affect  the fission chamber  calibrations. The 

predicted  2.5MeV  neutron calibration for  1994 operations  has been 

confirmed with  measurements of  neutron  activation  using the 

pneumatic  transport  system. Figure 23 shows a comparison  between 

recent  measurements and those of  the  previous campaign. 

47 

T E C H N I C AL 

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OF 

J ET 

The neutron  profile  monitor  provides detailed  information  on the 

fusion  performance  of JET plasmas that  cannot easily be obtained by 

other  means. The profile  monitor  is also furnished with  alternate 

detector  boxes, appropriate to the study of  bremsstrahlung  emission 

from fast electrons (FEB) accelerated  by the lower  hybrid current drive 

launcher. The FEB diagnostic  is regarded as essential for  LHCD studies. 

Due to this dual function  for the  profile  monitor,  the detector  boxes 

have to  be exchanged according to the  needs of the  programme. As a 

result, the  neutron camera was not always in place when  many 

interesting  discharges were  run. This upgraded  profile  monitor is 

designed to  provide simultaneous acquisition of  FEB and neutron data. 

Considerable effort  has been devoted to construction  and fitting-out 

of the  new profile  monitor.  It will  use separate detectors for  2.5MeV 

neutrons,  14MeV  neutrons and bremsstrahlung  radiation and will 

possess remotely  changeable collimation to  provide coarse optimization 

of  rate of  detected  events for  differing  plasma conditions. The existing 

profile  monitor will  be used until the new one is installed  in early-1995. 

Impurities 

Three spectrometers were used to  record the VUV, XUV and soft X-ray 

emission from the bulk  plasma during  1994 operations; a crystal 

spectrometer and two  grating  instruments. These spectral  regions are of 

particular value since they permit characteristic spectral lines to be 

monitored of all the intrinsic  impurities found in JET plasmas. The Bragg 

crystal spectrometer  consists of four  scanning monochrometers. A 

balance must be made between the temporal  resolution and the number 

of  lines accessed. Initially, a time resolution of  ~400ms was used and 

seven lines were routinely  monitored. With  increasing availability of 

spectroscopic data, a much improved temporal  resolution of  ~80ms was 

adopted with five lines being recorded. The SPRED spectrometer  is a 

VUV, grating  instrument  providing coverage of the wavelength  region 

from  100 to  1100Å with  a spectral  resolution of  -3Å  (see Fig.24). The 

XUV, SOXMOS instrument  is a grazing incidence spectrometer with  two 

detectors. Each views a wavelength  range of  -4CÅ and can be moved to 

record the spectrum at positions between  15 and 340Å. 

A new VUV/XUV diagnostic  has now  been installed to observe line 

radiation of  the main plasma  impurities emitted from the  divertor 

region. The instrument  consists of three spectrometers, a double 

48 

T E C H N I C AL 

S T A T US 

OF 

J ET 

Fig.24: Typical VUV spectrum from the SPRED instrument 

SPRED, which  combines  two  VUV spectrometers  in one  instrument 

and  a SOXMOS for  the  soft  X-ray  emission. All  systems started  to 

produce  spectra  on  a regular  basis at  the  end  of  1994, although  the 

SOXMOS only  produces  a weak  signal. The  reason for  this  is still 

under  investigation. 

Remote  Handling 

A comprehensive suite of  remote handling equipment  is being  prepared 

to support  operations during  periods when  man-access to the  machine 

is not  possible. During and immediately  after  periods of  D-T operation, 

parts of  the  machine will  become activated to an extent, where man-

access is severely limited and, under these conditions, all  maintenance 

and remedial work  in the restricted areas will  be performed  by traired 

operators  utilising  remote  handling  equipment. 

The  remote  maintenance  philosophy  is based on  the  principle 

that  maintenance  tasks  must  be  undertaken  by trained  operators 

using  remote  handling  machines. The operators  position  and  deploy 

the  equipment  so that  remote  maintenance  becomes  a direct 

extension  of  the  'hands-on'  maintenance  operations.  The  inbuilt 

intelligence  and  adaptability  of  the  human  operator  is retained  and is 

aided  and  enhanced  by the  robotic  devices, which  provide  force 

49 

T E C H N I C AL 

S T A T US 

OF 

J ET 

Fig.25: Computer generated  image of Mascot servo-manipulator 
attached to the Articulated  Boom performing  in-vessel tasks 

reflection from the work  face and include functions, which  enable the 

equipment  to  learn and repeat particular  motions and procedures. 

This approach  has enabled the overall  remote handling  equipment 

development  to  be minimally affected  by the continual  modification  and 

development  of JET since  1983. During  1992/3, a complete 

reconfiguration  and replacement  of  hardware took  place within  the 

torus. The only significant  effect  on the  remote handling system 

equipment was in the area of specialised tooling where development  of 

new cutting, welding  and handling tools was required. All other  major 

elements of the  remote handling  system remained unaffected. 

During  1994, the  Remote  Handling  Articulated  Boom and  special 

tooling  were  used to  facilitate  the  final  phase  of  Mark  I Divertor 

installation,  including  installation  of  the  Mark  I divertor  modules  and 

the  welding  of  the  divertor  module  cooling  pioework  in-vessel. 

In the  rest of  1994, all effort  within the  Remote Handling  Group was 

focused on  preparing for  the installation of the  Mark  IIA divertor  system 

during  1995 and the fully  remote exchange of  Mark  IIA tile carriers by 

the  Mark  IIB tile carriers planned immediately  after  DTE1. This remote 

tile exchange will  be the first fully  remote task  undertaken on the 

machine and will  also be the most significant fully  remote handling task 

50 

T E C H N I C AL 

S T A T US 

OF 

J ET 

Short  boom 

Door to man-access  cabin 

TCTF  enclosure 

Limb  brackets 

End-effector 

ISO container 

Base tile  carriers 

Fig.26:  Octant  No:1  Facility  designed  to  transport  tile  carriers  and tools  fully  remotely  into  and from  the  torus 

undertaken  on any fusion  device in the world. Accordingly,  extensive 

preparations are being made for  the proving of  the equipment  function, 

performance  and reliability together  with  the derivation and 

development  of task techniques and operator  training  using full scale 

mock-ups. The operations will  be performed  in-vessel by use of the 

Mascot servo-manipulator  mounted on the end of  the Articulated  Boom 

entering from  Octant  No:5 (see Fig.25). Special tools will  be used to 

handle the tile carriers and all components will  be transferred  into and 

out  of the torus through  Octant  No:1  main horizontal  port. 

The  remote  tile  exchange  concept  requires  the  transfer  of  tile 

carriers  and  tools  into  and from  the  torus  through  the  Octant  No:1 

main  horizontal  port.  This transfer  task  and  the  placing  of  tile  carriers 

in  a transport  container  will  be carried  out  fully  remotely  by  means  of 

a new Tile  Carrier  Transfer  Facility  which  will  be  installed  and 

operated  at  Octant  No:1  (see Fig.26). The facility  comprises  a so-

called short  boom  with  an  end  effector  suitable  for  carrying  the 

various  tile  carriers  and  tools  housed within  a sealed  enclosure  for 

contamination  control  but  with  interface  doors for  a man-access 

module  and  tile  carrier  transfer  containers.  Concept  design  for  the 

5  1 

T E C H N I C AL 

S T A T US 

OF 

J ET 

Fig.27: Picture mosaic of the inside of the JET vessel 

Control  and  Data 
Acquisition 
Due to  the high  number  of  components 

and their distribution  throughout  a large 

whole  facility  has been  completed  and  a contract  for  detail  design 

site,  the  operation  and  commissioning 

and  manufacture  of  the  short  boom  has been placed. 

of  JET is  supported  by  a  centralised 

Control  and  Data  Acquisition  System 

The new In-Vessel Inspection  System (MS), installed during the 

(CODAS).  This  system  is  based  on  a 

major shutdown, has been in operation since January  1994. It consists 

network  of  Norsk  Data  minicomputers 

interfaced 

to  the  experiment 

through 

of four  newly designed viewing  probes and add tional  lighting  units and 

CAMACinstrumentation 

(including  front 

its upgraded  enhancement.  As a result, clearer  pictures and a wider 

end  micro-processors) 

and 

signal 

conditioning  modules.  The  various 

range of view,  including the direct view of the divertor  have been 

components  have been logically  grouped 

obtained. Consequently, the system has become an increasingly 

into subsystems with  each one  controlled 

and  monitored  by a computer.  After  a 

important  instrument  in machine operation, allowing early detection  of 

pulse,  all  the  information 

from 

the 

any damage occurring  during  plasma operations and for  calibrating 

subsystem  is  merged 

together 

into  a 

single  file  on  the  storage  and  analysis 

several in-vessel diagnostics.  MS campaigns, during which the entire 

computer.  This file is then transmitted  to 

vessel is scanned, are presently carried out on a weekly  basis. Results are 

the IBM mainframe  computerfordetailed 

analysis. A summary of information 

from 

available in the form  of  picture  mosaics (Fig.27), from which  individual 

the  JET pulses  is held  in  the  JET Survey 

pictures can be easily chosen and enlarged for  display on the  monitor  of 

Data  Bank. 

any workstation  connected to the internal  network. 

In  1994, an  In-Vessel Training  Facility  was  built,  which  is a 

realistic  full-scale  replica  of  Four  Octants  of  the  inside  of  the JET 

vessel and  all  of  the  important  aspects of  the  in-vessel  components 

(Fig.28). This facility  has three  main  uses: 

•  a major concern for  future  shutdowns  has been the provision of 

adequate  in-vessel training. Due to  beryllium  contamination  of  the 

vessel and to  remain within  JET's self  imposed radiation exposure 

limit  of  10% of the  legal limit,  a large number  of workers are 

required. The facility  is used to ensure that  in-vessel work  can be 

done safely, efficiently  and with  a minimum  of  radiation exposure; 

52 

T E C H N I C AL 

S T A T US 

OF 

J ET 

Fig.28:  T he  In-vessel  T r a i n i ng  Facility 

• 

the  facility  is used to  check  the  details  of  new  components, 

develop  and test  all  new  installation  procedures,  and  interface 

problems  to  ensure that  the  Mark  II divertor  can  be installed  to 

the  desired accuracy  and  with  the  prevailing  boundary  conditions; 

• 

the  facility  will  be  used for  remote  handling  operations  trials. The 

preparation  of  equipment  and  operators  for  the  fully  remote 

exchange  of  the  tile  carriers  will  include  the  full-scale  testing  and 

proving  of  each  individual  task  both  under  normal  operating 

conditions  and  under  failure  case conditions.  The  forthcoming 

mock-up  task  trials will  include  extensive  testing  and  proving  of 

the  capability  to  recover  from  worst  case remote  handling 

equipment  failures  and  also from  worst  case failure  of  in-vessel 

components  such as seizure  of  bolts  or fractured  tiles. 

Waste  Management 

The Waste Management  Group  is responsible for  provision of  facilities 

in support  of  interventions  and shutdowns,  respiratory  protection 

equipment  and disposal of  radioactive and beryllium wastes. This 

involves operation  of  controlled  areas, including the Torus Access Cabin 

(TAC), the two  Beryllium  Handling facilities and Waste Handling. 

53 

T E C H N I C AL 

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OF 

J ET 

Fig.29: Torus  Access  Cabin 

The Torus Access Cabin (TAC) enables personnel to access the JET 

vessel in a manner which  prevents the spread of  radioactive of  beryllium 

contamination  beyond controlled areas and includes decontamination, 

workshop and waste transfer facilities. The latter  includes the ability to 

transfer  both drummed waste and waste  in ISO containers. These can 

be docked onto the TAC  by means of  a special double  PVC membrane. 

It is designed  in such a way as to  present a clean, closed face on  both 

controlled area openings to the Torus Hall environment. The containers, 

presenting a clean outside surface, can then  be transferred  through 

uncontrolled  areas and docked onto the J30 Waste  Handling  Facility. 

The arrangements, which are shown  in Fig.29, have reduced the 

transfer  time from one shift to 45  minutes. 

Although  the  TAC  attached  to  the  vacuum  vessel was  a semi 

permanent  arrangement  operating  continuously  for  22  months  of 

54 

T E C H N I C AL 

S T A T US  OF 

J ET 

the  divertor  shutdown,  improvements  have  been  made to  permit 

rapid  docking  to  the  machine  to  permit  access for  emergency 

interventions.  The TAC  is normally  parked  in the  Assembly  Hall and is 

craned  into  the  Torus  Hall when  required. Three  interventions  were 

made  during  1994 and the  in-line  time  involved  in connection  of  the 

TAC  (including  opening  of  the  pumping  chamber  door, 

establishment  of  services and  preparation  for  entry  in  pressurised 

suits)  has been  reduced  to  less than  one  shift. 

The  Beryllium  Handling  Facilities  have  been  operated  continuously 

Prediction,  Interpretation 
and  Analysis 

during  1994. These provide  containment  and contamination  facilities 

The  prediction 

of  performance 

by 

to  enable  the  carrying  out  of  cleaning  and  maintenance  on  beryllium-

contaminated  components. 

Control  and  Data  Management 

computer  simulation,  the 

interpretation 

of  data,  and  the  application  of  analytic 

plasma  theory  are of  major 

importance 

in  gaining  an  understanding  of  plasma 

behaviour  in JET. 

•  Prediction  work  continuously  checks 

The JET Control  and Data Acquisition  System (CODAS) is based on a 

the measured  behaviour  against  the 

network  of  minicomputers.  It is the only  interface to operate JET and it 

different  computational  models,  and 

provides  a  basis 

for 

long 

term 

provides centralised  control, monitoring  and data acquisition. The 

programme  planning; 

various components  of JET are logically  grouped  into subsystems like 

Vacuum, Toroidal  Field, Lower  Hybrid additional  heating, etc. Each 

subsystem is controlled  and monitored  by one dedicated  computer 

interfaced to the  machine and its diagnostics through  distributed  front-

end instrumentation.  Signal conditioning  and some data conversions are 

• 

Interpretation  plays  a key role  in  the 

assessment  of plasma  performance, 

and  hence  in  optimisation 

studies 

and  programme  planning; 

•  A  major  role  of  analytic  theory  is  to 

compare 

the  observed  behaviour 

against  that  expected  from  existing 

analysis,  and  to  modify 

the 

latter 

made using  EUROCARD modules. The rest of the  instrumentation is 

when  there  is  divergence. 

based on CAMAC  and VME standards.  Embedded front-end  intelligence 

is implemented through  microprocessors that are also used for  real-time 

applications. The actions of  the various computers are coordinated by 

superviso'y software  running  in the  Machine Console computer. This 

supervisory function  includes the countdown  sequence for  the  control 

and data acquisition  during each plasma discharge. 

During  1994, the  main  efforts  were  focused  on  completion  of  the 

A  central 

task  is  to  provide  a 

quantitative  model  of tokamak  plasmas 

with  the ultimate  objective  of  including 

all  the  important  effects  observed  in 

JET and other  tokamaks.  It is  preferable 

to understand  each effect 

theoretically, 

but  in some  cases it  may  be  necessary 

to  rely  on  an  empirical  description. 

For  carrying  out  these  tasks  it  is 

important 

that  JET data  is  held  in  a 

readily  accessible  and  understandable 

move  of  CODAS to  a UNIX  environment,  the  support  of  the  restart  of 

form. 

operation  in the  new  configuration  and the  implementation  of a 

significant  number  of  VME-based  front-end  applications. This 

required  full  integration  of  the  new  control  systems  into  the 

operation  and  deployment  of  new  front-end  intelligence  based  on 

the  VME standard  and  connected  by a dedicated  network. 

The mainframe computing  service is based on an IBM 3090/300J 

with  160 GigaBytes (GB) of  disc storage and a further  2000 GB of 

55 

T E C H N I C AL 

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OF 

J ET 

automated  cartridge tape storage. The service has operated since June 

1987 and the central computer was upgraded  in February  1990 from an 

IBM 3090/200E to an IBM 3090/300J with  three processors, two  vector 

facilities and 384  MegaBytes (MB) of  memory (128MB central and 

256MB  expanded), almost doubling the processing  capacity. 

In February  1992, a Memorex-Telex automated cartridge tape library 

(ATL) with a capacity of  1000GB was installed. The ATL not only provides 

storage for all raw JET data (JPFs) and archived processed JET data (PPFs), 

but also provides storage for  backup and dump tapes, previously  handled 

using manually operated cartridge tape drives. This together with the 

introduction  of automated operations via the product AutoMate/MVS has 

completely eliminated the requirement for operator  cover. In December 

1993, the eight tape drives of the ATL were upgraded doubling the data 

recording  density, and this doubled the capacity of the ATL. This disc 

system hardware was enhanced in two  stages during  1994, with the 

addition of a further  53GB of  a more recent generation of  IBM discs. 

The JET IBM Computer  Centre was originally  established at the 

UKAEA Harwell  Laboratory  but was relocated to  Building J2 at JET in 

July  1992. This service has run successfully from the JET location, 

providing the expected improvement  in communications,  integration 

with  the UNIX and PC systems, and a reduction  in staff.  The  integration 

was further  improved  early in  1994 with  the  introduction.of  TCP/IP 

services on the  mainframe  using the  Interlink  SNS/TCP access software 

to  provide more accessible means of  data transfer  between the 

mainframe and the  UNIX and PC or  Macintosh  systems. It also improved 

the terminal  emulation access from the  PCs and UNIX system. 

Summary  of  Machine  Operations 

During  1994, operations  were  made  up of  two  distinct  periods. 

(a)  First  Period  (Weeks  W4  to  W17) 

The Divertor  Shutdown  ended in January and the  machine 

integrated  commissioning  phase started. Commissioning  activities were 

undertaken  by a "Restart"  Task Force (Task Force R), whose main 

objectives were the  progressive and complete  integration  of  systems for: 

•  producing  the  first  plasma; 

•  establishing  a 2MA  divertor  plasma  with  up to  10 MW  of  heating. 

The first  plasma  pulse was followed  by a period  of  initial 

operation  for  plasma  commissioning  with  the  Plasma  Position  and 

56 

T E C H N I C AL 

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Current  Control  system  (PPCC) using the  magnetic  signals from  the 

magnetic  diagnostic  (KC1).  Development  and commissioning  of a 

new  improved  magnetic  diagnostic  (KO D) was carried  out  in  parallel 

and  progressively  integrated. The Task  Force  R  work  continued  until 

the  end of  April  1994 and obtained  reliable  3MA  plasma  discharges 

with  additional  heating.  65%  of  pulses were  successful.  Of these, 

60%  were  commissioning  pulses  and 40%  were  plasma  pulses. 

(b) Second  Period  (Weeks  W18  to  W51) 

Operation  time within  the  experimental  programme  was shared 

between  the  three  main Task  Forces:  Divertor  Assessment  (D);  High 

Performance  (H);  and Tokamak  Concept  Improvements  (T).  Some 

time  was  also devoted  to  plasma  commissioning  (C) and ICRF 

commissioning  (within  Task  Force RF). 

The experimental  programme  was  carried  out  in  double-shift 

operation,  six days a week.  The  number  of  shifts were  distributed as 

follows: 

TASK  FORCE 

NO:  OF  SHIFTS 

% OF  TOTAL 

TASK  FORCE  D 

TASK  FORCE  H 

TASK  FORCE  Τ 

TASK  FORCE  C 

TASK  FORCE  RF 

104 

84 

76 

48 

10 

32.30% 

26.10% 

23.60% 

14.90% 

3.10% 

The overall  ratio  of  successful  pulses  either  for  commissioning  or 

plasma  increased  to  77%  (compared  with  65%  during  the  period 

devoted  to Task  Force Restart). 

Interruptions 

to  Operation 

The time  available  for  operation  was  interrupted  by three  in-vessel 

interventions: 

• 

the first  interruption  occurred during Weeks W12 and W13.  Some 

large nickel  influxes, sparks and sometimes a long afterglow  were 

visible in most of the divertor  or  limiter  plasmas.  Detailed in-vessel 

inspection and completion  of  some diagnostic  systems were possible 

over this non-operational  period of  seven days; 

• 

the next  intervention, during Weeks W38 and W39, was spread over 

nine operational  days.  It was then  possible to  carry  out  repair  work 

57 

T E C H N I C AL 

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35 

30 

25 

öS  20 

s  15 

E 

10 

Total  pulses 
33223 

Tokamak  pulses 
24658 

Commissioning 
pulses 
8565 

83 

1984 

1985 

1986 

1987 

1988 

1989 

1990 

1991 

1992 

1993 

1994 

_J  Shutdown 

Fig.30: Cumulative totals  of JET Pulses: 1983-1994 

on  the  upper  saddle coils, following  some severe distortions  on 

the  cross-over  busbar  sections.  At  the  same time  several  other 

activities were  successfully  completed; 

• 

the  third  interruption  occurred  during  Weeks W44  and W45, 

when  nine  days were  used for  replacement  inside the  vessel of 

the three  beryllium  evaporator  heads  located  at  Octants  Nos.1, 5 

and  7.  These units were fitted  with  new  extension  pieces  whilst 

the  evaporator  at  Octant  No.3  was removed. 

Plasma  Current  Distribution  for  1994 

2-3MA 
58% 

3-4MA 
12% 

4-5MA 

0-1 MA 

1-2MA 

Fig.31:  Plasma  Current  Distribution  for  1994 

58 

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Overall Summary 

The total  number  of  pulses  carried  out  in  1994 was 4869  with  an 

overall  distribution  over  1983-1994  as shown  in  Fig.30.  An  analysis 

of  the  distribution  of  the  plasma  currents  showed  that  currents  of 

2-3MA  were  routinely  used during  1994. These  represented  82%  of 

the  tota!  plasma  pulses (Fig.31).  A total  of  43  pulses  in the  range 

4-5MA  were  also obtained. 

Technical  Developments  for 
Future  Operations 

Considerable  effort  was  devoted  during  the  year to  the  design  and 

procurement  of  equipment  for  installation  on the  machine  during 

future  shutdown  periods.  Reference  should  be  made to  the  section 

on  the  Future  Programme  of  JET to  relate these  technical 

developments  to  the  overall  JET Programme. 

Pellet  Injection 

The  injection  of  solid  hydrogen  pellets  is one  method  of  providirg c 

particle  source  inside the  recycling  boundary  layer  of  a future  fusion 

reactor  without  simultaneously  depositing  excessive  power.  The 

ablation  of  the  pellet  by hot  plasma  electrons  requires  very  high 

speed  pellets,  in order  to  penetrate  beyond  the  q=1  surface  to  the 

plasma  centre.  Work  is underway  on two  systems:  a high-speed 

pellet  launcher  and a pellet  centrifuge  for  shallow  deposition. 

The  high-speed  launcher  is designed  to  accelerate  sabot 

supported  6mm  deuterium  pellets  to  speeds  up to  4kms"1 by  using 

two-stage  light  gas gun  driver  technology  for  deep  central  plasma 

fuelling  purposes.  Sabots are small  plastic  pistons,  set  behind  the 

deuterium  pellets  to  protect  them  from  the  hot  driving  gas.  These 

sabots,  needed for  pellet  velocities  in excess of  about  3kms"'  anc 

being  accelerated  with  the  pellet,  are  made from  two  halves of  equal 

mass, which  split  after  leaving  the  barrel  due to  aerodynamic  forces 

and  can  be eliminated  from  the  pellet  path  by a shear  cone. 

This system  was  installed  on  the  machine  and  commissioned  with 

plasma  towards  the  end  of  1994.  First  pellet  launches  into  the 

plasma  have  been  performed. The two-stage  gun  performed  with 

few  problems  and  no further  modifications  were  required. Some 

limited  experiments  have  been  performed  with  pellet  injection  using 

59 

T E C H N I C AL 

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Fig.32: The  Pellet  Centrifuge 

polypropylene  (impurity  injection  experiment)  and  deuterium. The 

high-speed  pellet  activity  was terminated  at the  end  of  1994 and the 

equipment  was  mothballed. 

The pellet centrifuge  is to  provide a source of  deuterium  particles at 

varying depths beyond the  recycling layer and with  it a minimum 

recycling flow  into the divertor.  The injection  parameters chosen are 

pellet sizes 1.5-3mm with  repetition frequencies  up to 40s ' at speeds of 

50-60ms"' for  long pulses approaching  one minute.  Fig.32 shows the 

pellet centrifuge  unit on  its stand  in the Assembly  Hall.  It accelerates 

pellets mechanically from the hub to the tip of  a rotor  arm, where these 

leave at about  1.4 times the rotor tip speed. Each size of  deuterium  ice 

pellet will  be launched from one of  up to four  possible  individual 

extruder  units into the central  part of the centrifuge  rotor  hub, into 

which  reaches the stationary stop cylinder featuring  a hole to ensure the 

proper starting conditions for  the pellet on the  rotor  arm.  The design of 

centrifuge  rotor  and stop cylinder follows very closely that of  the 

centrifuge developed for ASDEX Upgrade  by IP3, Garching, Germany, 

who also advise JET under contract.  The extruder  is of  a new design by 

JET to  provide a much  larger  number  of  pellets per tokamak  pulse. 

60 

T E C H N I C AL 

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The first  extruder  for  3mm  pellets  has now  been  assembled, and 

it  has undergone  warm  pre-testing  of  its  piston  action.  The extruder 

will  be  installed  in the  main vessel  unit  in early  1995.  It will  then  be 

commissioned  and  made operational  from  initial  cool-down  to  tee 

extrusion  and  pellet  formation  phases, with  the  aim  of  preliminary 

application  of  pellet  injection  into  the  torus. 

Design  of  Mark  II  Divertor 

The Mark  I divertor  currently  installed  in JET has been successful from 

several points of  view.  In particular,  its power  exhaust capability  in the 

sweeping  mode  has been very good, and the cryopump works 

effectively.  The fact that  Mark  I is relatively wide  has made it possible to 

investigate a large variety of  magnetic  equilibria with  strike points on 

both the horizontal and vertical sections of the target tiles. 

However, the openness of the divertor  also has drawbacks.  Such an 

open divertor  is not as effective  at restraining  neutrals in the  divertor 

region as a more closed design, particularly at high powers and low  to 

moderate  main chamber  densities.  Under these conditions, the  divertor 

plasma itself  is not opaque enough to  re-ionize the recycling  neutrals. 

Thus, if was decided to design and construct a series of  more closed 

divertors,  beginning with  Mark  HA, to  be installed  in  1995, and followed 

by Mark  IIGB (Gas-Box) in  1997.  The purpose of these closed divertors 

is to  permit  a high  neutral density  in the divertor  volume while 

maintaining  a low  neutral density in the main chamber.  The first 

condition  is required to exhaust  power to a large fraction of the total 

divertor  wall area by the volumetric  processes of  radiation and charge 

exchange (i.e. to operate  in the  highly radiating, partially  or fully 

detached regime).  The second requirement, that  of  low  neutral density 

in the main chamber,  is necessary for  highest main plasma  performance. 

Mark  IIA is a moderately  closed semi-V divertor  which flares out  at the 

upper  region of  the side plates in order to accommodate  a wide variety 

of  equilibria.  These include several low X-point  equilibria  of  varying 

degrees of flux expansion, which  have strike zones on the  "outwa'd 

facing"  dome targets, as well as higher X-point  equilibria  which  contact 

the  "inward  facing"  vertical targets.  The target tiles are 40 cm long in 

the poloidal  direction, providing  a high toroidal wetted  length, relative 

to  Mark  I.  This feature, along with  the inclination  of  the  targets,  gives 

a wetted  area  sufficiently  large  that  the  need for  sweeping  is 

6  1 

T E C H N I C AL 

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OF 

J ET 

Fig.33:  Mark  II Gas-Box  divertor:  (a)  w i th  lower  side-wall  tiles,  and  (b)  w i t h o ut  lower  side-wall  tiles 

eliminated.  Simulations  indicate that the  pumping  performance  should 

be somewhat  better than that of  Mark  I.  Mark  HA is intended to satisfy 

the dual and sometimes conflicting  needs of  performing  well in both 

the  high performance  campaigns (low to  moderate density,  magnetic 

axis not too far above the midplane) and ITER-related high density 

radiating divertor  studies. The design of  Mark  HA was finalised  in  1992 

and procurement  and fabrication  went  ahead throughout  1993/94.  It is 

scheduled to  be installed starting  in June 1995. 

The term  "Gas-Box"  is used to denote a divertor  characterized  by a 

close fitting  baffle  near the X-point,  below which there  is a relatively 

large volume in which the  recycling  neutrals (both  hydrogenic  and any 

seeded recycling  impurity which  may be used) can circulate freely in 

order to-re-enter  the divertor  plasma along its entire surface.  This 

promotes the volume loss of  energy and momentum  required to  prevent 

excessive energy deposition on the target  "strike zones".  The JET Gas-

Box divertor  uses the same coils, pump, and support  structure as the 

Mark  HA divertor,  but  has new target tiles and tile carriers to  change the 

geometry from  moderate-V to  Gas-Box configuration.  The system has 

been designed so that the changeover  can be carried out via  remote 

handling, following  the D-T campaign scheduled for  late  1996. 

To provide  a specific  test  of  the  Gas-Box  concept,  the  range of 

equilibria  must  be restricted  to  allow  for  a baffle  which  is long 

enough,  and close fitting  enough, to  prevent  the  back-flow  of 

neutrals  into  the  main  chamber.  In practice,  this  means  that  only 

high  X-point  equilibria  can  be investigated. 

62 

T E C H N I C AL 

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Two versions of the Gas-Box are being designed and are shown in 

Figs.33 (a) and (b). These use a common  baffle, base target, and septum 

assembly.  The  "vertical target"  version shown  in Fig.33(a)  has a nearly 

vertical tile  placed just  below the baffle on both the inside and outside 

legs, and operates with  the strike  points on these vertical tiles.  In this 

case, the  recycling of  neutrals takes place through the private flux 

region only.  Such vertical target  divertors  have certain advantages, and 

current  ITER thinking  favours such a design.  Among the advantages are 

a higher wetted surface for  handling  power during  non-detached 

phases of  the discharge, such as start-up or  in ELMs.  In the second Gas-

Box configuration  (Fig.33(b)), the vertical target tiles are removed and 

the strike points are placed on the bottom  plates.  The wetted surface is 

now smaller,  but the pumping  is somewhat  stronger. 

In contrast to the design of  Mark  IIA, the tile carriers for  Mark IIGB 

will  probably  be fabricated from  CFC plates.  The anticipated  radiative 

loads on the carriers, including the septum, are sufficiently  high that 

bare inconel could not  be used, and would  need to  be clad.  Thus, the 

adoption  of  CFC as the structural  member  should lead to  a simpler 

engineering  solution, and tests are in progress to validate this approach. 

The target  plate and the  baffle  material  is CFC.  In principle,  beryllium 

could be used for  a later version, if justified  by continuing  interest  from 

ITER and favourable  results from the  Mark  I beryllium tile test  in  1995. 

At  present,  it  is planned  to  test  both  versions  of  the  Gas-Box 

divertor.  However,  the  scheduling  is very  tight,  and  it  may be 

necessary  to  omit  one  of  them.  The verlical  taiget  version  is the  one 

current y favoured  by JET (as well  as by ITER), but  the  design is 

sufficiently  flexible  to  permit  either  version  to  be operated  first. 

Tritium  Handling 

The purpose of the Active Gas Handling  System is to  pump the torus, to 

collect gases from various systems (torus, neutral beam injection, pellet 

injection  and various diagnostics), to purify  and isotopically separate 

these gas mixtures (consisting of  the six hydrogen  molecules, helium 

and impurities  such as hydrocarbons, oxygen, nitrogen, etc) and to re 

inject  pure tritium  and deuterium  gas into the torus. 

The system is situated  in a separate building and can be separated 

into sub-systems as shown  in the block  diagram (Fig.34).  During the 

year, the inactive commissioning  phase of the sub-systems continued  in 

6 3 

T E C H N I C AL 

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J ET 

From  Torus 

Mechanical 
forevacuum 

Exhaust 
detritiation 

r^ 

Cryogenic 
forevacuum 

Impurity 
processing 

Analytical 
laboratory 

Intermediate storage 

Isotope 
separation 
Gas chroma 
tography 

Isotope 
separation 
Cryo-
distilfation 

D + T 

Product 
storage 

To 
NI 
-4-

To 
Torus 

J 25 

Fig.34:  Block  diagram  of  Active  Gas  Handling  System  (AGHS) 

accordance with the JET programme for  D-T operations  in  1996.  Some 

sub-systems are almost  ready for  active commissioning with  tritium 

trace amounts, which  is planned for  mid-1995. 

With  the  commissioning  of  the  individual  sub-systems  being 

substantially  complete, the  emphasis  has been  on operation  of  the 

system  in  an  integrated  way  under  conditions  relevant  to  the  actual 

tritium  processing  cycle.  For example, the  gas  chromatography 

isotope  separation  system  has been  operated  in conjunction  with  the 

intermediate  and  product  storage  using  uranium  beds  in these sub 

systems for  hydrogen  isotope  pumping. 

The completion  of  Safety Related Commissioning  Procedures is a 

pre-requisite for  obtaining  approval for the start of  commissioning  with 

64 

T E C H N I C AL 

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J ET 

Preliminary  Safety 
Report 
(PSR) 

Pre-Construction 
Saftey 
Report  (PCSR) 

Approval  of  Desig 
Concept  and 

Principles 

Approval  of  Detailed 
Design  against  Safety 
Principles 

Pre-Commissioning 
Safety 
Report  (PCMSR) 

Inactive 
Commissioning 
Tests 

Approval  of  Active 
Commissioning 

Operational 
Approval 

Pre-Operational 
Safety 
Report  (POSR) 

^ "v 

Active 
Commissioning 
Tests 

J 

^y 

Fig.35:  JET  Safety  Submissions 

tritium.  These cover specific conditions  such as leak  rates, ability  to 

transfer  gas to special tanks for  processing, and verification of  the 

correct operation  of  all valves from the control system.  A block  diagram 

of  the safety submissions  required are set out  in Fig.35. 

The formal  safety  documents  which  comprise  the  safety  case for 

the  AGHS have  been  endorsed  by the  Fusion  Safety  Committee  and 

the  UKAEA's  Safety  and  Reliability  Directorate  (SRD).  In  particular, 

the  Pre-Commissioning  Safety  Report  (PCMSR) was  endorsed  during 

1994 and a number  of  procedural  aspects  requiring  resolution  were 

identified.  Once these  are cleared, approval  for  tritium 

commissioning  of  the  AGHS is expected  from  UKAEA, which  should 

enable  it  to  proceed  in  mid-1995. 

The  UKAEA's  Safety  and  Reliability  Directorate  (SRD) endorsed  the 

Preliminary  Safety  Report  (PSR) for  the  torus  which  provides  part  of 

the  required  documentation  to  use tritium.  Work  then  began  on  the 

65 

T E C H N I C AL 

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next  phase  of  documentation,  the  Pre-Construction  Safety  Report 

(PCSR).  This report  will  provide  a comprehensive  description  of  the 

torus  and  its auxiliary  systems, as well  as safety  analyses of 

postulated  accidents  and  probabilistic  risk assessments. 

An  analysis was  undertaken  to  assess the torus  and  its  auxiliary 

systems for  their  intrinsic  radiological  hazard.  This work  was 

completed  and  conclusions  accepted  by the  UKAEA  and the  Fusion 

Safety  Committee.  Much  of  the  supporting  work  for  the PCSR 

analyses was  completed  in  1994 and  a few  remaining jobs  initiated. 

This work  is expected  to  be completed  in early  1995. 

Technical  Preparations  for  D-T  Experiments 

A  DTE1 Technical  Preparations  Group  was set  up to  identify  the 

technical  work  which  needed to  be completed  before the  DTE1 

experiments  could  take  place  in  1996.  In particular,  it  was  important 

to  identify  all the work  which  needed to  be  ncluded  in the  1995 

shutdown,  bearing  in  mind  that  this  would  be the  last  major 

shutdown  before  DTE1.  The group  also took  note  of  additional  work 

which  would  be  required  for  DTE2  in  1999,  but  the  main  emphasis 

was on  DTE1.  For this  purpose,  DTE1 was taken  to  be a programme 

producing  =2 χ 10J0D-T  neutrons  in a three-month  period  at  the  end 

of  1996, with  provision  for  a remote  handling  divertor  configuration 

change  immediately  after  the  experiment. 

The technical  work  required  to  prepare  for  DTE1 was  identified 

and  in-vessel  work  to  be accomplished  in  1995  was  included  in the 

in-vessel  planning. The ex-vessel work  was  included  in the  ex-vessel 

planning,  and  a large  amount  of  ex-vessel work  could  be 

accomplished  within  the  planned  shutdown  duration. 

It was shown  that  there  were  no technical  or  approval  obstacles 

which  could  not  be overcome  in time  for  DTE1 to  take  place at  the 

end of  1996  and  experience  with  the  PTE meant  that  there was 

confidence  that  the  programme  could  be executed  in a safe and 

professional  manner.  However,  it was  clear  than  improvement  in the 

reliability  of  JET operation  was  required  before  DTE1 could  take  place 

efficiently. 

When  the  present  operating  campaign  comes to  an end, the 

process of  specifying  the  content  of  the  DTE1 programme  in  more 

detail  should  be started.  This would  be difficult  in  1995  when  there 

66 

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was  no experience  of  Mark  II divertor  operation.  Nevertheless,  the 

preparation  of  specific  experimental  proposals  should  begin. The 

proposals  should  address the  main  areas already  identified: 

(a)  refine  the  objectives  of  each  experiment; 

(b)  estimate  the  experimental  time  required; 

(c)  estimate  the  required  neutron  production; 

(d)  define  the  D-D  comparison  discharges; and 

(e)  specify  the  plasma  performance  to  be demonstrated  in  D-D 

before  the  experiment  could  proceed. 

67 

S C I E N T I F IC 

A D V A N C ES 

D U R I NG 

1 9 94 

Scientific  Advances  During  1994 

Introduction 

The overall  objective  of  the  Project  is to  study  plasma  in  conditions 

and with  dimensions  close to  those  needed  in a fusion  reactor.  The 

central  values of  temperature,  density  and  energy  confinement  time 

required  for  a reactor  operating  with  deuterium  and  tritium  are such 

Magnetic  Field  Configuration 

that  the  fusion  triple  product,  (πτΕΤ),  must  exceed the  value  of 

5x102'm"3skeV.  Typical  values for  these  parameters  which  must  be 

attained  simultaneously  in a reactor,  are given  in Table  6. With  ohmic 

The toroidal  and poloidal  magnetic  fields 

combine  to  form  helical  magnetic 

field 

lines,  which  define  a  set  of  magnetic 

surfaces. As thestrengths  of the  magnetic 

fields  vary across the minor  cross-section 

heating  alone  in JET, temperatures  of  3keV  and 4keV for  the  ions 

of the machine,  the pitch  of the field lines 

and  electrons,  respectively,  densities  of  4x10,9nr3  and  energy 

confinement  times  of  1s are the  limits  that  have  been achieved. 

These parameters  were  obtained  simultaneously  during  one 

discharge  and  resulted  in a fusion  product  of  1.2x1020m3keVs. 

Poloidal 
Field 

Major 
Radius 

Minor 
Radius 

Toroidal 
Field 

Resultant Helical Field 

69 

vary and usually decrease with  increasing 

minor  radius.  The number  of turns a field 

line  must 

traverse  around 

the  major 

direction  of  the  torus,  before  closing  on 

itself, is denoted  by  the safety  factor,  q. 

Of  special  importance  are  the  positions 

where  q is numerically  equal  to  the  ratio 

of  small  integers,  as  these  regions  are 

specially  sensitive 

to 

perturbations. 

Instabilities 

arising 

from 

these 

perturbations 

can  result 

in  enhanced 

energy losses. 

In  addition, 

the  maximum  plasma 

pressure,  which  can be maintained  by a 

given  magnetic 

field  is  dependent  on 

the  plasma 

current 

value. 

The 

effectiveness  with  which  the  magnetic 

field  confines  the  plasma  is given  by  β, 

which  is defined  as the  ratio  of  plasma 

pressure  to  the magnetic  field  pressure. 

JET can be operated  with  elongated 

plasma cross-section ratherthan  circular. 

This enables  larger  plasma  currents 

to 

be  carried  for  given  values of  magnetic 

field,  major  radius  and  minor  radius,  as 

well  as producing 

larger  values  of  β. 

S C I E N T I F IC 

A D V A N C ES 

D U R I NG 

1 9 94 

TABLE 6: REACTOR  PARAMETERS 

CENTRAL  ION  DENSITY,  n 

2.5x1020nr: 

GLOBAL  ENERGY CONFINEMENT  TIME, τΕ 

1-2s 

CENTRAL ION TEMPERATURE, Τ 

1 0-20keV 

Breakeven 

This condition is reached when the power 
produced from fusion reactions is equal 
to  that  necessary  for  maintaining  the 
required temperature and density in the 
plasma  volume. 

However,  higher  peak  values  of  electron  anc  ion  temperature  have 

been  reached  using  additional  radio  frequency  heating  and  neutral 

beam  heating  and  combinations  of  these  methods.  Even  so,  these 

substantial  increases  in  temperature  were  associated  w i th  a 

reduction  in  energy  confinement  time  as  the  heating  power  was 

increased.  Thus,  gains  in  plasma  temperature  have  been  partly  offset 

by  degradation  in  energy  confinement  time,  t he  fusion  product 

values  obtained  have  not  shown  the  full  gains  anticipated  over 

Ignition 
Ignition of a mixture of deuterium and 

conditions  with  ohmic  heating  only.  However,  a  substantial  increase 

tritium would be reached if  the power 

in  the  values  of  the  fusion  product  has  been  achieved,  by  operating 

produced by the alpha-particles (20% of 

the total thermonuclear power) released 

in  the  so-called  magnetic  limiter  (X-point)  configuration.  During  the 

from the fusion reactions is sufficient to 

1991/92  campaign,  values  of  9-10x1020rrv3skeV  were  obtained  using 

maintain the temperature of the plasma. 

up  to  16MW  of  additional  heating. 

Higher  values  of  temperature,  density  and  energy  confinement 

have  been  obtained  individually  in  separate  experiments,  but  not 

simultaneously  during  one  discharge.  These  include  peak  ion 

temperature  up  to  30keV,  energy  confinement  times  up  to  1.8s  and 

central  densities  up  to  4x1020nrr3. 

Experimental  Programme 

The  strategy  of  JET  is to  optimise  the  fusion  product  by  building  up 

a  high  density  and  high  temperature  plasma  in  the  centre  of  the 

discharge,  while  still  maintaining  an  acceptable  high  confinement 

time.  These  conditions  should  ensure  that  sufficient  alpha-particles 

are  produced  with  deuterium-tritium  operation  so  that  their 

confinement  and  subsequent  heating  of  the  plasma  can  be  studied. 

The  original  scientific  programme  of  JET was  divided  into  four 

phases. The  Ohmic  Heating,  Phase  1, was  completed  in  September  1984 

and  Phase  II - Additional  Heating  Studies  - started  early  in  1985.  By 

December  1986,  the  first  part,  Phase  IIA,  had  been  completed.  The 

machine  then  entered  a  planned  shutdown  for  extensive  modifications 

70 

S C I E N T I F IC 

A D V A N C ES 

D U R I NG 

1 9 94 

and enhancements  before  the  second  part  of  the  Additional  Heating 

Studies,  Phase IIB, which  started  in June  1987. The objective  of  :his 

phase, from  mid-1987  until  late-1988,  was to  explore  the  most 

promisirg  regimes  for  energy  confinement  and  high  fusion  yield  and 

to  optimise  conditions  with  full  additional  heating  in the  plasma. 

Experiments  were  carried  out  with  plasma  currents  up to  7MA  in the 

material  limiter  mode  and  up to  5MA  in the  magnetic  limiter  (X-

point)  mode and with  increased  radio  frequency  heating  power  up  to 

18MW  and  neutral  beam  heating  power  exceeding  20MW  at  80kV. 

The  ultimate  objective  was  to  achieve full  performance  with  all 

systems operating  simultaneously.  Phase III of  the  programme  on  Full 

Power  Optimisation  Studies  started  in  1989  and was  completed  in 

early  1992.  In  1991, JET's lifetime  was  prolonged  by four  years  until 

the  end  of  1996. The extension  was approved  to  allow  JET to 

implement  the  new  Pumped  Divertor  Phase of  operation,  the 

objective  of  which  is to  establish  effective  control  of  plasma 

impurities  in operating  conditions  close to  those  of  the  Next Step. 

This programme  of  studies  will  be pursued  before  the  final  phase of 

full  D-T  operations  in JET. 

The  1991/92  experimental  period  concentrated  on:  optimisation 

of  plasma  performance;  advancing  understanding  in certain  key 

areas of  tokamak  physics, (such  as: physics  of  the  Η-mode;  energy 

transport  and confinement;  and transport  of  particles  and 

impurities);  and  establishing  the  basis for  pumped  divertor  and  Next 

Step physics  (including  a Preliminary  Iritium  Experiment). 

Operating  Modes 

Under  normal  operating  conditions  the 

magnetic  surfaces are nested inside each 

other.  The edge of the plasma is defined 

by the magnetic  surface which  intersects 

the limiter.  The only magnetic  field  lines 

intersecting 

the  walls  of  the  chamber 

are  those  beyond  the  region  bounded 

by  the limiters  as shown  in  the  diagram 

on theleft.  This is termed  material  limiter 

operation. 

The  magnetic 

field 

configuration 

on JET can  be  modified  so  that  one  of 

the  closed  surfaces  near  the  limiter  is 

opened  up so that  it  intersects  with  the 

vacuum vessel wall. In this  configuration, 

the  magnetic  separatrix  is  moved 

to 

within  the  vacuum  chamber. 

This so called X-point  configuration 

(or  magnetic  limiter)  can  be  operated 

with 

the  two  nulls  of  the  separatrix 

within 

the  vacuum  chamber 

(double 

null)  or with  only  one inside  (single  null) 

as shown  in  the  diagram  on  the  right. 

During  X-point  operation 

with 

additional  heating, 

the  plasma  can 

behave,  with  respect  to  confinement, 

as  though 

its  edge  were  bounded  by 

limiters.  This is called  the Low  (L)-mode. 

Under  certain 

circumstances, 

the 

plasma  can  be  induced  to  behave  in  a 

different  manner  which produces  better 

plasma  confinement.  This is termed  the 

High  (H)-mode  of  operation. 

Limiter  Operation 

Divertor  Operation 

2.0 

1.0 

-1.0 

-2.0 

Belt limiters 

2.0 

Poloidal 
miter 

1.0 

-1.0 

-2.0 

Target  plates 

1.0 

3.0 
2.0 
Major radius (m) 

4.0 

3.0 
2.0 
Major radius (m) 

7  1 

S C I E N T I F IC A D V A N C ES D U R I NG 1 9 94 

The machine entered a major shutdown in early 1992, which 

lasted throughout 1992 and 1993, to prepare the machine for the 

divertor phase of operation. At the start of 1994, JET was nearing the 

end of the longest and most extensive modifications since initial 

assembly of the device. During the long shutdown, the interior of the 

vacuum vessel was essentially replaced. Following completion of this 

work, JET was in a position to begin its planned programme of 

operations to demonstrate effective methods of power exhaust and 

impurity control in conditions close to those envisaged for ITER. 

Main Scientific Results 

The main objectives oí lhe 1994 campaign period were: to assess 

divertor performance on the vertical and horizontal targets of the Mark I 

Disruptions 

divertor; to demonstrate the high performance capability of the JET 

There is a maximum value of density 

which can be contained with a given 

pumped divertor; and to study those physics areas in which JET could, _ 

plasma current. If this value is exceeded 

a disruption occurs when the plasma 

confinement is suddenly destroyed and 

the plasma current falls to zero in a short 

period of time. Under these conditions 

high mechanical and thermal stresses 

are produced on the machine structure. 

Disruptions are thought to be caused by 

instabilities mostly developing on the 

magnetic surface where q=2. 

make important contributions to ITER and DEMO. The main scientific 

results obtained during the 1994 campaign are set out below. 

Power Handling and Particle Control w i th 
the Mark I Pumped Divertor 
Power Handling 

The study of divertor performance under steady-state conditions at high 

power has been a central goal of the programmie. Particular care was 

taken in the design and installation of the power handling surfaces of the 

Mark I divertor target to optimise power handling, in particular, by 

avoiding exposure of tile edges. The strike points could be swept over the 

target plates at 4Hz with an amplitude of -10cm to increase the area of 

the target effectively "wetted" by the plasma. The Mark I divertor 

showed excellent power handling capability, with no "carbon blooms" 

limiting the performance, as seen in the past. The performance of the 

target was investigated over a wide range of plasma conditions, at 

powers up to 26MW and in steady-state Η-modes lasting up to 20s. In no 

case, including hot-ion Η-modes where performance was previously 

limited by impurity influxes, has a bloom of carbon impurity been 

observed. The combined effects of sweeping and giant ELMs, which are 

now a feature of Η-modes, are such that the surface temperature of the 

target has not exceeded 1200°C, even in cases where over 100MJ of 

energy has been delivered to the horizontal target. 

72 

S C I E N T I F IC A D V A N C ES D U R I NG 1 9 94 

1 

\ 
V_____ 

5 
I. 10 

ΐ Γ~ 

Neuton  reactioji^__^-__-^~ 

D. 

l> 2 

ID 

b 1 
S 0 

d 

rate , / '^ 

D« 

Ô 

-
o 
CL 800 
E 
Φ 
r­
E 600 
E 
I40O 

Γ.  . . . | _. 
12.0 

^_ 

V 

ν 

L 

Avjout 

| 

I 

!  " 
|  |  ;j 

I 
12.5 

Ί 

ι 
13.0 

Time  (s) 

1 
13.5 

3 

I 
14.0 

Fig.36: Time  evolution  of  tile  temperature  during  swept  and  unswept  cases 

From the  power  handling  point  of  view,  the  appearance  of  edge 

instabilities  (ELMs)  had two  effects.  Firstly, they  deposit  power  on 

surfaces  other  than  the  strike  zones, and  secondly,  since they  create 

a spike  ir  the  radiated  power,  the  power  conducted  to  the  strike 

zones  decreased  immediately  after  the  ELM event.  In this sense, 

ELMs are extremely  effective  in alleviating  the  heat  load  on the  strike 

zones,  but  the  penalty  is the  uncontrolled  deposition  of  power  in 

zones which  may have a poor  power  handling  capability.  In Fig.36, 

the  time  evolution  of  the  target  temperature  in the  two  strike 

regions  is shown  for  a discharge  in the  hot-ion  Η-mode  regime 

(lp=5MA,  B,=3.4T,  Pm=20MW).  Between  giant  ELMs, the  power 

imbalance  changed  between  inner  and  outer  strike zones. 

Sweeping of  the strike points over the target  plates spreads the 

power  load and improved the power  handling  capability of  the tiles. 

Fig.37 shows the  power  injected during two  pulses, one with  and one 

without  sweeping: the current  in the divertor  coil  D2 is an indicator  of 

the sweep, and the time evolution of  the  maximum temperature  of  the 

tiles is shown for  the two  cases.  In the swept case, the peak 

temperature  decreased when the sweep  began and stabilised at 

~550°C, in contrast  to the  unswept  case where the peak  temperature 

73 

Density  Control 

Increasing  the  density  can  be  achieved 

by  introducing  additional  gas  into  the 

vacuum  vessel,  by  the 

injection 

of 

energetic  neutral  atoms  (neutral  beam 

heating)  and  by solid  pellet 

injection. 

Increasing  the  input  power  to  the 

plasma 

through  additional 

heating 

raises 

the  electron 

density 

limit. 

However,  problems  can  occur  when 

this  heating  power 

is switched  off,  if 

the  electron  density  is  too  high.  To 

overcome  this  problem, 

the  plasma  is 

moved,  prior  to the switch-off  point,  so 

thatitbearson 

thecarbon 

tilescovering 

the  inner  wall.  The  tiles  have  been 

found, 

to  provide 

a 

pumping 

mechanism 

for  removing  particles  so 

that  the  density  can be reduced  below 

the  critical 

limit. 

S C I E N T I F IC A D V A N C ES D U R I NG 1 9 94 

Swept 

Unswept 

■ Pulse No: 30092, swept 
Pulse No: 30093, unswept 

O  800­

o5 6 0 0-

400 

10 

12 

20 

22 

Fig.37: Time evolution of the surface temperature in the two strike regions 
for a discharge in the hot­ion mode 

continued to increase steadily. With sweeping, the effective power seen 

by the target was ­66% of the power in the unswept case. 

An example of the power handling capabi ¡ties is shown in Fig.38, in 

which the discharge is subjected to 25MW of combined neutral beam 

and ICRF heating. 72MJ of energy was conducted to the tiles and their 

surface temperature increased ~650°C. In other discharges, almost 

140MJ of neutral beam energy was injected during ELMy H­mode 

plasmas, raising the surface temperature of the tiles to ~550°C. In the 

X­point configuration of 1991/92, an input energy of about 15MJ was 

sufficient to raise the tile temperature above 1500°C and provoke a 

severe impurity influx (carbon "bloom") which quenched the discharge. 

The vertical side plates, intended primarily as protection for the 

divertor coils, have also been used as the targets for ohmic and ELM­

free Η­modes with neutral beam heating. In line with model 

predictions, the density scrape­off layer is narrower and the 

temperature profile is inverted with operation on the vertical target. 

This result gives confidence that the basic modelling assumptions are 

correct and with further validation and refinement will provide the 

tools to help design a reliable ITER divertor. 

74 

Sawteeth 

Perturbations  on  the  q=1  magnetic 

surface  can  result  in  the  formation  of 

large 

fluctuations 

in 

the 

central 

temperature 

and  density. 

These 

fluctuations 

have  been 

termed 

'sawteeth'.  They  are  also  associated 

with 

the  expulsion  of  energetic 

ions 

from  the  central  region  of  the  plasma. 

Understanding  this process is 

important 

as  the  alpha-particles  produced 

from 

deuterium-tritium 

fusion 

reactions 

might  be lost  before  they  can  produce 

any  effective  heating  of  the  plasma. 

S C I E N T I F IC A D V A N C ES D U R I NG 1 9 94 

3ulse No: 31407 

2 MA/2.8 Τ 

rf^\ AynVwv//wv^VA 

O 

600 

400 

Ü 

500 

400 I­

Ί  \nita>Mir*:  w 

■^Λ^^^ί^!NΨMk^  (d) 

4^fff^%i 

(e) 

VA ΛΑ 

12 

13 

14 
Time (s) 

15 

16 

'7 

Fig.38:  Time  evolution  of  a  2 MA  ELMy  Η-mode  w i th  2 6 MW  of  combined 

heating,  illustrating  the  low  tile  surface  temperatures  achieved  w i th  the 

Mark  I divertor  ((d)  and  (e))  w h en  strike  point  sweeping  ((c))  is  applied 

Particle  Control 

Active  particle  control,  including  efficient  helium  exhaust,  is  required 

for  steady-state  operation  in a reactor.  The study  of  the  effects  of 

fuelling  and  active  pumping  on  plasma  properties  has been  one of 

the  main  topics  of  the  divertor  campaign.  In  particular,  an  upgraded 

gas introduction  system, characterised  by toroidally  and  poloidally 

distributed  gas sources  in the  vacuum  vessel, and  an entirely  new 

internal  cryopump  capable  of  providing  active  pumping  during 

plasma  pulses  have  been  extensively  used. 

The torus  cryopump  has proven  to  be effective  for  plasma  purity 

and  density  control,  removing  neutral  deuterium  for  a wide  range  of 

strike  point  positions  on  both  the  horizontal  and vertical  target 

plates.  The  particle  removal  rates  for  the  different  strike  point 

positions  are shown  in  Fig.39 for  typical  ohmic  plasmas. The  variation 

in the  removal  rate  is useful  in facilitating  density  control  in  steady-

state  ELMy  Η-mode  plasmas as shown  in  Fig.40 for  three  different 

magnetic  configurations.  The  particle  removal  rate  is several  times 

the  maximum  neutral  beam fuelling  and  this  helps  deplete  gas 

reservoirs  in  plasma  facing  surfaces  and  reduce recycling. 

75 

S C I E N T I F IC A D V A N C ES D U R I NG 1 9 94 

(a) 

— 6 

S 4 

o 
E 
ω 
rr 

2.82 

2.90 2.98 

3.06 -1.70 -1.62 

1.54 

Strike point major radius (m) 

Strike point height (m) 

Fig.39: Particle removal rate as a function of separatrix position on the horizontal plates (a) and on the vertical 
plates (b). The full sweep of separatrix position is carried out in ­10s 

The efficiency of the cryopump in reducing recycling has been 

exploited in high performance hot­ion Η­modes to the point where the 

neutral beam particle source dominates fuelling from recycling. Hence, 

peaked density profiles, favoured for improved neutral beam penetration 

in this regime, are maintained and lead to improved fusion performance. 

1x1020 

\ 

\® 
\ 
I 
S 
\ 
\ 

V"Ô,, 

■\ \ 

\l 

Λ l'A 
f \ / \\\4 
/ / '^ 

V H·· 

\ypXs'fX 

/ 

J 
/ 
1 
1 
1 . ...·■ 

/ 

1 
1 
1 
ι! 
1; 

Γ 

/ 

5x1019 

Tl 

­

η 
12 14 16 18 20 

1 

I 

1 
22 

24 

Time (s) 

Fig.40: Density control in ELMy Η­modes with 15MW NB heating showing 
variation of strike point relative to divertor cryopump: (1) Strike point near 
pump, no added gas; (2) Strike point near centre of horizontal target, no 
added gas; (3) As (2) but with added gas flow of ~2x1022D"s ' 

76 

S C I E N T I F IC A D V A N C ES D U R I NG 1 9 94 

10 

8 

5 
2 
È 6 
o 
Q. 
+ 
m 
a. 4 
II 
Ι­
Ο 
r­
Q. 

2 

C» 

Π 

D 

a* 
r¿n D 

f D g DO 

DD Π Π% Π 

uc 
) 2 4 6 

I ι 

D L 
• L ­H 

I 
8 10 

Volume averaged density n8 (1019 

m­3) 

Fig.41: Power threshold for Η­modes at 2MA as a function of density 

Current Profile Control 

The highest current density exists at the 

centre of the plasma, as this is the hottest 

region and the electrical resistivity 

decreases as the temperature increases. 

Without sawteeth, which occur on the 

q=1 surface, this high current density 

region would be squeezed or pinched 

inwards. Selective heating outside the 

central region would remove the q=1 

surface from the plasma and so avoid 

the onset of the sawteeth. Another way 

is to decouple the plasma current and 

CO 

en 

temperature profiles. On JET, it is 

intended that an electric current, 

additional to that generated by 

transformer action, should be produced 

by neutral beams and by radio-frequency 

power at 3.7GHz (called Lower Hybrid 

Current Drive (LHCD)). 

Η­Mode Threshold and Confinement 

Transition from low confinement (L­mode) to high confinement (H­mode) 

regimes have been achieved with similar input powers as those required 

during previous campaigns. During 1994, a more thorough study was 

carried out and demonstrated a clear scaling of Η­mode threshold power 

with plasma density and with toroidal magnetic field. Figure 41 shows 

the power threshold as a function of density carried out at constant 

toroidal field. When combined with results from other tokamaks 

operated with the same geometry (for example, ASDEX­Upgrade and 

COMPASS in Europe), a more accurate size scaling of threshold power 

will be determined and will be used in predictions for ITER. 

In previous operation, edge instabilities (ELMs), were rarely 

encountered in the Η­mode and were only observed either just above 

the L­H transition threshold; with extremely strong gas fuelling; or at 

high beta after a period free of ELM instabilities. The plasma in the new 

configuration now develops regular ELMing as standard behaviour. 

Long ELM­free periods (greater than 1s) can only be obtained with 

special preparation. These edge instabilities have the characteristics of 

'giant' ELMs identified on other machines and have an ELM frequency 

which increases with input power and reduces with current. 

77 

S C I E N T I F IC A D V A N C ES D U R I NG 1 9 94 

(a) 

1.0 

ELMy H­mode x 2·2/ 

+ 1.0MA /' ,­' 
O 1.5MA 
D 2.0MA 
Δ 2.5MA 
Ο 3.0MA / ' /' γ * . ­ '' 
• 4.0MA / ' , ·' ¿J 
▼ 5.0MA ML^rf1'' 

s 

T3 

H 0.5 

,­'x2.0 

_,­'x1.6 

* 

(b) /'x3.0 ,­'x2.4 , . ·' 

1.0 

ELM­free H­mode / ,­'' , / x 2 .0 

01.5MA / HÖ / 
D2.0MA / 'o Á / 
Δ2.5ΜΑ ,¿. /o / 
O3.0MA ^ä3S>' ° '' + 
+ 4.0MA '

' Æ B P / '0 

/^IffiW 

/' /

/U 

en 
S 0.5 

/y ° 

0 
/' I 
( 
) 0.5 

2 χ τΕ (ITER89­P) (s) 

2 χ τΕ (ITER89­P) (s) 

I 

3 
1.0 

η 

/' I 
3 0.5 1. 

Ö 

Fig.42: (a) Measured confinement time (xEDIA) for 1994 JET ELMy Η­mode dataset versus twice the prediction from 
ITER89­P scaling law (tEITER89P) (Data are quasi­steady­state); (b) for 1994 ELM­free Η­modes with TELMFREE>0.5S 

As a consequence, two approaches to high performance can be 

pursued: long pulse, ELMy Η­modes, which maintain steady plasma 

conditions for many energy confinement times (considered the most 

credible mode of operation for ITER); and ELM­free, hot­ion H­modes 

in which the highest performance is achieved transiently. 

ELMy Η­modes show a typical enhancement factorin confinement 

time, H=1.8, but can rise to H=2.5 in certain circumstances, as shown in 

Fig.42(a). However, confinement in the ELM­free phase is typically a 

factor, H=2.2­2.4 higher than low confinement (L­mode) scalings (see 

Fig.42(b)). This value is not significantly different from past campaigns, 

although the absolute level of confinement is, as expected, about 15% 

lower, since the plasma volume with the divertor is now smaller. 

Long Pulse Steady­state H­modes 

The excellent power handling capabilities of the divertor target, together 

with the in­vessel cryopump and the facility for sweeping the strike zone 

position, have been exploited to produce long pulse steady­state 

H­modes. The H­modes produced are characterised by frequent regular 

giant ELM behaviour, a regime which is now the most commonly 

observed at JET. 

An example of the temporal evolution of a discharge of this type is 

shown in Fig.43. This shows the longest duration Η­mode achieved, 

78 

S C I E N T I F IC A D V A N C ES D U R I NG 1 9 94 

Pulse No: 31703 

ilIIUlltlüi 

2MA/2.1T 

D« 

J^rj¿l¿p(w)!MMH4&^'*^^ 

> 

Wd 

. « A V r ^ V ^ / W ^ V W / ' / ^ Î ^ V V ï. 

u^^jii ι·' iliLll.il I.lililí l'tlj 

Prad 

Fig.43: Time evolution of a long pulse steady­state ELMy Η­mode at 2MA 

with a pulse length of 20s, which was limited only by machine 

engineering constraints. The Η­mode phase in this 2MA discharge lasts 

for ­50 energy confinement times (~50τΕ). The regime has been 

extended up to plasma currents of 3MA for at least 20 energy 

confinement times. The plasma density, the effective charge (Zeff), stored 

energy, recycling conditions, plasma temperature, plasma beta and 

radiated power all remained constant for 20s. The radiated power (at 20­

40% of input power) and Zef| (­1.6) were typically low in the ELMy phase 

of these plasmas. An important result is that the divertor target 

temperature remained far below the design limit during the pulse. 

During the Η­mode phase of this discharge, plasma fuelling was 

by neutral beam injection alone, and the divertor cryopump 

controlled the density, with no saturation of the pumping. This 

contrasted with previous operation in 1991, when wall pumping 

saturated after about 10s and no true steady­state could be 

achieved. The energy confinement time was enhanced by a factor, 

H=1.8, above L­mode confinement. The demonstration of these 

steady­state H­modes with edge safety factors, qg5, as low as 2.9 ¡s a 

relevant ignition scenario and is a significant step along the road to 

the successful operation of ITER. 

79 

MARFE 
A MARFE (Multifaceted Asymmetric 

Radiation From the Edge) is a toroidally 

symmetric band of cold, highly radiating 

plasma which normally forms at the 

plasma inner wall. It can occur when the 

plasma edge density is high and results 

from of an imbalance between the power 

flowing along magnetic field lines in the 

edge and the power lost locally due to 

radiation. A MARFE grows rapidly, on a 

timescale of =10-100 milliseconds, but 

it can persist for several seconds. In 

some cases, the MARFE leads to a 

disruption, but in others the main 

consequence is a reduction in the edge 

density. 

ELM 
An ELM (Edge Localized Mode) is an 

edge instability which occurs in the high 

confinement (Η-mode) regime. It affects 

a narrow region in the plasma edge and 

leads to a loss of particles and energy 

from the edge on a timescale <1 

millisecond and therefore is a rapid, but 

transient, instability. However, ELM'scan 

occur as repetitive instabilities which 

cause a reduction in the time-averaged 

energy and particle confinement time. 

S C I E N T I F IC 

A D V A N C ES 

D U R I NG 

1 9 94 

1.0 

-^k-

1 
I 
I 
1 

0.8 

0.6 

1 
t 
I 
/ 
/ 
/ 
/ 
—  / 
/ 
/ 
/ 
/ 
/ 

0.4 

Α. 

*7~ 

0 . 2-

target 

\ 

• 

\ 
\ 
\ 
\ 

/ 
J 
r 

_ /\ 
W 
/ 

\ 
\ 

\ 
\ 
\ 
L 

Γι 

·· 
\ Γ 

γ ι, target 

\ · 

\ ­­D' 

Ο·' Λ. 

• 

/ 
·/ 

"°··. 

\ 
\ 

\ 

\ .■ 

Λ 
/ \ 

\ 
\ 

S 

Ό 

1 
3.0 

1 
4.0 
ne(101 9m­3) 

poa 

• 

N^^ · 

\* 

­ ­ Α­

Ι 
5.0 

Fig.44: Time evolution of deuterium flux to target (Γ\), recycling light 
(Dn) and carbon influx (Γ.) showing development towards detachment 
at high density 
Gas Target/Radiative Divertor Plasmas 

While poloidal divertors have been employed successfully in many 

tokamaks to reduce impurity contamination of the plasma core, an 

operating scenario which scales to reactor plasmas has not yet been 

demonstrated. The exhaust of particles, momentum, and power from a 

burning plasma is such that, the divertor target plates would not survive 

unless a significant fraction of the exhaust is redirected across magnetic 

fields lines by volume loss processes. In addition, it has been shown that 

radiation alone cannot be used to solve the power loading problem in a 

so­called high recycling divertor where there is no loss of momentum 

along the scrape­off­layer (SOL) field lines. A pressure drop of a factor 

of ten is consistent with a radiative fraction su­'ficient to satisfy 

constraints on ITER target loads (~5MWm2). 

When the core plasma density is increased by some fuelling 

method, the plasma density in the divertor typically increases, the 

dependence being as high as the third power. This staggering rate of 

increase is self­limiting, because at the same time the temperature in 

the divertor and the SOL is rapidly decreasing and losses, such as charge 

exchange and radiation, begin to dominate. A: this stage, the density at 

80 

S C I E N T I F IC 

A D V A N C ES 

D U R I NG 

1 9 94 

h—  Div instability 

Pulse No: 31204 

2MA,  2.8T 

JS AT  (Inner) 

unUfcHoUJ  ill,UKk4q» 

JS AT  (Outer) 

mHnWfrif ■> 

___. iW' 

y—|/ |f 

ISAT (ref) 
/ 

0.4 

0 
<  0.4 

""> 
E 

Ό 

2 

ri 
m 

0 

5 

0 

4 

0 

20 

0 

Divertor 

JET was originally configured as a "limiter 

Tokamak" where the edge oftheplasma 

(the "last closed flux surface"-L CFS) was 

defined by contact with a material 

boundary called a limiter, which absorbs 

the exhaust power of the plasma. Since 

the edge of the plasma is quite hot, 

material is eroded by sputtering and the 

sputtered impurities enter the plasma 

relatively easily. This enhances radiative 

14 16 18 20 22 

24 

losses and dilutes the plasma, which 

Time (s) 

Fig.45: Time d e v e l o p m e nt to d e t a c h m e nt in a 2 MA L-mode discharge 

the divertor plates decreases as the core density, is increased even 

further. This regime, where not only the density but also the flows of 

particles, momentum, and energy are decreasing functions of increasing 

core density, is the so-called detached plasma regime and is the 

favoured candidate for future burning plasma tokamaks. 

However, the onset of volume losses is gradual as the SOL/divertor 

temperature drops and presents a problem of defining exactly when a 

plasma is detached. Complete detachment and the concomitant 

reduction of heat flux to the divertor target plates was first seen during 

the 1991/92 campaign. With the Mark I divertor, this regime has been 

extended to cover a broad range of plasma conditions and magnetic 

configurations at high density in ohmic, L-mode and Η-mode plasmas. 

lowers the fusion reaction rate. 

The JET vacuum vessel and 

magnetic field system have been 

modified to operate in a "divertor" 

mode. The field configuration includes 

an "X-point" so that the LCFS (in this 

case designated the separatrix) 

bounding the main plasma does not 

intersect the wall. The power crossing 

the separatrix is transmitted in a thin 

layer called the scrape-off layer (SOL) to 

the divertor at the bottom of the vessel 

and is absorbed by the divertor "target 

plates". Divertor operation reduces the 

impurity content of the main plasma 

through a combination of effects. The 

divertor plasma is generally much cooler 

than the main plasma edge, so that 

sputtering and erosion are reduced. 

Moreover, impurities which are 

produced at the target plate tend to be 

retained in the divertor area by friction 

with the plasma streaming towards the 

divertor plates. In addition tocontrolling 

Figure 44 shows results from an ohmic detached discharge where 

impurity content, divertor operation 

the density was increased from 3χ10, 9ητ3 to 6 x 1 0, 9m3 over several 

tends to allow higher SOL temperatures, 

thus facilitating access to high 

seconds. As the density increased, the particle flux density (Γ) to the 

confinement regimes. 

target first increased (the high recycling regime), and then decreased 

dramatically w i th the onset of detachment. The recycling light (Da) 

also increased as the number of excitations per ionisation increased 

rapidly at very low temperature. The carbon influx (Γ.) from the 

divertor plasma temperature fell below that required for significant 

S C I E N T I F IC A D V A N C ES D U R I NG 1 9 94 

W/m3 

500000 

W/m3 

500000 

400000 

300000 

200000 

100000 

400000 

300000 

200000 

1000C0 

2.0 2.5 3.0 3.5 2.0 2.5 

10 -2.0 

3.0 3.5 

Fig.45: Radiation distribution in (a) attached state and (b) detached state 

sputtering, and then decreased further with decreasing temperature 

and particle flux during the detached phase. 

Figure 45 shows an L­mode discharge in which detachment was 

achieved at a power level of 5MW. The gas input was controlled by 

feedback from the ¡on saturation current of one of the Langmuir probes 

close to the inner target. There was a gradual fall in the ion saturation 

current between 16s and 20s, while the density was increasing. The 

pressure at the probe fell while the total radiated power was about 

50%. Plasma radiation was localised near the target plates during the 

attached phase of such discharges, and near the X­point during the 

(partially) detached phase (Fig.46). Studies are in progress to control the 

location of the radiating region. Attempts to achieve complete 

detachment in Η­mode plasmas, by puffing deuterium into steady state 

H­modes, precipitate an increase in the ELM frequency, and eventually 

the plasmas return to L­mode before complete detachment can occur. 

A step forward was achieved by puffing a mixture of nitrogen and 

deuterium into the plasma. With this mixture, it was possible to maintain 

quasi­steady­state ELM behaviour which had not been possible, with a 

mixture of neon and deuterium. The quasi­steady­state regime could be 

established with ­80% radiation (Fig.47) and initial analysis showed that 

about 70% of this was emitted from outside the separatrix. The degree 

82 

S C I E N T I F IC 

A D V A N C ES 

D U R I NG 

1 9 94 

18 
Time  (s) 

20 

22 

End of  D2/N2 
puffing 

Fig.47: Time  evolution  of  an  ELMy  Η-mode  w i th  N2/D2  gas  puffing 
showing  achievement  of  quasi-steady-state  ELM  behaviour  w i th  - 8 0% 

radiated  power 

of detachment  in these discharges is still being investigated, but it seems 

that there is complete detachment  between ELMs, though each ELM 

results in a "burn through"  of the divertor  plasma and a spike in the ion 

saturation current.  The confinement  enhancement  in these discharges 

was relatively modest (H=1.4). The H factor  increased, however, when 

ICRF power was coupled successfully to these discharges, indicating, 

perhaps, a dependence on heating profile.  So far, combined  heating 

powers of  up to 24MW  have been coupled into these plasmas with the 

divertor  target  tile temperature  remaining  constant  at 350'C, without 

sweeping, for the 2.5s duration of the heating pulse.  Further studies at 

higher combined  heating power should provide the most convincing 

demonstration of the relevance of this regime to ITER. 

High  Performance 

The major objective was to develop plasmas with  high core plasma 

performance  in the new configuration. This included the optimisation of 

parameters such as temperature, density, stored energy and fusion yield, 

with a view to defining  regimes of  relevance to future  D-T operation  in 

JET and ITER. The pumped divertor was designed for operation of 

diverted configurations with  plasma currents up to 6MA with safety 

factors of, typically q95~3 (and down to 2.2), in line with  mainstream ITER 

83 

S C I E N T I F IC 

A D V A N C ES 

D U R I NG 

1 9 94 

Pulse No: 17386 

PNB  (MW) 

. J » J L / ^-

10 

PNB (MW)  "u—IT 

Pulse No: 30726 

RDD(1015S 

Teo(keV) >/rA-^KI-VUuvt-

0 
8 
4 

6 

2 

4 
2 

3 
2 
4 
2 
20 
10 
6 

Fig.48:  (a)  Typical  Η-mode  behaviour  in the  old  JET for  1 0 MW  of  beams  into  a  3 MA  plasma  (Note  the  long  ELM-

free  period):  (b)  Typical  Η-mode  behaviour  for  standard  FAT  configuration  in  the  new  JET for  1 0 MW  of  beams 

and  plasma  current  of  3 MA 

design. The improved current  capability  in diverted  configurations 

opened up the domain  of  Η-mode operation and raised the question of 

scaling of  confinement with  plasma current  at such low safety factor. 

In the  previous configuration,  ELM-free  H-modes  of  several seconds 

were the  natural  regime for  any high power  heated  diverted 

configuration, as shown  in Fig.48(a). Above a clearly defined  threshold 

power, the  plasma would enter and remain in this regime  until, 

typically,  increasing  radiated power  reduced the exhaust  power  below 

threshold.  In this case, optimisation  of  performance  required  low 

recycling, density control, optimum  beam deposition  but were limited  in 

plasma current. The hot-ion  Η-mode regime in the old JET was transient 

limited  by carbon blooms, earlier appearance of  giant  ELMs and, in 

some cases, limited  by sawteeth and other  internal MHD. 

In the  new  configuration, the  natural  Η-mode  regime  is ELMy as 

shown  in  Fig.48(b).  In this  case, the  initial  ELM-free  period is 

relatively  short  and  is followed  by repetitive  large  ELMs.  These  larger 

ELMs lead to  steady-state  conditions,  which  can  be  maintained  for 

many energy  replacement  times. This  is directly  relevant  to  ITER, and, 

as such, has been  exploited  effectively.  This difference  in  behaviour 

has provided  a challenge  and  a rethink  of  strategy  for  performance 

optimisation  has had to  be  implemented. 

84 

S C I E N T I F IC A D V A N C ES D U R I NG 1 9 94 

* 1991-1992 (3MA) 
O 1994 FAT 1 ( 2 . 5 - 3 M A) 
Δ 1994 Triangular J Bake at 250°C 

θ 1994 Fat 
Δ 1994 Triangular 
Β 1994 High flux 
expansion 

O 1994 Fat 
Δ 1994 Triangular 
Π 1994 High flux 
expansion 

Ops. at 250°C 

( 2 . 5 - 3 M A) 

■ Bake at 320°C 
Ops. at 320°C 

( 2 . 5 - 3 M A) 
- Bake at 320°C 
Ops. at 250°C 

1991 - 1 9 92 
Be evap + He GDC 

10 

9 

D 

D 

D * 

·:· 

D 

D 

D 

* 

*□ 

* 
**Qn 
* 
D 

Π 

* 

α 

c 

* 
en 

r-, □ 
tbn 

*υ 

* 

I 6 

ro o 

-

-

-

^ C P EP 

* D 

ο ο0 □ 0 
°π ο 0 □ 

° D n □ o ° π 

o o 

ι 

I 
16 
Normalized recycling 10-7 ph/st/s/rrr2 

I 
8 

12 

20 

Full Cryo 

He GDC 

I 

Be evap + He GDC, 

Be evap + He GDC +1/2 Cryo 

Fig.49: The dependence of D-D fusion reaction rate on normalised main 
chamber recycling flux. Data is for 2.5-3MA hot-ion Η-mode plasmas from 
1994 with 15-18MW neutral beam heating. Comparison is made with 3MA 
hot-ion H-modes from the 1991/92 campaign 

A picture is emerging whereby the optimisation of core conditions is 

constraired by a complex web of interrelated physics. Progress has been 

made in understanding some of the key links, such as that between 

configuration and ELM-free period; between recycling and ELM-free 

period; and between recycling, refuelling and core density profiles. 

The high heat load capability of the Mark I divertor target means 

that carbon "blooms", which previously terminated high performance, 

have been eliminated. This has shown, as previously suspected, that 

MHD modes, which precede fast collapses (ELMs) and slow "roll-overs" 

of storec energy and neutron yield, limit high performance. The 

achievement of high performance ELM-free H-modes is strongly 

affected by vessel conditioning, gas recycling from in-vessel components 

and the magnetic configuration. Recycling in the new configuration has 

been higher than previously and this has impacted on performance. 

Impurities 
Impurities released from interactions 
between the plasma and material 
surfaces can have major effects on 
plasma behaviour by causing: 
(a) increased radiation losses; 
(b) dilution of the number of ions 

available in the plasma between 
which fusion reactions can occur. 
A measure of the overall impurity 
level is given by Zefl which is defined as 
the average charge carried by the 
nuclei in the plasma. A pure hydrogen 
plasma would have Zùff =1 and any 
impurities in the plasma would cause 
this value to be increased. In JET, Zefl is 
generally in the range from 1.2-3. 

Major energy losses can result from 

two radiation processes: 

• Bremsstrahlung Radiation - radiation 

is emitted when electrons are 
decelerated in the electric field of 
an ion. The amount of radiation 
emitted increases with Zelr 
Bremsstrahlung radiation imposes a 
fundamental limit to the minimum 
plasma temperature that must be 
attained in a fusion reactor; 
• Line Radiation - heavy impurities 

will not be fully ionised even in the 
centre of the plasma and energy 
can therefore be lost through line 
radiation. 

Considerable effort is made to keep 
the level of impurities in the JET plasma 
to a minimum. The vacuum vessel is 
baked at 300 °Cto remove gas particles 
trappedon the vessel walls which might 
be released by plasma bombardment. 

Interactions between the plasma 
and vacuum vessel walls would result 
in the release of heavy metal impurities. 
To reduce this possibility, the edge of 
the plasma is defined by upper and 
lower belt limiters. These are cooled 
structures circling the outboard torus 
wall with carbon or beryllium tiles 
attached. Carbon and beryllium havea 
relatively low electric charge on the 
nucleus. 

S C I E N T I F IC 

A D V A N C ES 

D U R I NG 

1 9 94 

Pulse No: 33040 

5MA/3.4T 

PNB (MW) 

Fig.50: Time evolution  of  a 5MA  ELMy  Η-mode with  15MW  of  neutral 

beam  injection 

Special attention  has therefore  been paid to  identifying  and reducing 

the source of  higher  recycling and thereby  extending the  ELM-free 

period and improving fusion yield. 

In particular, the divertor targets and the vacuum vessel have been 

baked to  high temperatures  (200°C  and 320°C. respectively)  and the 

high pumping capacity of the cryopump  has been used to deplete neutral 

gas reservoirs in the tiles and walls. Specific magnetic  configurations 

("high  flux  expansion"  in the  divertor)  have  been  developed  to 

reduce  leakage  of  neutrals  into  the  main  plasma from  the  divertor. 

The  importance  of  low  recycling  in achieving  a high  fusion  yield is 

shown  in  Fig.49  where  recycling  levels  in  present  hot-ion  discharges 

are compared  with  those  in  past  campaigns. 

Specific studies have been undertaken to establish the  role of  the 

magnetic  configuration  in improving edge stability  and reducing the 

occurrence of  ELMs. Studies at constant  input  power, constant current 

and constant toroidal field showed that: high flux expansion at high edge 

shear increased the ELM-free period, while  increasing edge shear and 

triangularity  at high flux expansion also increased the ELM-free period. 

In practice, the increased  ELM-free period could not always be 

exploited for  increased stored energy and neutron yield, since a slow 

86 

S C I E N T I F IC 

A D V A N C ES 

D U R I NG 

1 9 94 

(a)  Pulse No: 33643 

4MA/3.4T 

200 

100 

B  C 

-100 

-200 

Fig.51: (a) Time  evolution  of  discharge  with  highest  fusion  yield  in  deuterium; and  (b)  High flux  expansion 

equilibrium  of  discharge  with  highest  fusion  yield  in  deuterium  (compared  with  best  1994 discharge  at  3MA) 

200 

300 

400 

500 

R (cm) 

"roll-over"  often  occured  in these high  performance  plasmas. This 

phenomenon  is accompanied  by a variety of  MHD activity.  In these high 

performance  discharges, it  is this phenomenon  (already  noted in some 

of the best hot-ion  H-modes  of  1991/92) that  limits plasma pressure to 

a normalised  ß„=2.  In hot-ion  H-modes  at higher  plasma current (3.5-

4MA), the phenomenon  was much weaker,  or was entirely absent. The 

limitation then appeared to  be the giant  ELMs or sawteeth. 

High  Plasma  Current,  Stored  Energy  and 
Fusion  Yield 

Experiments in  1994 achieved plasma currents up to  5MA.  At  5MA, 

ELMy H-modes were obtained with  up to  15MW of  neutral beam 

injection (see Fig.50). At 4MA current, steady-state  ELMy H-modes  lasting 

more than four  energy confinement  times, were obtained with  more 

than  18MW of additional  heating.  The maximum stored energy was 

-8MJ  and the fusion triple product  (nDTtE) reached 2.6x1020m3keVs.  The 

highest stored energy achieved was  11.3MJ at the end of the ELM-free 

period of a 4MA discharge with  19MW of  additional heating. 

The best fusion  performance  achieved in  1994 was comparable to 

the  best of  that  in the  past, even though the new  plasmas were  -15% 

smaller  in volume. The record  neutron  rate in  1991/92  deuterium 

plasmas was exceeded. In the  ELM-free phase of  a 3.8MA  hot-ion 

87 

S C I E N T I F IC A D V A N C ES D U R I NG 1 9 94 

JET Fusion Triple Product (nD(0)­TE­T¡(0) 

JG95.81/3C 

12 

;, '■{;·. Q = 1 (Steady state Q(Th)) Pa (Th> 

"loss 

10 

ξ 8 

■= 4 

10 

14 18 22 

26 

30 

T¡ (0) keV) 

η Ττ values have been averaged over 0.2s 

D HOT ION PLASMAS (T¡ > 1.5 « Te) PRIOR TO 1992 

A HOT ION PLASMAS (T¡ > 1.5 · Τβ) 1994 

A­»­­+A Typical Variations between FULL FIT and RAW DATA 

Fig.52: Fusion triple product for 1994 compared with previous campaigns 

H­mode, the neutron rate reached 4.7x10,6s"', more than 10% higher 

Plasma Beta 

The economic efficiency of a tokamak 

reactor is determined, partly, by the 

maximum plasma pressure which can be 

contained by the magnetic fields in the 

device. In particular, the important 

parameter is the plasma beta ßr defined 

as the ratio of plasma pressure to the 

pressure of the confining magnetic field 
(βτ is proportional to nT/B2, where η is 
the plasma density, Τ the plasma 

temperature and Β Tthe toroidal magnetic 

field). This limit expected theoretically, is 

the so-called Troyon limit 

ßT(%)=2.8lp(MA),BJT)a(m), where lp is the 

plasma current and a is the minor radius. 

than the previous best with deuterium in 1991/92. The highest fusion 

triple product (nDTtE) exceeded 8x1020m"3keVs, within ­10% of the 

previous best. The time evolution of this discharge is shown in 

Fig.51(a), together with its magnetic equilibrium (Fig.51(b)). 

The fusion triple product achieved in a representative selection of 

1994/95 discharges is shown in Fig.52, compared with a similar 

selection from earlier campaigns. 

Advanced Tokamak Studies 

Studies have been undertaken to investigate coherent steady­state 

tokamak concepts (the so­called Advanced Tokamak Scenarios) for 

ITER and DEMO. Advanced Tokamak scenarios for steady­state 

operation require a high bootstrap current fraction (>70%), high 

beta and good Η­mode energy confinement. The broad current density 

profiles typical for high bootstrap current operation are prone to MHD 

instabilities. Active profile control is therefore required. Such a plasma 

would also need to exceed the Troyon ß limit, if very high toroidal fields 

88 

S C I E N T I F IC 

A D V A N C ES 

D U R I NG 

1 9 94 

7 
βΝ = 4/  βΝ = 3/ 

7~ 

-7 

/ 

·/ 
/ 

# 

/ 
/ 
· 

«/ 

Λ 

ι 

/ 

/ 
1 / 
Advanced γ 
Tokamak  j 
Scenarios 
' 
/ 
/  * 

/ 

y 

/ 
Χ 

* 
· 

/*.. 

3.0 

2.5 

2.0 

PP 1.5 

1.0 

0.5 

/  r 

/. 

/ / i/  -fi £ 
//■Hr · · 

/ / / \ · 1994 data 
Λ7/ Δ Typical 1991 

ƒ 
0 

// data 

■195 

10 12 

14 

Fig.53: Achieved values of ßp plotted against the safety factor q95 

were to be avoided. Investigation of the stability and confinement 

characteristics of plasmas in this domain have been undertaken to assess 

the prospects for steady­state reactor concepts. 

In previous experiments, a bootstrap current fraction of 0.7 was 

obtained in plasmas with β ~2 and high confinement compared with 

the usual L­mode and Η­mode scalings. However, these discharges were 

obtained at qg5>10, which would be uneconomic and technically 

difficult to achieve in a reactor. In such plasmas, high values of β can 

be achieved without reaching the Troyon β limit. Nevertheless, these 

discharges were not stable and collapsed with a large ELM after =2s. 

The cause of the collapse has not yet been unambiguously identified. 

The new pumped divertor configuration has several features which 

differ from the old configuration. The plasma volume is smaller and, 

consequently, for similar plasmas current and toroidal field, the safety 

factor, q, is lower. On the other hand, the power handling of the divertor 

has improved. Long ELM­free periods and very high confinement have 

not yet been obtained in high β; experiments with the new configuration. 

While the cause of this is still under investigation, the benefits of ELMy H­

mode plasmas have been exploited to meet the aim of achieving quasi­

steady­state plasmas at least with respect to density control. 

89 

Energy Confinement 

Energy confinement in tokamaks when 

the plasma is bounded by a material 

limiter generally degrades as the input 

power to the plasma increases. The result 

is that the energy confinement time, rp 

falls approximately as the square root of 

the input power. This regime is said to 

exhibit L(low)-mode confinement. In 

plasmas with a magnetic limiter (that is 

with an internal magnetic separatrix or 

X-point), a transition can occur above a 

certain threshold input power to a regime 

in which the energy confinement time is 

increased by a factor of two or more 

greater than in the L-mode situation. 

This has been called H(high)-mode 

confinement. However, a similar 

degradation with input power is 

observed. 

In addition to the improved energy 

confinement time, enhanced particle 

confinement is observed and the 

temperature and density close to the 

separatrix can increase substantially, 

resulting in the formation of plasma 

profiles with an edge 'pedestal'. The 

precise conditions for the transition 

Into the Η-mode vary with plasma 

parameters. For example, the threshold 

power for the transition increases at 

least linearly with the toroidal magnetic 

field. In recent years, the H-mode 

transition has also been observed in 

plasmas with a material limiter, 

although the power threshold is usually 

significantly higher than in magnetic 

limiter (X-point) plasmas. 

S C I E N T I F IC 

A D V A N C ES 

D U R I NG 

1 9 94 

Pulse No: 32847 

Padd (MW) 

15 
10 
5 
0 

NBI 

ICRH 

ne( 0 ) ( x 1 01 9m -3 
l 
18 

L 
20 

Time (s) 

ι 
22 

24 

Fig.54: Time  history  of  heating  and  plasma  parameters  for  a high  β  pulse 

High β  plasmas have been investigated  ir  a much wider  range of 

plasma parameters  during  1994 than in previous JET campaigns. 

Figure 53 shows the achieved values of  β  plotted against the safety 

factor  qg5, and typical data from  previous experiments are also shown 

for  comparison. The  1994 data are significantly  closer to the  'Advanced 

Tokamak'  domain; whereas previous JET data were obtained  at modest 

values of  βΝ, values up to  3.7  have now  been achieved. 

The time evolution  of a typical  high  β  plasma with  a plasma current 

of  1MA and toroidal field of  2.8T  is shown  in Fig.54. The combined 

heating  power of the  neutral beam ICRF heating systems was 20MW, 

which  resulted  in β  value of  2.0.  However,  ir  previous experiments, this 

value had been achieved with  <10MW  of  heating  power  in an ELM-free 

very high confinement  (VH) mode  plasma. The plasma has reached 

quasi-steady conditions with  respect to stored energy, radiated  power 

and density, whereas the previous VH-mode  conditions were achieved 

only transiently. 

The bootstrap current fraction  calculated for  these pulses is about 

0.6, somewhat  lower than  previously. This is largely due to the  higher 

non-thermal  fraction  of the plasma stored energy (not  included  in the 

bootstrap current  calculation), which  results from the  increased 

90 

S C I E N T I F IC 

A D V A N C ES 

D U R I NG 

1 9 94 

Pulse No: 29711 

LHCD Current drive 

9 
8 
Time (s) 

Fig.55: LHCD discharge  at  lp=2MA,Bt=2.8T:  (a)  LH power,  (b)  OH  flux 
consumption, (c) measured, corrected  and  calculated  loop  voltage,  (d) 

central  and  line averaged  electron  density, (e) central  electron 

temperature,  (f)  energy  content  from  diamagnetic  and  equilibrium  field 

measurements  and  (g)  internal  inductance 

additional  heating  power.  In some cases, around  30% of the plasma 

stored energy is estimated  as non-thermal.  The calculated  beam driven 

current can account for  about  20% of the plasma current. 

Three high  power  noninductive  current drive systems are available 

now on JET: lower  hybrid current  drive (LHCD), fast wave current  drive 

(FWCD), and high energy neutral  beam current  drive (NBCD). 

Experiments  have been performed with the full  LHCD system in the 

plasma current  range of  0.7-3MA.  Maximum  LH power  of  6.5MW  and a 

energy  of  36MJ have been coupled to the plasma. A  representative 

2MA  discharge at a line averaged density of <ne>=1.8x10,9irr3  is shown 

in Fig.55. The LH-current  drive with  4MW coupled power  replaces the 

ohmic flux consumption. The measured surface loop voltage  U 

drops 

to zero. The electron temperature  rises from  3 to  5keV and the total 

9  1 

S C I E N T I F IC A D V A N C ES D U R I NG 1 9 94 

• 2.5MA (LHCD + ICRH) 
♦ 1.OMA (LHCD) 

■ 1.5MA (LHCD) 
A 2.0MA (LHCD) 
x 0.4MA (LHCD) 

• 

■ 

• 

• 

A 

• 

• 

• 

0.3 

0.2 

ι 
Q_ 

rr 

Q 
υ 
fr 

0.1 

I I 
0.5 

1.0 1.5 

<Te> (keV) 

I 
2.0 

2.5 

Fig.56: Scaling of LH­current drive efficiency with volume averaged 
electron temperature 

energy content doubles. A decrease of the internal inductance, i\, 

indicates a broadening of the current profile. Full current drive 

conditions have been achieved at plasma currents up to 2.5MA. 

The current drive efficiency, r\CD, increases with electron 

temperature, for LHCD alone and in combined heating/current drive 

scenarios. The scaling of r\CD with the volume averaged temperature 

is shown in Fig.56 for full current drive discharges with LH alone and 

for partial and full current drive in combined operation with LHCD 

and ICRF heating over a wide range of powers. 

Current Profile Control 

A main goal of lower hybrid current drive application on JET is the 

active control of the plasma current profile in order to improve MHD 

stability and energy confinement and to explore scenarios for 

tokamak performance improvement in combination with other 

heating and current drive methods. In a first stage, the extent of 

deposition profile variations and the resulting modifications of the 

current profile were assessed. 

The current profile is broadened in most cases with LHCD, and a 

strong dependence of the LH deposition profile was found with 

92 

S C I E N T I F IC 

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1 9 94 

Pulse No: 33166 

Fig.57: Temporal evolution of LH  power, central ECE emission, ne(0), Te(0), 
internal inductance and q(0) during sawtooth stabilisation with LHCD 

plasma  current.  Stabilisation  of  sawtooth  instabilities was achieved 

with  nearly full  LH current  drive. The waveforms of a 3MA  discharge at 

a density  ne(0)=2x10,9m 3 are shown  in Fig.57. The LH power was 

ramped  up within  1s to  5.5MW.  The last sawtooth was seen on the ECE 

signal  1.5s  after  start of the  LH wave. m=1 oscillations continued  for 

about 0.5s and then disappeared, but only transiently. The central 

electron temperature  rose, with  an increase from  3.8 to  5.4keV. 

Toroidicity-induced  Alfven  Eigenmodes 

Toroidicity-induced  Alfven  Eigenmodes  (TAEs) can be driven  unstable 

by fast  particles  originating  from  neutral  beam  injection, ICRF 

heating  or  fusion  reactions.  Déstabilisation  of  such  modes  in  reactors 

is a subject  of  concern  as they  may  lead to  anomalouly  rapid  losses 

of  energetic  alpha-particles,  reducing  alpha  heating  efficiency  and 

potentially  damaging  the  first  wall.  These  modes  are characterised  by 

high  frequency  (typically  100-200kHz  in JET), which  depends  on  the 

Alfven  speed. 

First experiments  have been carried out,  in which Alfven  eigenmodes 

have been externally excited using the saddle coils driven by a high 

frequency  amplifier.  In ohmic  plasmas, the Alfven  nature of  the 

93 

S C I E N T I F IC A D V A N C ES D U R I NG 1 9 94 

ω 
ω 
co 
­C 
Q. 

bpoi (a.u.) j 4 ^ t ^ H < ( v j. 

1.3 1.4 

f (xlQSHz) 

1.5 160 

Fig.58: (a) Resonance detected at the Alfven frequency when plasma is excited using the Saddle Coils; (b) 

resonant frequencies observed in toroidal field scan compared to calculated Toroidal Alfven Eigenmode 

frequencies; and (c) multipeak structure of resonant frequencies on magnetic and reflectometer data 

above the Alfven frequency 

observed resonances has been verified by scanning the toroidal 

magnetic field and comparing the measured 'esonance frequency 

with that calculated (Fig.58(a and b)). The measured damping of the 

modes (γ/ω in the range 0.1 to 0.01) are being compared with 

predictions. With 6MW of ICRF and 2.5MW of LH powers (T/T>2 

and ß=0.3%), a strong multi­peak structure has been seen on 

magnetic and reflectometer data (Fig.58(c)) at frequencies above the 

continuum damping and with very low damping. This has been 

identified as the first observation of kinetic toroidicity­induced Alfven 

eigenmodes. Experiments will continue in 1995 to observe their 

effect on the plasma. 

N e xt Step Issues 

Certain work has been carried out on issues of particular relevance to 

Next Step devices, especially for ITER. The main highlights are set out 

below. 

ITER Η-mode Database 

Energy confinement predictions for Η­mode operation in ITER require 

a scaling law based on tokamaks of different dimensions. JET has 

94 

S C I E N T I F IC A D V A N C ES D U R I NG 1 9 94 

continued to add new data to the Η-mode database for global 

confinement scaling, throughout 1994, at the request of the ITER 

Project. The work is performed as a combined effort from JET and 

from other tokamaks (Dlll-D (General Atomics, USA), ASDEX (IPP 

Garching, FRG), JFT2M (JAERI, Japan) PBXM and PDX (PPPL, USA)). 

The database now contains measurements from a variety of heated 

H-modes (electron cyclotron resonance, ICRF and neutral beams). 

The scaling of thermal Η-mode confinement using two-term power 

law models has been studied. A linear scaling has been assumed for 

the thermal enerqy content, W,., of the form W =W +τ Ρ, with W , 

^J ' I h' th o inc ' o' 

τιηΕ and Ρ being the offset incremental confinement time and loss 

power, respectively. Using the standard dataset of the H-mode 

database, the following ELM-free and ELMy power laws for Wo and 

Tnchave been obtained from non-linear least squares regression: 

ELM-free: 

Wo=0.0117 l,0B05n075M-025R32K005(a/R)° 

τΓΚ=0.0493 I1 °B-° 3m' °M2 °R-' °κ2 3(a/R)° 

ELMy: 

Wn=0.0055 l ^ ' B ^ n ^ M ^ R4 0^ 5(a/R)03 

τ =0.0851 l,05B'0,n-045M,5R■,'K035(a/R)°3 

ine v ' 

The units are [Wo(MJ), tnc(s), l(MA), B(T), n, M, R(m), k]. 

ITER L-mode Database 

The L-mode database has been expanded by adding more 

information about the experimental configurations, conditioning, 

heating and radiative power, and fast ion energy content. The 

database has also been extended by new data from various devices. 

These changes have made the L-mode database more complete and 

compatible with the ITER Η-mode database. The updated L-mode 

database consists of data from eleven different tokamaks (ASDEX, 

Dill, Dlll-D, FTU, JET, JFT-2M, JT-60, PBX-M, PDX, TORE-SUPRA and 

TEXTOR). 

The scaling analyses of total and thermal confinement times have 

not yet been completed. 

ITER Η-mode Threshold Database 

The Η-mode threshold database has been extended and improved by 

the addition of new data. The database now consists of data from 

95 

S C I E N T I F IC A D V A N C ES D U R I NG 1 9 94 

DAMU13b) 

1.5 

1.0 

„ 0.5 

­ 0 

­0.5 

-1.0 

a) o AlcatorC­Mod 
+ ASDEX 
Δ Dlll­D 
D JET 
■ JFT­2M 
• PBX­M 
O ASDEX­U 
χ Compass­D 

0.5 

1.0 1.5 
log,o (ne°·75 BTS) 

2.0 

2.5 

1.5 

1.0 

0.5 -

­0.5 

-1.0 

o AlcatorC­Mod 
+ ASDEX 
Δ Dlll­D 
D JET 
■ JFT­2M 
• PBX­M 
O ASDEX­U 
χ Compass­D 

1.0 1.5 
log,o (ne BTFt") 

Fig.59: Tests of the variations (a)P~n075BS, and (b) P~nBR25 

ALCATOR C­MOD, ASDEX, ASDEX Upgrade, COMPASS­D, Dlll­D, JET, 

JFT­2M and PBX­M. It has been suggested that the threshold power 

increases linearly with the product of density (n) and toroidal field (B) 

for each device and that the plasma surface area, S, could be used to 

unify the multi­machine database. However, the form P~nBS is not 

dimensionally correct, but it can be made to depend only on 

dimensionless parameters. Two main forms have been investigated: 

P~n075BS (1) 

P~nBR25 (2) 

Figure 59(a) and (b) shows tests of Eqs (1) and (2). The straight 

lines are approximate lower bounds to the threshold power, and are 

given by the expressions P=0.035n075BS (Fig.59(a)) and 0.4nBR2 5 

(Fig.59(b)) in units of 1020nr3, T, m2, m. Assuming n=5x10,9rrr3 for 

ITER, the first scaling yields a threshold power for the Η­mode of 

= 100 MW, while the second gives =200MW. 

Toroidal Field Ripple Experiments 

In 1991, JET carried out an experiment in which only one of two sets 

of toroidal field coils was energised. This inceased the toroidal field 

(TF) ripple at the plasma edge from about 1 % to 10%. The effects of 

high toroidal field ripple on plasma behaviour in general, and on fast 

particle losses, in particular, were studied. Results from these 

experiments were sufficiently encouraging to prepare for a second 

ripple experiment. In this case, the two sets of coils are operated at 

different currents. This modification will make possible the 

exploration of the intermediate ripple regimes, with ripples in the 

96 

S C I E N T I F IC 

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RF window-

RH welds-

dow 
cooling 

Waveguide 
cooling 

Plug/ 
Hyperguide 
cooling 

TEO.3 mode 
converter 

Fig.60:  Schematic  view  of  a  module  of  a  LHCD launcher  design  for  ITER 

range  of  0.5%  to  3 %,  in  the  pumped  divertor  configuration.  The 

first  ripple  experiment  showed  that  at  the  fixed  high  ripple  level  of 

1 0%  the  plasma  was  strongly  degraded.  The  second  ripple 

experiment  should  allow  quantification  and  investigation  of 

threshold  ripple  values  and  other  ripple  effects.  These  experiments 

have  strong  relevance  to  Next  Step  devices,  in  general,  as  there  will 

always  be  engineering  pressure  to  minimise  the  number  of  toroidal 

field  coils.  The  constraints  imposed  by  the  physics  on  this 

minimisation  need  to  be  established. 

Technical  preparation  for  this  experiment  is  now  well  advarced. 

A  study  was  made  of  the  trajectories  of  lost  particles  and  local 

heatloads  on  in-vessel  components  due  to  the  ripple.  As  a  result, 

maximum  input  powers  and  energies  for  the  heating  systems  have 

been  formulated. 

97 

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1 9 94 

LHCD  Launcher  Design  for  ITER 

A conceptual  design for  a Lower  Hybrid (LH)  Heating  and  Current 

Drive  launcher  for  ITER was  made  in cooperation  with  NET and CEA, 

Cadarache,  France.  Lower  hybrid  current  drive  has the  feature  to 

retain  a good  current  drive  efficiency  in  plasma  regions  of  low 

temperature,  or  in  low  β plasmas. These aspects are  important  for 

the  'Advanced  Tokamak'  scenarios. The conceptual  launcher 

operates  at  5GHz and couples  typically  20MW  through  one  main 

horizontal  port.  It  uses a combination  active/passive  waveguide  grill 

with  hyperguide  and  mode  convertors  feeds to  give a structure 

which  combines  a robust  grill  with  simple  supply  networks  and  good 

neutron  attenuation  at the  first  wall  (see Fig.60). 

Summary  of  Scientific  Programme  in  1994 

The first  plasma  in the  pumped  divertor  configuration  was  produced 

in  mid-February  and  by mid-March  successful  2MA  divertor  plasmas 

had  been  established.  Subsequent  achievements  have been: 

• 

the  plasma  current  was  increased  to  5MA,  the  total  heating 

power  to  26MW,  the  stored  energy  to  11.3MJ  and the  neutron 

rate to  4.7x1016 neutrons  per second; 

•  1994 saw significant  progress  in optimising  peak fusion  performance 

(the best neutron  rates of  past campaigns  in deuterium  have now 

been exceeded) and extending operation to the  reactor  relevant 

steady-state  ELMy Η-mode, which  has now  been obtained  under a 

variety of conditions (plasma currents  up to 4MA,  power  levels up to 

26MW,  in the  high β  regime, in discharges with  shear  reversal, and 

at high βΝ).  The high  β  regime was also extended to  steady-state 

and to the  reactor  relevant domain; 

• 

the  high  power  handling  capability  of  the  Mark  I divertor  target 

was demonstrated  and the  severe  impurity  influxes  (carbon 

"blooms"),  which  previously  terminated  high  performance 

plasmas,  have  been  eliminated. About  140MJ  of  neutral  beam 

energy  was  injected  during  ELMy  Η-mode  plasmas,  raising  the 

surface  temperature  of  the  target  tiles  to  less than  550°C; 

• 

the  cryopump  improved vacuum  conditions,  reduced  recycling, 

eliminated  the  effects  of  wall  saturation  (observed  in  previous 

long  pulse  operation),  allowed  effective  particle  control, and 

generally  permitted  higher  performance; 

98 

S C I E N T I F IC 

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• 

the  two  neutral  beam  injectors  routinely  injected  up to  19MW  of 

power  into  plasmas.  133MJ of  power  in a 20s  ELMy H-mode 

represented  a JET record; 

•  up to  13MW  of  ICRF power  was  coupled  to  JET plasmas. This 

power  limit  is set  by a combination  of  unsatisfactory  control 

electronics,  unequal  coupling  of  the  straps of  the  antennae  array 

and  low  power  transfer  to  the  plasma  under  some  phase 

conditions.  New  control  electronics  systems are  being  installed 

and will  be tested  with  plasmas  in early  1995.  Further  antenna 

modifications  are scheduled  for  the  beryllium  tile  exchange  and 

during  the  Mark  II divertor  shutdowns  which  are  planned  for 

mid-1995; 

•  up to  7.3MW,  and  up to  46MJ, of  lower  hybrid  power  has been 

coupled  to  plasmas  using  position  control  feedback.  Up to 

2.5MA  has been  driven  non-inductively; 

• 

the  saddle  coils  have  been  used for  initial  experiments  on TAE 

modes  and  the  disruption  feedback  stabilisation  system  is in  final 

stages  of  commissioning.  Only the  lower  saddle  coils are  now 

available  for  experiments,  since the  upper  saddle  coils  were 

diabled  after  being  damaged  in  September  1994; 

• 

the  prototype  high  speed  pellet  launcher  has been  brought  into 

operation  and  used to  deliver  its first  deuterium  pellets  into 

plasma  in commissioning  pulses. 

Progress  towards  a  Reactor 

An assessment  of  the  fusion  performance  in the  new  pumped 

divertor  configuration  showed  that,  by the  end  of  1994, a D-D 

reactivity  of  7.75x10,6s"'  had  been  achieved  with  an  input  power  of 

18MW  in  a plasma  with  3MA  current  and toroidal  field  of  3.4T.  This 

gives  a fusion  efficiency,  QDD, of  2.6x103,  which  would  correspond  to 

an equivalent  thermonuclear  fusion  efficiency  (if  deuterium  and 

tritium  mixture  had  been  used ) of  QDT -0.6. 

The triple  fusion  product  ( n t j)  obtained  was  7.4x1020nrr3s  keV, 

which  was  about  75%  of  the  previous  best value  obtained  during  the 

1991-92  campaign.  It  is expected  that  further  progress  will  be made 

in  increasing  this  and the  thermonuclear  fusion  efficiency  in  the 

continuing  campaign  during  1995. 

99 

S C I E N T I F IC 

A D V A N C ES 

D U R I NG 

1 9 94 

Year 
t 
ι 

■1994 
I 

­1980 
I 

1000-

σιοο-

Inaccessible 
Region 

Ignition 

' D T =1 

' f *·  JT-60U 

J E T · / J E T· 

JT-30U/4  ,  o  oTFTR 
.TFTR 
»TFTR 

J JW  JET 

TFTR·  D| | | _D.  / 
/ 

• 
J T - 6 0/ 

ALC-C 

<?/ 

TFTR  QDT=0.1 
»Dlll-D 

c 

Ό 
O 

C 
o 
'ω 

W 

| 
Reactor-relevant 
Conditions 

/ D I I I -D 
p\SDEX 

1-

ALC-A 

/ Τ 

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»TFR 
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PLT. 

T10· 

/ 
/ 
TFR^ 

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0.1_ 

T3 

/ 

/ 
/ 
/ 

I— 
10 

- 1— 
100 

­1970 

o D-T Exp 

JGS4.47/1 

1965 

1000 

Central Ion Temperature T¡ (M°C) 

Fig.61: Triple fusion product as a funtion of ion temperature, T,, for a 

number of tokamaks w o r l d w i de 

The carbon bloom, which was a restriction on JET performance in 

the old divertor configuration, has been eliminated by the new 

divertor. However, at high values of the stored energy (W­12MJ), 

the fusion performance is limited now by a variety of MHD 

phenomena (slow rollover, giant ELMs, etc). A comprehensive 

experimental and theoretical study of these events is being 

undertaken in an effort to alleviate the problem. 

During 1994, considerable progress was made on other 

experiments world­wide. In the TFTR machine at PPPL, USA, 

experiments have continued using ­50:50 mixtures of deuterium and 

tritium. A fusion power exceeding 10MW has now been reached 

transiently, with an input beam power of 39MW into a "supershot" 

plasma. The improved yield has been obtained by increasing the 

toroidal magnetic field to 5.5T and the plasma current to 2.7MA. 

In the large Japanese tokamak, JT­60U, the highest fusion product 

(ητΕΤ) of 12.3x1020m3skeV has been achieved at an ion temperature of 

■100­

S C I E N T I F IC 

A D V A N C ES 

D U R I NG 

1 9 94 

41 keV, with  an equivalent  QDT value of  -0.6.  In addition, a quasi steady-

state  ELMy Η-mode was produced  in deuterium with  a high fusion 

product  of 4.5x1020rTT3skeV and an equivalent  QDT of  0.25-0.36. 

The fusion  triple  product  values of  the  high  performance  pulses  in 

both  imoure  deuterium  and  in the  D-T  pulses for  JET and TFTR are 

compared  in  Fig.61  with  the  latest  results from  other  machines 

world-wide  to  illustrate  the  progress  that  has been  made  over  the 

last  30 years. 

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F U T U RE P R O G R A M ME 

Future Programme 

Introduction 

In 1978, the original objectives of JET were set out in the JET Design 

Proposal, EUR-JET-R5, as follows: 

'The essential objective of JET is to obtain and study a plasma in 

conditions and dimensions approaching those needed in a 

thermonuclear reactor. These studies will be aimed at defining the 

parameters, the size and the working conditions of a Tokamak 

reactor. The realisation of this objective involves four main areas 

of work: 

i) the scaling of plasma behaviour as parameters approach the 

Objectives of JET 

The essential objective of JET is to obtain 

and study plasma in conditions and with 

dimensions approaching those needed in 

a thermonuclear reactor. These studies 

will be aimed at: 

1. Scaling of plasma behaviour as 

parameters approach the reactor range; 

2. Plasma-wall interactions in these 

conditions; 

3. Plasma heating; and 

4. Alpha-particle production, confinement 

and consequent plasma heating. 

reactor range; 

ii) the plasma-wall interaction in these conditions; 

Hi) the study of plasma heating; and 

iv) the study of alpha-particle production, confinement and 

consequent plasma heating. 

The problems of plasma-wall interaction and of heating the 

plasma must, in any case, be solved in order to approach the 

conditions of interest. 

An important part of the experimental programme will be to use 

JET to extend to a reactor-like plasma, results obtained and 

innovations made in smaller apparatus as a part of the general 

tokamak programme. These would include: various additional 

heating methods, first wall materials, the control of the plasma 

profiles and plasma formation. ' 

At the start of 1994, JET neared the end of the longest and most 

extensive modifications since initial assembly of the device. During the 

■103-

F U T U RE P R O G R A M ME 

long shutdown, the interior of the vacuum vessel was essentially 

replaced. Following completion, JET was ready to begin its planned 

programme to demonstrate effective methods of power exhaust and 

impurity control in conditions close to those envisaged for ITER. 

Since the beginning of its experimental campaign, extensive studies 

had been made in the first and third areas of work of JET's objectives: 

reactor relevant temperatures (up to 30keV), densities (up to 4x1020nr3) 

and energy confinement times (up to 1.7s) had been achieved in 

separate discharges. The second area of work had been well covered in 

the limiter configuration for which JET was originally designed. 

However, the highest performance JET discharges had been obtained 

with a 'magnetic limiter', (or X-point configuration). The duration of the 

high performance phase of these discharges exceeded 1.5s; this was 

achieved by careful design of the targets and specific operation 

techniques, but is limited, ultimately, by an unacceptably high influx of 

impurities, characterized by a rapid increase in electron density, effective 

ionic discharge and radiated power (referred to as the 'bloom'). 

The fourth area of work had been started by earlier studies of 

energetic particles produced as fusion products or by ion cyclotron 

resonance heating (ICRH). It was addressed further during 1991 by the 

first tokamak plasma experiments in deuterium-tritium mixtures. The 

high performance achieved in deuterium discharges, together with the 

experience gained in making substantial modifications to JET in a 

beryllium environment and with significant vessel activation, gave 

confidence that an experiment with about 10% tritium in the plasma 

could be performed and would provide data that could be used to plan 

an effective campaign of deuterium-tritium experiments in 1996. 

During 1991, the JET Council had approved the policy of a step-wise 

approach to the introduction of tritium in advance of the full D-T phase 

of JET operations. As a first such step, after having obtained all 

necessary regulatory approvals, JET successfully carried out a preliminary 

tritium experiment (PTE-1) in November 1991. A release of fusion 

energy in the megawatt range in a controlled fusion device had been 

achieved for the first time in the world. 

In the 1991/92 campaign, JET achieved plasma parameters 

approaching breakeven values for about a second, resulting in large 

bursts of neutrons. However, in spite of the plasma pulse continuing 

■ 1 0 4-

F U T U RE P R O G R A M ME 

for many seconds after reaching peak plasma values, the neutron count 

fell away rapidly as impurities entered the plasma and lowered its 

performance. This limitation on the time for which the near­breakeven 

conditions could be maintained was due to the poisoning of the plasma 

by impu'ities (the 'bloom'). This further emphasised the need to provide 

a scheme of impurity control suitable for a Next Step device. 

In late 1991, the Council of Ministers approved a modification to the 

JET Statutes, which prolonged its statutory lifetime by four years until 

31st December 1996. The extension will allow JET to implement the 

new Pumped Divertor Phase of operation, the objective of which is to 

establish the effective control of plasma impurities in operating 

conditions close to those of the Next Step. This programme of studies 

will be pursued before the final phase of full D­T operations in JET. 

During 1992/93, a large effort was devoted to shutdown work for 

the pumped divertor phase. The first stage of the shutdown in 1992 

had involved replacement of faulty toroidal magnetic field coils. The 

second stage in 1992/93 involved assembly of the four divertor coils and 

casings inside the vacuum vessel. The third stage began in mid­1993, 

with the final positioning of the coils. The shutdown was successfully 

completed in January 1994. The first plasma in the Pumped Divertor 

Characterisation Phase was produced in mid­February and by mid­

March 2MA divertor plasmas had been established. During 1994, the 

plasma current was increased to 5MA, the heating power to 26MW, the 

stored energy to 11.3MJ and the neutron rate to 4x1016 neutrons/s. 

199¿ saw significant progress in optimising peak fusion 

performance and extending operation to the reactor relevant steady­

state ELMy Η­mode, which has now been obtained under a variety of 

conditions (plasma currents up to 4MA, power levels up to 26MW, in 

the high ßp regime, in discharges with negative central magnetic 

shear, and at high ßM). The high ß_ regime has also been extended to 

steady­state and to the reactor relevant domain. 

The high power handling capability of the Mark I divertor target was 

demonstrated and the severe impurity influxes (carbon "blooms"), 

which previously terminated high performance plasmas, have been 

eliminated. The cryopump reduces recycling, eliminates the effects of 

wall saturation (observed in previous long pulse operation), allows 

effective particle control, and generally allows higher performance. 

■105­

F U T U RE P R O G R A M ME 

JET Strategy 

These achievements show that the main objectives of JET are being 

actively addressed. The overall aim can be summarised as a strategy 

"to optimise the fusion product (ηΤτΕ) For the energy confinement 

time, tE, this involves maintaining, with full additional heating, the 

values that have already been reached. For the density and ion 

temperature, it means increasing their central values n(0) and T(0) to 

such an extent that D-T operation would produce alpha-particles in 

sufficient quantities to be able to analyse their effects on the plasma. 

JET aims to build up high density and high temperature plasma in 

the centre of the discharge (with minimum impurity levels) where alpha-

particles could be observed, while maintaining an acceptably high global 

energy confinement time tE. The mechanisms involved are to decouple 

the temperature profile from the current density profile using lower 

hybrid current drive and neutral beam injection to ensure that, at higher 

central temperatures, the current density in the centre does not reach 

the critical value that causes sawteeth oscillations. 

This involves the following: 

a) Increasing the Central Deuterium Density nD(0) by: 

• injecting high energy deuterium beams to fuel the plasma 

centre and dilute impurities; 

• injecting pellets to control the influx of edge material; 

• stabilising the m=2, n=1 magnetic oscillations present at the 

onset of a disruption with magnetic perturbations produced 

from a set of feedback controlled internal saddle coils; 

b) Increasing the Central Ion Temperature, T(0) by: 

• trying to lengthen the sawtooth period; 

• controlling the current profile (by lower hybrid current drive in 

the outer region, and by counter neutral beam injection near 

the centre) to flatten the profile; 

• on-axis heating using the full NB and ICRF additional heating 

power (24MW, ICRF, and 20MW, NB) 

c) Increasing the Energy Confinement time τΕ by: 

• increasing to 6MA the plasma current in full power, H-mode 

operation in the X-point configuration; 

d) Reducing the impurity content, by: 

• using beryllium as a first-wall material to decrease the impurities; 

■106-

F U T U RE P R O G R A M ME 

• controlling impurities and edge material using the pumped 

divertor configuration. 

In parallel, preparations for the D-T phase of operations have 

continued. JET has completed installation of the main components of the 

active gas handling system and pre-tritium commissioning has continued. 

During 1994, the inactive commissioning phase of subsystems continued 

in accordance with the programme for D-T operations in 1996. 

JET is now continuing its programme to demonstrate effective 

methods of power exhaust and impurity control in operational 

conditions close to those envisaged for ITER. ITER relevant studies will 

provide stimulation to JET and JET's results will make an important 

contribution to development of the ITER design. 

Future Plans 

The JET Programme was divided into phases governed by the availability 

of new equipment and fitting within the accepted life time of the 

Project. Phase I (Ohmic Heating Studies) was completed in September 

1984, and Phase II (Additional Heating Studies) in October 1988. Phase 

III (Full Power Optimization Studies) ended in February 1992. The 

scientific aims of Phase III were to obtain maximum performance in 

limiter configuration (currents up to 7MA) and to optimize X-Point 

Opera:ion (currents up to 6MA) including a comparison of H-modes in 

X-point configuration using beryllium (loWer X-point) with carbon 

(upper X-point) dump plates. 

JET future plans are dominated by the new phase of the Project 

(Phase IV: Pumped Divertor Configuration and Next Step Oriented 

Studies). This phase is subdivided into a Divertor Characterization 

Plasma and an ITER Support Phase, and extends the lifetime of :he 

Project to the end of 1996. 

The present campaign on divertor characterization (to end in May 

1995) has still to address: high power combined heating; further 

exploration of the detached divertor regime and its compatibility with 

ELMy Η-mode operation; helium pumping and transport using argon 

frost on the cryopump; current profile control; toroidal magnetic field 

ripple studies; use of the saddle coils with the disruption feedback 

contro 1er and for generating error fields; and the extension of X-point 

operat on towards 6MA. In March 1995, the present CFC divertor target 

■107-

F U T U RE P R O G R A M ME 

1991 

1992 

1993 

1994 

1995 

1996 

1997 

1998 

1999 

Pumped Divertor 
Characterisation 
Phase 

PTE 

ITER Support Programme 

ITER­EDA Support Phase . Final D­T Phase ' 

1 
I 

Divertor Concept 'Characterisation 
Studiesand ¡of ITER­specific 

DTE1 Divertor 

D T E 1, 

I 

I 

Plasma ι 
Optimisation and ι 
DTE2 I 

DTE 2, 

ι ι 

'T 
I 

• Pumped Divertor 

(Mark I) 

■ Divertor Modifications 
(Mark IIA) 

■ Divertor Modifications 
(Mark II—GB) using 
Remote Handling 

• Preparations for DTE 1 

1 Modifications to Divertor, 
if necessary 

1 Preparations for DTE 2 

JG94.26/1 (Rev.4.5) 

PTE: Preliminary Tritium Experiment 
DTE 1 : D­T Experiments with < 2 χ 1020 neutrons 
DTE 2: D­T Experiments with < 5 χ 1021 neutrons 

Fig.62: The JET P r o g r a m me S c h e d u le to 1999 

tiles will be exchanged for beryllium target tiles and a comparison will 

be made for a few specific plasma configurations. 

The next major milestone is then to demonstrate compatibility 

between the high confinement reactor­relevant ELMy Η­mode regime 

and detached divertor operation. This might require the more closed 

Mark II divertor structure, which will be installed in 1995 and tested 

in 1996 with particular emphasis on the effect of geometry on gas 

target/radiative divertor plasmas, which forms the physics basis for 

the divertor concept favoured by the ITER Joint Central Team. 

Extension of the Programme to end of 1999 

An extension of the JET Programme to the end of 1999 is currently being 

proposed. The purpose of the extension would be to provide further data 

of direct relevance to ITER, especially for the ITER­EDA, before entering 

into a final phase of D­T operation. In particular, the extension would: 

(i) make essential contributions to the development and 

demonstration of a viable divertor concept for ITER, and 

(ii) carry out experiments using deuterium­tritium plasmas in an ITER­

like configuration, which will provide a firm basis for the D­T 

operation of ITER; 

while allowing key ITER­relevant technology activités, such as 

demonstration of remote handling and tritium handling, to be 

carried out. 

Extending JET by three years to the end of 1999 would allow a 

three year ITER­EDA Support Phase to be inserted into the JET 

­108­

F U T U RE P R O G R A M ME 

Programme before the final D-T phase begins. The proposed 

programme to the end of 1999 is shown in Fig.62. 

More information on the future phases are indicated below. 

Pumped Divertor Characterisation Phase 
(1992 to mid-May 1995) 

The shutdown begun in February 1992 was completed in 

January 1994. Experiments during the 1994/5 are focused on: 

establishing reliable operation in the new configuration; 

- studying control of impurities, plasma density and exhaust; 

assessing power handling using the full range of ancillary 

equipment; 

- extending performance to high power, long pulse operation; 

studying specific physics and ITER related issues. 

The relatively open geometry of the Mark I Divertor accepts a wide 

range of plasma equilibria, and divertor studies are possible for a range 

of configurations within the designed maximum plasma current 

capab lity of 6MA. Divertor pumping and exhaust issues, particularly 

helium exhaust, are being addressed with the torus cryopump, which 

has already shown clear benefits for plasma purity and density control. 

Carbon fibre composite divertor target plate tiles were installed for 

initial operations with the Mark I Divertor and will be used for most of 

the campaign. During a short intervention in March/April 1995, these will 

be replaced by beryllium tiles. A comparison will then be made between 

CFC and beryllium target tiles for a few specific plasma configurations. 

ITER-EDA Support Phase 
(mid-1995 to mid-1998) 

The next major shutdown is scheduled to start in May 1995 and to be 

completed in 9 months. As agreed by the JET Council, the Mark IIA 

Divertor will be installed. This will have a considerably higher unswept 

power handling capability than the Mark I divertor, particularly on the 

vertical side plates. The new divertor configuration will use close-fitting, 

precisely-aligned large target tiles which result in a much larger plasma 

"footprint" on the targets. The Mark IIA divertor target will also be 

more "closed", facilitating the production of a low temperature, high 

density, high recycling, radiative divertor plasma in which atomic 

processes (such as radiation and charge-exchanged neutral losses) 

■109-

F U T U RE P R O G R A M ME 

reduce the conducted power to the targets. During the shutdown, 

work will also be undertaken to bring all JET systems and sub-systems to 

a level of tritium compatibility adequate for the D-T operations planned 

for 1996. This will include a number of modifications to strengthen or 

remove components in order to ensure the reliability of the machine. 

During 1996, the programme will develop toward long pulse, high 

performance operation. The Mark IIA divertor target structure can 

accommodate a wide range of plasma configurations and operating 

conditions, making possible high power, high current operation on both 

the horizontal and vertical target plates. The programme will extend 

many of the issues now being addressed initially with the Mark I divertor 

and give further emphasis to studies of the effect of geometry on gas 

target/detached plasmas which form the physics basis for the divertor 

concept favoured by the ITER Joint Central Team. 

A period of D-T operation (DTE-1) is scheduled for the last four 

months of 1996. It will assess whether the more favourable 

confinement found during D-T operation in TFTR extends also to the 

ITER-relevant divertor and operating conditions in JET. These 

experiments will allow more accurate scalings for the size and heating 

requirements of ITER to be determined. In addition, it will demonstrate 

long pulse fusion power production (fusion amplification factor, Q=1 

with more than 10MW of fusion power). As a result, DTE-1 could last 

up to four months and produce up to 2 χ 1020 neutrons. DTE-1 will also 

address the important technology issues of reactor relevant tritium 

processing and remote handling. In particular, DTE-1 will demonstrate 

the ability of the Active Gas Handling System to process tritium while 

supporting a reacting tokamak plasma. 

In a six-month shutdown in early-1997, the Mark IIA divertor target 

structure will be exchanged for a second target structure, an 

ITER-specific divertor of the "Gas-Box" class (Mark IIGB).The exchange 

will be accomplished by remote handling since radiation levels will be 

too high during 1997 for a manned in-vessel intervention. This remote 

handling operation will demonstrate for the first time one of the central 

technologies required for ITER and for a fusion reactor. 

In a divertor of the "Gas-Box" type, such as the Mark IIGB, energy 

and momentum are removed from the divertor plasma and spread more 

uniformly over the divertor sidewalls. Neutrals can recirculate freely 

■1 1 0-

F U T U RE 

P R O G R A M ME 

through a large relatively open volume below a fairly  narrow  entrance 

baffle, which  is placed as high as possible near the X-point. The  1997/98 

operations  period will validate experimentally  high power,  high 

performance  physics in a closed divertor  configuration  similar to that 

currently  proposed for ITER. 

Final  Phase  of  D-T  Operation 
(mid-1998  to  end-1999) 

A four-month  shutdown  in  1998 will  permit  any  necessary 

modifications  to  the  divertor  and final  preparations  for  DTE-2. 

Manned  interventions  will  again  be possible  in this  shutdown.  The 

subsequent  experimental  programme  in  late  1998 and early  1999 

will  then  prepare for  the  final  phase  of  D-T operation  by  optimising 

plasma  performance  and  establishing  reliable  operation. 

A period of  high  performance  D-T operation  (DTE-2) is scheduled to 

take place in  1999. This will capitalise on performance  improvements 

achieved in the preceding experimental campaigns with  deuterium. 

DTE-2 experiments could last up to eight  months and could produce up 

to  5 χ 102' neutrons. Actual  neutron  production, within  this defined 

limit, will  be reassessed in light  of the experience gained on JET (DTE-1 

in  1996) and on TFTR (in  1994/95). Every effort will  be made to  reduce 

the activation  produced while still satisfying JET's role in supporting ITER 

and the World  Fusion Programme. This period of  D-T operation will also 

provide a full-scale test of  the technology  of  processing tritium  in 

conjunction with  an operating  tokamak. 

- 1 1 1-

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M E M B E RS A ND O R G A N I S A T I ON 

Members and Organisation 

Members 

The JET Joint Undertaking has the following Members: 

The European Atomic Energy Community (EURATOM); 

The Belgian State, acting for its own part ('Laboratoire de Physique 

des Plasmas de l'École Royale Militaire ­ Laboratorium voor 

plasma­physica van de Koninklijke Militaire School') and on behalf 

of the Université Libre de Bruxelles' ('Service de physique 

statistique, plasmas et optique non­linéaire de l'ULB'); and of the 

'Centre d'Études de l'Energie Nucléaire (CEN)/'Studiecentrum voor 

Kernenergie' (SCK); 

The Centra de Investigaciones Energéticas Medioambientales y 

Tecnológicas (CIEMAT), Spain; 

The Commissariat à l'Énergie Atomique (CEA), France; 

The 'Ente per le Nuova Tecnologie, l'Energia e l'Ambiente' ('ENEA') 

representing all Italian activities falling within the Euratom Fusion 

Programme including that of the 'Consiglio Nazionale delle 

Ricerche', (CNR); 

The Hellenic Republic, Greece; 

The Forskningscenter Risø (Risø), Denmark; 

The Grand Duchy of Luxembourg, Luxembourg; 

The Junta Nacional de Investigação Cientifica e Tecnológica (JNICT), 

Portugal; 

Ireland; 

The Forschungszentrum Jülich GmbH(KFA), Germany; 

The Max­Planck­Gesellschaft zur Förderung der Wissenschaften 

e.V. Institut für Plasmaphysik (IPP), Germany; 

■1 1 3­

M E M B E RS A ND O R G A N I S A T I ON 

COUNCIL 

Chairman 

H. von BÜLOW 

Γ 

EXECUTIVE COMMITTEE 

Chairman 
F. TURVEY 

1 
SCIENTIFIC COUNCIL 

Chairman 
K. LACKNER 

DIRECTOR 

M. KEILHACKER 

Associate Director, 
Head of Administration 
Department 

Deputy Director, 
Head of Torus and 
Measurements Department 

Associate Director. 
Head of Heating and 
Operations Department 

G.W. O'HARA 

A. GIBSON 

J. JACQUINOT 

Fig.63: Overall Project Structure 

The Swedish Natural Science Research Council (NFR), Sweden; 

The Swiss Confederation, Switzerland; 

The Stichting voor Fundamenteel Onderzoek der Materie (FOM), 

The Netherlands; 

The United Kingdom Atomic Energy Authority (UKAEA), Host 

Organisation. 

Management 

The JET Joint Undertaking is governed by Statutes which were adopted 

by the Council of the European Communities on 30 May 1978. The 

organs of the Joint Undertaking are the JET Council and the Director of 

the Project. The JET Council is assisted by the JET Executive Committee 

and is advised by the JET Scientific Council (see Fig.63). 

JET Council 

Each member of the Joint Undertaking is represented on the JET 

Council, which is required to meet at least twice yearly. The Council is 

responsible for the management of the Joint Undertaking and for: 

• Nomination of the Director and Senior Staff of the Project 

with a view to their appointment by the Commission or the 

Host Organisation as appropriate; 

• Approval of the annual budget, including staffing, as well as 

the Project Development Plan and the Project Cost Estimates; 

­1 1 4­

M E M B E RS A ND O R G A N I S A T I ON 

• Ensuring collaboration between the Associated Laboratories 

and the Joint Undertaking in the execution of the Project, 

including the establishment of rules on the operation and 

exploitation of JET. 

Four meetings of the JET Council were held during the year: on 7th 

February, 24th March, 6th-7th July and 12th-13th October 1994. The 

membership of the JET Council is shown in Appendix I. 

JET Executive Committee 

The JET Executive Committee is required to meet at least six times a 

year. Its functions include: 

• Advising the JET Council and the Director of the Project on 

the status of the Project on the basis of regular reports; 

• Commenting and making recommendations to the JET 

Council on the Project Cost Estimates and the Draft Budget, 

including the establishment of staff, drawn up by the Director 

of the Project; 

• Approving, in accordance with the rules on the award of 

contracts established by the JET Council, the tendering 

procedure and the award of contracts; 

• Promoting and developing collaboration between the 

Associated Laboratories and the Joint Undertaking in the 

execution of the Project. 

The membership of the JET Executive Committee is shown in 

Appendix II. The Committee met six times during the year: on 10th-

11th February, 21st April, 14th July, 1 5th-16th September, 27th-

28th October and 8th-9th December 1994. 

JET Scientific Council 

The Statutes confer the following functions on the JET Scientific Council: 

• Upon the request of the JET Council, to advise on scientific 

and technical matters, including proposals involving a 

significant change in the design of JET, its exploitation, and its 

long-term scientific implications; 

• To perform such other tasks as the JET Council may request it 

to undertake. 

The Scientific Council met three times during the year: on 8th-9th 

February, 14th - 1 5th June and 21st - 22nd September. It also 

■1 ι 5-

M E M B E RS A ND O R G A N I S A T I ON 

organised a one day workshop, which was also attended by other 

experts from the Associated Laboratories, to examine: 

• how the JET programme can best satisfy the divertor research 

needs of ITER. 

The JET SC Chairman reported to the JET Council on three occasions, on: 

• the JET Director's proposal to extend the JET Programme 

beyond 1996 (the Mark II divertor programme; options for D-T 

experiments including the scope and strategic importance of 

DTE-1; and enhanced plasma performance and coherent 

advanced tokamak concepts); 

• the 1994 Experimental Campaign with the Mark I divertor; and 

• the reliability of the JET device for future operations 

(following the recommendations of a Joint Assessment Group 

involving JET Scientific Council members and JET staff). 

The full Scientific Council membership is detailed in Appendix III. 

Host Organisation 

The United Kingdom Atomic Energy Authority, as the Host Organisation 

for the JET Joint Undertaking, has made available to the Joint 

Undertaking, the land, buildings, goods and services required for the 

implementation of the Project. The details of such support, as well as 

the procedures for co-operation between the Joint Undertaking and the 

Host Organisation, are covered by a 'Support Agreement' between both 

parties. In addition to providing staff to the JET team, the Host 

Organisation provides support staff and services, at proven cost, to meet 

the requirements of the JET Project. 

Project Team Structure 
The Director of the Project 

The Director of the Project, Dr. M. Keilhacker, is the chief executive 

of the Joint Undertaking and its legal representative. He is 

responsible to the JET Council for the execution of the Project 

Development Plan, which specifies the programme, and for the 

execution of all elements of the Project. The Project Development 

Plan covers the whole term of the Joint Undertaking and is regularly 

updated. The Director is also required to provide the JET Council and 

other subsidiary bodies with all information necessary for the 

performance of their functions. 

■1 1 6-

M E M B E RS 

A ND 

O R G A N I S A T I ON 

Internal  Organisation 

The internal  organisation  of  the  Project  consists of  three 

Departments  and  the  Directorate.  The three  Departments  are: 

•  Torus and  Measurements  Department; 

•  Heating  and  Operations  Department; 

•  Administration  Department. 

The  Project  Departmental  and  Divisional  structure  is shown  in Fig.64. 

Directorate 

The Heads of  the Departments  report  to the  Director of  the Project and 

together  with  the  Director form the JET Directorate. Various special 

functions are carried out  by the  Director's Office. The Internal  Audit 

Office  monitors the financial activities and provides advice on 

accounting and control  procedures as well as maintaining  links with the 

Court  of Auditors. The Project Control  Office  is responsible for  financial 

planning and for  the preparation  of the  Project  Development  Plan and 

Project  Cost  Estimates. The JET Council Secretariat  provides  Secretarial 

Services to the JET Council and to the Executive Committee and also to 

the JET Project Board. 

Directorate 

Co-ordinating Staff  Unit:  J-P  Pofte 

Data Analysis and Modelling Unit:  J.G.Cordey 

Next Step  Unit:  B.Tubbing 

Torus  & Measurements  Department 

X 

DIRECTOR: 
M.Keilhacker 

Deputy  Director: 
A.Gibson 

Chief  Engineer: 
E.Bertolini 

Heating & Operations  Department 

Administration  Department 

Associate  Director & Head of  Department: 
A.Gibson 
Deputy  Head of  Department: 
R.Haange 

Associate Director  & Head of  Department: 
J.Jacquinot 
Deputy Head of  Department: 
E.Bertolini 

Associate  Directors  Head of  Department: 

G.W.O 

Hara 

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Fig.64:  JET Departmental  and  Divisional  Structure 

■1 1 7-

M E M B E RS 

A ND 

O R G A N I S A T I ON 

Within  the  Directorate,  are four  technical  units  and  a Chief 

Engineer,  reporting  directly  to  the  Director.  The  main  responsibilities 

are as follows: 

(a)  The Co-ordinating  Staff  Unit  is responsible  for  the  availability  of a 

comprehensive  health  physics and  safety  project  organisation; 

and for  the  provision  of  centralised  engineering  support  services. 

It  comprises  four  Groups:  Health  Physics and  Safety  Group; 

Quality  Group; Technical  Services  Group;  and  Drawing  Office. 

(b)  The Data Analysis and  Modelling  Unit  is responsible  for  the 

provision  of  software  for  the  acquisition  and  processing  of  data 

from  JET diagnostics;  for  confirming  the  internal  consistency  of 

the  processed  data  and assembling  it  into  public  databases;  and 

the  development  and testing  of  theoretical  models  against JET 

data.  In addition, the  Unit  is responsible for  prediction  by 

computer  simulation  of  JET performance,  interpretation  of JET 

data  and the  application  of  analytic  plasma  theory  to  gain  an 

understanding  of  JET physics.  It comprises  three  groups:  Analytic 

Theory  Group;  Simulation  Group;  and  Data  Processing  and 

Analysis  Group 

(c)  The Next Step Unit  is responsible for  co-ordinating  contributions 

from  JET to  the  European  effort  in support  of  the  ITER-EDA. This 

responsibility  includes  drawing  up  proposals,  initiating  relevant 

work  programmes  on JET and taking  part  in their  execution  and 

evaluation. 

(d)  The Divertor Interface Unit is responsible for  assessing the impact of 

developments  in the experimental  programme and operation  on the 

design  requirements for JET divertors. This includes a high  level of 

participation  in the experimental  programme on divertor physics, 

thermomechanical  analysis of  plasma  induced  loads on the  divertor, 

and the definition  of  advanced divertor concepts. 

In addition,  there  is a Chief Engineer, who  reports  to  the  Director, 

and  is responsible  for  ensuring  the  overall  coherence  of  technical 

aspects  of  JET operations. 

Torus  and  Measurements  Department 

The Torus and Measurements  Department  has overall  responsibility  for 

the  performance  capacity of  the  machine: this  ncludes  enhancements 

directly  related to this (excluding  heating) and the  long term  planning 

-1  1 8-

M E M B E RS A ND O R G A N I S A T I ON 

associated with integration of these elements to achieve ultimate 

performance. The Department is also responsible: for fusion technology 

requirements for the active phase including tritium handling and 

processing; for construction and operation of necessary measurement 

diagnostic systems and the interpretation of experiment data; and for 

data systems comprising data control, acquisition and management. The 

main functions of the Department are: 

• to design, procure and implement enhancements to the JET 

device; 

• to provide and maintain clean conditions inside the vessel which 

lead to high quality plasma discharges; 

• to conceive and define a set of coherent measurements; 

• to be responsible for construction of necessary diagnostics; 

• to be responsible for diagnostics operation, quality of 

measurements and definition of plasma parameters; 

• to organise and implement data acquisition and computing; 

• to design and develop remote handling methods and tools to 

cope with JET requirements; 

• to design and construct facilities for handling tritium and for 

waste management. 

The Department consists of five Divisions and three Groups 

(Machine Assembly, Diagnostic Engineering and Data Management): 

(a) First Wall Division, which is responsible for the vital area of 

plasma wall interactions. Its main tasks include the provision and 

maintenance inside the vacuum vessel of conditions leading to 

high quality plasma discharges. The Division develops, designs, 

procures and installs the first wall systems and its components 

sucn as limiters, wall protections and internal pumping dev ces. 

The area of responsibility encompasses the mechanical integrity 

of the vacuum vessel as a whole and the development and 

implementation of mechanical and remote handling techniques; 

(b) Fusion Technology Division, is responsible for all nuclear engineering 

aspects of the Project including tritium and gas handling, vacuum 

systems, waste management and regulatory approvals; 

(c) Control and Data Acquisition System Division (CODAS), which is 

responsible for the implementation, upgrading and operation of 

computer-based control and data acquisition systems for JET; 

■1 1 9-

M E M B E RS 

A ND 

O R G A N I S A T I ON 

(d)  Experimental  Division  I  (EDI),  which  is responsible  for 

specification,  procurement  and  operation  of  about  half  the JET 

diagnostic  systems.  ED1  undertakes  electrical  measurements, 

electron  temperature  measurements,  surface  and  limiter  physics 

and  neutron  diagnostics; 

(e)  Experimental  Division 2  (ED2),  which  is responsible  for 

specification,  procurement  and  operation  of  the  other  half  of  the 

JET diagnostic  systems.  ED2 undertakes  all  spectroscopic 

diagnostics,  bolometry,  ¡nterferometry,  the  soft  X-ray  and  neutral 

particle analysis. 

Heating  and  Operations  Department 

The overall  responsibility  of  the  Heating  and  Operations  Department 

is for  the  efficient  and effective  day-to-day  operation  of  the  machine. 

In addition, the  Department  has responsibility  for  plasma  heating  and 

auxiliary  equipment  and  related  physics; the  design  and operation  of 

power  supplies  as well  as contributing  to  the  execution  and 

evaluation  of  JET's experimental  programme.  The  main  functions  of 

the  Department  are: 

•  preparing  and  co-ordinating  operation  of  the  machine  across 

Departments  and  Divisions; 

•  heating  and  current  drive  and  analysis  of  its effects  in the 

plasma; 

•  plasma  fuelling,  including  pellet  injection; 

•  designing  and employing  power  supplies for  ensuring  efficient 

operation  and control  of  the  machine. 

The  Department  consist  of  two  Groups  (Machine  Operations  and 

Machine  Services)  and four  Divisions: 

(a)  Operations  Division plays a major  role  in the  efficient  planning 

and  execution  of  JET's experimental  programme  and  in the 

integration  of  existing  or  imminent  systems  into  an  effective 

experimental  programme.  In addition,  it  is responsible  for 

effective  methods  of  fuelling  the  plasma  including  the 

development  of  methods  based on  solid  high  speed  hydrogen 

pellets;  development  of  new  plasma  wall  conditioning 

techniques;  plasma  control  systems; development  of  disruption 

control  methods;  training  of  operations  staff;  and  monitoring  of 

machine  operations; 

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M E M B E RS 

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O R G A N I S A T I ON 

(b)  Neutral  Beam Heating  Division, which  is responsible  for 

construction,  installation,  commissioning  and operation  of  the 

neutral  injection  system, including  development  towards  full 

power  operation. The  Division  is also  responsible  for  all cryo-

systems and also  participates  in studies  of  physics of  neutral 

beam  heating; 

(c)  Radio Frequency Heating  Division, which  is responsible  for  the 

design,  construction, commissioning  and operating  RF heating 

and current  drive  systems during  the  different  stages of  its 

development  to  full  power.  The  Division  is also  responsible  for  the 

TAE excitation  system  and also  participates  in studies  of  the 

physics  of  RF heating; 

(d)  Magnet  and  Power Supplies Division is responsible  for  the  design, 

construction,  installation, operation  and  maintenance  of  the 

electromagnetic  system  and  plasma  control. The area  of 

responsibility  encompasses  the  toroidal,  poloidal  and  divertor 

magnets,  mechanical  structure;  and all  power  supply  equipment 

needed for  magnets,  plasma  control, additional  heating  and 

auxiliaries. 

Administration  Department 

The Administration  Department  is responsible  for  providing 

Contracts,  Finance  and  Personnel  services to  the  Project.  In addition, 

the  Department  is responsible  for  the  administration  of  Publications 

and  Public  Relations  Groups. 

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A D M I N I S T R A T I ON 

Administration 

Introduction 

The three main aspects of JET's administration - Finance, Contracts and 

Personnel - are reported in this section. There are also contributions on 

Safety and Health Physics, Public Relations and Publications Group. 

Finance 

The initial budgets for 1994 were approved at 81.70 MioECU for 

Commitments and 100.16 MioECU for both Income and Payments, 

subject to a 10.0 MioECU blockage in all Budgets only to be released 

on the agreement of the JET Council. As a result of the efforts made 

to achieve longer term reductions in annual costs, the JET Director 

did not need to request release of the 10.0 MioECU. Therefore, the 

final 1994 budgets were 71.70 MioECU for Commitments and 

90.16 MioECU for both Income and Payments. 

Due to the temporary blockage by the European Parliament of 

part of the Community Fusion Programme Budget for 1994, the 

original budget could not be adopted until August. In the meantime, 

the Project operated within a series of constrained budgets and 

rigorous expenditure restrictions were enforced. 

The Commitments and Payments Budgets each are divided irto two 

phases of the Project - Extension to Full Performance and the 

Operational Phase; subdivisions distinguish between investment, 

operating, and personnel costs, each with further detailed cost codes. 

Commitments 

Of the total final appropriations in 1994 of 89.48 MioECU (including 

1 7.78 MioECU brought forward from previous years), 79.36 MioECU 

■1 2 3-

A D M I N I S T R A T I ON 

COMMITMENT APPROPRIATIONS 

MioECU 

INITIAL COMMITMENTS BUDGET FOR 1994 

AMOUNTS BROUGHT FORWARD FROM 

PREVIOUS YEARS. 

COMMITMENTS MADE DURING THE YEAR 

BALANCE OF APPROPRIATIONS AT 31 DECEMBER 

1994 AVAILABLE FOR USE IN 1995 

Table 7: Commitment Appropriations for 1994 

71.70 

17.78 

89.48 

79.36 

10.12 

was committed and the balance of 10.12 MioECU was available for 

carrying forward to 1995. The details of the commitment 

appropriations available (Table 7) and of the amounts committed in 

each Phase during the year (Table 8) are summarised as follows: 

• In the extension to Full Performance Phase, 0.62 MioECU was 

committed leaving 1.13 MioECU commitment appropriations not 

utilised at 31 December 1994, to be carried forward to 1995; 

• In the Operational Phase, 78.74 MioECU was committed leaving 

a balance of 8.99 MioECU to be carried forward to 1995. 

BUDGET HEADING 

COMMITMENTS 

PAYMENTS 

BUDGET 
APPRO­
PRIATIONS 
MioECU 

OUTTURN 

MioECU 

BUDGET 
APPRO­
PRIATIONS 
MioECU 

OUTTURN 

MioECU 

PHASE 2 

TITLE 1 

EXTENSION TO FULL 
PERFORMANCE 
PROJECT INVESTMENTS 

PHASE 3 
TITLE 1 
TITLE 2 
TITLE 3 
TOTAL PHASE 3 

OPERATIONAL 
PROJECT INVESTMENTS 
OPERATING COSTS 
PERSONNEL COSTS 

1.75 

0.62 

3.00 

2.73 

5.55 
33.92 
48.26 
87.73 

4.81 
28.82 
45.11 
78.74 

8.42 
35.78 
48.31 
92.51 

7.14 
34.17 
46.19 
87.50 

PROJECT TOTAL ­ ALL PHASES 

89.48 

79.36 

95.51 

90.23 

Table 8: Commitments and Payments for 1994 

■124­

A D M I N I S T R A T I ON 

INCOME  AND  PAYMENTS 

MioECU 

INCOME 

BUDGET FOR  1994 

INCOME  RECEIVED  DURING  1994 

(I)  MEMBERS'  CONTRIBUTIONS 

(II)  BANK  INTEREST 

(III)  MISCELLANEOUS 

(IV)  UNUSED APPROPRIATIONS  BROUGHT 

FORWARD  FROM  PREVIOUS YEARS 

TOTAL  INCOME 

VARIATION  FROM  BUDGET 

PAYMENTS 

BUDGET FOR  1994 

90.16 

90.16 
-

81.67 

0.95 

0.02 

7.52 

90.16 

AMOUNTS  AVAILABLE  IN THE SPECIAL ACCOUNT TO  MEET 

OUTSTANDING  COMMITMENTS  AT  31  DECEMBER  1993. 

5.35 

TOTAL AVAILABLE  APPROPRIATIONS  FOR  1994 

95.51 

ACTUAL  PAYMENTS  DURING  1994 

FROM  SPECIAL ACCOUNT TRANSFERRED TO INCOME. 

90.23 

0.14 

UNUTILISED  APPROPRIATIONS  AT 31  DECEMBER  1993 

CARRIED  FORWARD  IN THE SPECIAL ACCOUNT TO MEET 

OUTSTANDING  COMMITMENTS  AT THAT DATE. 

90.37 

5.14 

Table  9: Income  and  Payments  for  1994 

Income  and  Payments 

The  actual  income  for  1994  was  82.64  MioECU  to  which  was  added 

7.52  MioECU  available  appropriations  brought  forward  from  p-evious 

years,  giving  a  total  of  90.16  MioECU.  Total  payment  appropriations 

for  1994  were  95.51  MioECU;  payments  in  the  year  amounted  to 

90.23  MioECU  and  0.14  MioECU  was  transferred  from  the  Special 

Account  to  income.  The  balance  of  5.14  MioECU  was  transferred  to 

the  Special  Reserve  Account  to  meet  commitments  outstanding  at 

31  December  1994.  (Payments  are  summarised  in  Tables  8  and  9). 

Contributions  from  Members 

The  budget  for  Members'  contributions  was  81.67  MioECU  as follows: 

8 0%  from  the  general  budget  of  the  European  Atomic  Energy 

Community  (Euratom); 

1 0%  from  the  UK  Atomic  Energy  Authority  as  Host 

Organisation; 

1 0%  from  members  who  have  Contracts  of  Association  with 

Euratom  in  proportion  to  the  previous  year's  contribution 

from  Euratom  towards  the  cost  of  their  Association  Contracts. 

Table  10  gives  contributions  from  Members  for  1994. 

•125-

A D M I N I S T R A T I ON 

MEMBER 

% 

Mio ECU 

EURATOM 
BELGIUM 
CIEMAT, SPAIN 
CEA, FRANCE 
ENEA, ITALY 
RISO, DENMARK 
LUXEMBOURG 
JNICT 
KFA, GERMANY 
IPP, GERMANY 
KFK, GERMANY 
NFR, SWEDEN 
SWITZERLAND 
FOM, NETHERLANDS 
UKAEA 

80.0000 
0.2745 
0.2809 
1.8143 
1.9316 
0.0756 
0.0018 
0.0752 
0.7947 
2.4602 
0.7316 
0.2400 
0.5171 
0.3403 
10.4622 
100.0000 

65.34 
0.22 
0.23 
1.48 
1.58 
0.06 
0.00 
0.06 
0.65 
2.01 
0.60 
0.20 
0.42 
0.28 
8.54 
81.67 

Table 10: Percentage Contributions to JET for 1994 

Bank Interest 

During the year, funds are normally received on a quarterly basis in 

respect of Members' contributions and intermittently for other items. 

Therefore, the Project has funds not immediately required for discharge 

of its commitments; these funds are placed on deposit accounts at 

market interest rates. During 1994, earned interest was 0.95MioECU. 

Appropriations from Earlier Years 

Unused payment appropriations and excess income over budget of 

1.632 MioECU arising in 1992 and 5.883 MioECU arising in 1993 

were transferred to income in 1994. 

Summary 

Table 11 summarises the financial transactions of the JET Joint 

Undertaking as at 31 December 1994, which have yet to be audited. 

The final audited accounts will be published in due course. 

FINANCIAL TRANSACTIONS 

MioECU 

CUMULATIVE COMMITMENTS 1,485.5 

CUMULATIVE PAYMENTS 1,462.3 

UNPAID COMMITMENTS 23.2 

AMOUNT CARRIED FORWARD IN THE SPECIAL ACCOUNT 5.1 

AMOUNT AVAILABLE FROM 1993 AND 1994 TO SET OFF 

AGAINST FUTURE CONTRIBUTIONS FROM MEMBERS 1.5 

Table 11: Summary of Financial Transactions at 31 December 1994 

■126-

A D M I N I S T R A T I ON 

Contracts  Service 
Contracts  Activity 

During  1994, tender/contract  work was undertaken  by the Contracts 

Service as shown in Table  12. Many larger contracts involved advance 

and retention  payments for which  bank guarantees were required by 

JET. The value of guarantees  held at 31 December  1994 was 4.5MioECU. 

Imports  and  Exports  Services 

Contracts  Service  is also  responsible  for  the  import  and export  of JET 

goods.  712  imports  and  292  exports  were  handled  in  1994. There 

were  also  1274  issues of  goods  to  UK firms.  The total  value  of  issues 

to  all  countries  for  the  year was  8.027MioECU. 

Stores  Organisation 

The bulk  of  material  is procured  on  a  "just  in time"  basis and  stores 

organisation  provides  a receipts  and  delivery  service for  this  material. 

The total  number  of  such  receipts  in  1994 was  17,700. 

Administration  of  Contracts 

The distribution  of  contracts  between  countries  is shown  in Tables  13 

and  14. Table  13  includes  all  contracts  with  a value  of  10.000ECU 

and  above  placed  prior  to  1984, together  with  all contracts  placed 

during  the  period  1984-94. Table  14 is an allocation  of  "high-tech" 

contracts,  which  is based on figures  shown  in Table  13 but  excludes 

all  contracts  below  5.000ECU  and  contracts  covering  civil works, 

FORMAL  TENDER  ACTIONS 

NUMBER 

SUPPLY 

SERVICE 

PERSONNEL 

TOTAL 

55 

23 

29 

107 

CONTRACTS  PLACED 

QUANTITY 

VALUE  (MioECU) 

MAJOR  (>75  kECU) 

MINOR  (<75kECU) 

DIRECT ORDERS 

AMENDMENTS  AND 

WARRANTS 

TOTAL 

50 

4,402 

11,852 

658 

16,962 

7.851 

16.058 

2.516 

23.010 

49.435 

Table12:  Formal Tender  Actions  and  Contracts  placed  during  1994 

•127-

A D M I N I S T R A T I ON 

COUNTRY 

VALUE  (kECU) 

%  OF TOTAL 

UK 

GERMANY 

FRANCE 

ITALY 

SWITZERLAND 

DENMARK 

NETHERLANDS 

BELGIUM 

SWEDEN 

IRELAND 

OTHERS 

TOTALS 

564,354 

163,473 

88,530 

58,914 

43,068 

13,340 

17,465 

11,662 

6,848 

975 

37,716 

56.07 

16.24 

8.80 

5.85 

4.28 

1.33 

1.74 

1.16 

0.68 

0.10 

3.75 

1,006,345 

100.00 

Table  13:  Allocation  of  JET Contracts 

installation,  pipework,  consumables  (including  gases),  maintenance 

operations  and  office  equipment  (including  PCs). 

Personnel  Service 
Staffing 

The JET Team strength  has reduced again this year (Table 15). 

Departures of  Euratom staff  continued at a high rate (over  10%) and 

replacement with  Euratom  recruits was restricted. However, UKAEA 

staff  numbers have increased as a result of  active  recruitment to fill 

vacancies in  1994. The aim is to  have some growth  in team staff 

COUNTRY 

VALUE  (kECU) 

%  OF  TOTAL 

UK 

GERMANY 

FRANCE 

ITALY 

SWITZERLAND 

DENMARK 

NETHERLANDS 

BELGIUM 

SWEDEN 

IRELAND 

OTHERS 

TOTALS 

138,916 

142,951 

77,775 

51,242 

34,795 

7,448 

16,277 

5,062 

4,482 

389 

28,649 

507,986 

27.34 

28.14 

15.31 

10.09 

6.85 

1.47 

3.20 

1.00 

0.88 

0.08 

5.64 

100.00 

Table  14:  Allocation  of  JET "High-Tech"  Contracts 

\ 

-128-

A D M I N I S T R A T I ON 

Germany  22 

Netherlands  11 

Italy  27  France  23 f 

* 

Ireland  14 

J, 

n 

Others  8 

Total  351.5 

Denmark  8 

Sweden  7 

Belgium  6 
Switzerland 4 
Austria  1 
Greece  1 
Luxembourg  1 
Spain  1 

Fig.65: Composition of the JET team by nationality 

numbers in  1995 and  1996 to  meet work  programme  needs.  Figure 65 

shows the composition  of the JET team  by nationality.  In addition to the 

JET Team staff,  222 contract  staff  were on site at 31  December  1994, 

charged to Title 3 of the JET Budget. 

All  team  staff  assignments  expired  at the  end  of  1994, and  once 

the  budgetary  position  for  1995/6  was  clarified, all  existing  team 

staff  were  reassigned  until  31  December  1996. 

Thirty-four  staff  (including  one  part-time  member)  left  the  Project 

during  1994  (16  Euratom,  1 DGXII and  17  UKAEA) the  same total as 

in  1993.  Among  UKAEA  leavers were  five  retirements  but  most 

leavers f'om  the  Euratom  category  obtained  posts  in the  services of 

the  Commission. Two  leavers  (both  UKAEA)  moved  to ITER. 

Recruitment  and  Promotions 

Recruitment  was  restricted  to  UKAEA  candidates  during  1994.  There 

were  six  recruits: four  were  already  employed  by  UKAEA  and  two 

were  newly  recruited  on  fixed  term  contracts. 

TEAM  POSTS 

POSTS FILLED 

POSTS FILLED 

END  1993 

END  1994 

UKAEA 

EURATOM 

DGXII 

TOTAL JET TEAM 

228.5 

143 

7 

378.5 

218.5 

127 

6 

351.5 

Table  15: Team  posts  filled 

-129-

A D M I N I S T R A T I ON 

Three existing team staff  were selected to fill  Group Leader 

vacancies during the year. Following the annual  promotion  exercise 32 

staff  were promoted to  higher  grades. Changes in the  Commission's 

promotion  procedures  necessitated  a review of the  promotion 

procedures for  Euratom staff  at JET. Revised arrangements for assessing 

eligible staff  (Euratom and UKAEA) were introduced. 

Staffing  Policy 

JET staffing  policy was reviewed by the JET Executive Committee  in the 

light of the  need to  maintain a team of trained, experienced and 

dedicated  personnel to operate JET if a proposed extension to  1999 was 

approved. The retention of  staff  remains a problem  due to  uncertainty 

about the  immediate and longer-term future of the  Project, and 

concerns about  post-JET careers. The Project  is exploring  possible 

measures with the  respective employers to  address this problem. 

The employers'  conditions  of  service were  administered, as 

appropriate.  The  UKAEA  Retention  of  Experience  Allowance  was  paid 

to  the  majority  of  UKAEA Team staff  in  December  1994. 

Staff  Relations 

At  the  beginning  of  May  1994 the  European  Parliament  formally 

voted  to  make  available  from  the  Community  Budget  a sum of 

2MioECU for  disbursement  to  UKAEA  staff  ass gned to  JET. These 

disbursements  were  to  be  part  of  a package  of  measures  to  help 

resolve  the  UKAEA  staff  petition  to  the  European  Parliament.  The JET 

Council  agreed  that  JET could  make the  disbursement,  acting  as an 

agent  outside  its  normal  budget 

Consultants 

Twelve  consultants  were  appointed  by JET in  1994 to  work  for  a total 

of  132  man-days  on  scientific/technical  matters. This compares  with 

204  man-days  in  1993. 

Assigned  Associated  Staff 

1994 saw  an  increase  of  58%  in  man-years  of  effort  provided  at JET 

from  the  Associations,  other  than  the  UKAEA,  under  the  Assigned 

Associate  Staff  Agreement.  The total  contribution  for  1994 was  29.1 

man-years  compared  with  23.8  man-years  in  1993; the  UKAEA 

-130-

A D M I N I S T R A T I ON 

LABORATORY 

UKAEA 

(UK) 

IPP 

CEA 

NFR 

ENEA 

CRPP 

JNICT 

(GERMANY) 

(FRANCE) 

(SWEDEN) 

(ITALY) 

(SWITZERLAND) 

(PORTUGAL) 

CIEMAT 

(SPAIN) 

TOTAL 

MAN-YEARS 

12.0 

4.6 

2.9 

2.8 

2.7 

2.1 

1.1 

0.9 

29.1 

Table  16: Staff  Assignments  from  Associated  Laboratories  during  1994 

contribution  within  this  was  12  man-years  in  1994.  There  appeared 

to  be  an  increased  interest  in  assignments  when  JET  operations 

resumed  at  the  end  of  February  1994.  Tables  16  and  17  show  the 

contribution  made  by  the  Associations  in  1994  and  the  distribution 

of  the  Assigned  Associated  Staff  within  the  Project. 

Visiting  Scientists 

Appointments  to  the  Visiting  Scientist  scheme  are  made  through  the 

UKAEA  as  Temporary  Research  Associates.  During  1994,  three  new 

appointments  were  made,  one  each  from  USA,  Japan  and  Russia. 

There  were  seven  Visiting  Scientists  on  site  at  the  year  end. 

Fellows 

During  1994,  eight  new  JET Fellowships  (under  the  Commission's 

Fellowship  Scheme)  were  awarded  for  t wo  years,  three  to  engineers 

and  five  to  scientists.  At  the  end  of  1994,  there  were  26  Fellows  on 

site  (14  engineers  and  12  scientists).  Of  the  research  projects 

undertaken,  five  were  at  post-doctoral  level  and  21  at  a  lower 

post-graduate  level. 

DEPARTMENT 

MAN-YEARS 

TORUS AND MEASUREMENTS  DEPT 

DATA ANALYSIS AND MODELLING UNIT 

HEATING AND OPERATIONS DEPT' 

DIRECTORATE 

TOTAL 

1 7.5 

4.7 

5.9 

1.0 

29.1 

Table  17: Assigned  Staff  within  the  Project  during  1994 

- 1 3 1-

A D M I N I S T R A T I ON 

Students 

The  number  of  student  contracts  awarded  reduced  slightly  during 

1994:  53 students  worked  at JET during  1994, contributing  881 

man-weeks.  70%  of  the  appointments  were  placements  of  13 

weeks  or  more.  No applications  were  received from  students  in 

Luxembourg,  Portugal,  Switzerland  and  Sweden  during  1994,  but 

students were  recruited  from  all  other  Member  states. 

Training 

During  1994,  individual  staff  needs for  skills training  were  met  where 

possible, through  attendance  on  external  short  courses (about  230 

man-days).  Safety  courses  are described  in the  section  on  Safety  and 

Health  Physics. Nine Team staff  were  supported  by JET to  obtain 

vocational  or  academic  qualifications,  in  addition  to  the  53 

undergraduate  students  gaining  work  experience  at JET and  the 

26  Fellows  and  7 other  PhD students. 

There were  60 JET-based  Science  Meetings  and  Symposia.  In 

addition,  training  opportunities  in  languages  and  European  Union 

institutions  and  policy  were  provided  for  Team  staff,  following  the 

recommendations  of  a joint  JET staff-management  working  party. 

Tuition  was  offered  in three  Community  languages,  French,  German 

and  Italian.  100 Team staff  were  allocated  places for  tuition  between 

October  and  December  1994. 

Administration 

Cost  reductions  have  been  achieved,  in  particular  in the  areas of 

JET's telecommunications  and  missions.  The  number  of  missions 

undertaken  by JET staff  in  1994  fell  by 39%. 

JET hosted  16  meetings  of  its governing  anc  advisory  bodies,  for 

which  transport  and  accommodation  was orgarised. 

Health  Physics  And  Safety 

The  Director  is responsible  for  safety  and  is required  by the JET 

Statutes  to  undertake  all  organisational  measures  to  satisfy  relevant 

safety  requirements.  JET continues  to  meet  all  the  requirements  of 

relevant  UK and  EC legislation  and, in accordance  with  the  Host 

Support  Agreement,  JET complies  with  the  safety  regulations  of  the 

Host  Organisation.  Responsibilities  for  Safety  and  Health  Physics 

-132-

A D M I N I S T R A T I ON 

matters  are discharged  by the  Health  Physics and  Safety  Group 

within  the  Coordinating  Staff  Unit. 

Safety 

The Group  provides  a general  safety  service that  incorporates  safety 

related  training,  monitoring,  co-ordination  and  planning  of  statutory 

inspections.  It  ensures there  is an awareness  of  any  new  legal 

requirements  or  changes  to  existing  legislation.  During  1994,  new 

requirements  for  Pressure  Systems  resulted  in the  need for  a written 

scheme of  examination  to  be in  place. This was  in  place to  meet  the 

legislative  implementation  date  of  July  1994. 

During the year, safety  related training continued  but as the year 

was predominantly  associated with  machine operation the  numbers 

show some differences to those of  1993, which was a long  shutdown 

year.  For comparison, figures for  1993 are set out  in brackets. 

Throughout  1994, 320 (470) people attended  Safety Induction  and 98 

(137) attended the  Basic Safety courses.  Radiological  Protection courses 

were attended  by 67 (125)  people, while 46 (125) attended the 

Beryllium  Introduction  courses. Cardio-pulmonary  resuscitation courses 

continue to  be well attended with  220 (250) attending.  In total, there 

were  1474 (2299) attendances at safety related courses for  28 course 

topics during  1994. Throughout  1994, there were 78 (135) accidents 

reported to the  Safety Section. All were classed as minor except for  2 (6) 

which were  reported to the UK Health and Safety Executive sirce they 

resulted in more than three days absence from work.  Neither  accident 

was work  related but occurred within JET buildings. The total  lost time 

from  all accidents in  1994 was 45 (163) days. There was only one 

Incicent  Safety Review Panel that  needed to  be set up in  1994 and the 

Panel report was published during the year. 

Health  Physics 

The  Group  provides  a comprehensive  radiological  protection  and 

occupational  hygiene  service, dosimetry  service,  beryllium  analysis 

and  environmental  monitoring,  both  on  and  off-site. 

Following  the  resumption  of  operations  in January,  after  the 

divertor  shutdown,  there  were  three  short  interventions  into  the 

torus  vessel  during  March,  September  and  November  1994. The 

highest  radiation  dose-rate  measured  in the  torus  during  the year 

•133-

A D M I N I S T R A T I ON 

was 78μ5νΕτ'. The collective dose accrued as a result of torus 

intervention was 0.021 man-Sv, which was approximately 50% of 

that recorded in 1993. The collective dose during the calendar year 

1994 for the Project was 0.026man-Sv and reflects the focus of work 

during long operational periods. The maximum individual dose 

arising from in-vessel work was 1.37mSv, which is similar to the 

1.43mSv individual maximum recorded in 1993. The radiation doses 

incurred by individuals during 1994 remained low and within the JET 

dose limitation policy and well within UK statutory limits. On- and 

off-site radiological monitoring of grass, soil, water and crop samples 

continued to indicate a negligible impact on the local environment as 

a result of JE I operations. 

The Ionising Radiation (Outside Workers) Regulations came into 

force on 1st January 1994 and these Regulations are aimed at 

ensuring the radiological safety of those staff who work at a number 

of locations. There are requirements for the provision and exchange 

of information between the employer and the outside worker. This 

is achieved by the issue of Radiation Passbooks and the Group have 

set the arrangements to ensure compliance with the regulations. 

Beryllium is commonly in use at JET but its hazardous properties 

are well known and controlled through the application of the JET 

Code of Practice for its safe use, which includes the designation of 

Beryllium Controlled Areas and Beryllium Workers. Work in these 

controlled areas continued throughout 1994 with the exposure to 

airborne beryllium quantified by the means of personal air sampling. 

Samples are processed and analysed in the Health Physics Beryllium 

Laboratory and 4715 (14,281) personal air samples were analysed in 

1994. All exposures were below the maximum exposure limit of 

2μgm 3 for beryllium as specified in the UK's Control of Substances 

Hazardous to Health Regulations. 

Safety-related Committees 

There are currently two committees on safety-related matters: 

The Fusion Safety Committee: this Committee chaired by the Head 

of CSU is attended by representatives of relevant operational 

Divisions/Groups at JET and has several independent members, 

including one from another site handling tritium. Its function is to 

keep under review the safety aspects of the project during design, 

■134-

A D M I N I S T R A T I ON 

commissioning and operation, which arise from the use of tritium. Its 

members provide advice on safety issues arising from safety documents 

prepared by JET and related peer reviews, and in particular it considers 

proposals for modifications which could significantly affect safety. It also 

receives reports of significant incidents. 

The JET Safety Working Group: the Group is concerned with day-

to-day aspects of safety and in particular it monitors and makes 

recommendations on the JET Safety at Work System. It follows up 

safety inspections and actions from all incident investigations. 

Meetings are chaired by the Head of the Coordinating Staff Unit and 

there is representation from all operational and advisory groups at 

JET nvolved in safety, plus an independent UKAEA member and staff 

representatives. 

Press and Public Relations 

Whilst the year was a relatively quiet one from a public relations 

viewpoint, there was a continuing interest from the media and the 

general public in both nuclear fusion matters in general and JET 

progress in particular. The restart of JET operations in the new divertor 

configuration attracted interest mainly from the specialist media. 

During 1994, five television crews visited JET to record interviews 

with senior staff and to film background shots of JET and its 

facilities. Among these was a BBC crew accompanied by Lord 

Marshall of Goring, formerly Chairman of the UKAEA. He recalled, 

the protracted political discussions during 1976/77 on the siting of 

JET when he was Chief Scientific Advisor to the UK Government. 

Four radio reporters visited the Project during the year to record 

interviews with senior staff. In addition, five interviews were given for 

local radio stations. In the written media, eight science correspondents 

came to JET to prepare articles on the progress and developments of the 

Project. The main technical aspects in which the media took an interest, 

revolved around the 'new JET' which embodies the pumped divertor 

and on the results of the tritium experiments on the TFTR fusion 

experiment at Princeton Plasma Physics Laboratory, USA. Other topics 

which provoked press interest included the blocking of the Euratom 

Fusion Budget by the European Parliament and the outcome of the 

UKAEA staff's petition to the European Parliament. 

■1 3 5-

A D M I N I S T R A T I ON 

Among the more distinguished visitors were Dr R Dautray, High 

Commissioner of the CEA, France; Sir William Stewart, Chief Scientific 

Advisor to the UK Government; the Russian theoretician, Academician Β 

Kadomtsev; the former Director General of CERN, Professor C Rubbia, 

and UK energy expert Professor Ian Fells. Politicians visiting JET during 

the year included Mr Tim Eggar, the UK Energy Minister. 

As in previous years many groups from Universities, Colleges and 

Schools from several European countries made visits to JET - all of 

whom were given introductory talks and a tour of JET laboratories. 

In addition, groups from professional bodies and local organisations 

visited the Project. In all, about 100 groups were given tours of the 

laboratories. Also JE I hosted visits from prize winners of national 

science competitions in Germany and Portugal. 

In November, during the European Week for Scientific Culture 

organised by the European Commission, younger members of JET staff 

gave illustrated talks on the European Fusion Research Programme to 

local schools. This resulted in requests for follow up visits to JET. 

Publications Group 

The Publications Group provides a Graphics, Phototypesetting, 

Photographic and Reprographics service for the Project. The Group is led 

by the Publications Officer, who is also responsible for the clearance, 

production and distribution of all JET documents. In addition, the Group 

arranges attendance at major International Conferences, and prepares 

papers and posters for these Conferences and Meetings. 

Conferences 

JET provided contributions to a number of major meetings, as follows: 

• 11th International Conference on Plasma Surface Interactions, 

Mito-shi, Japan, May 1994, (2 Invited Papers and 16 Posters); 

• 21th European Conference on Controlled Fusion and Plasma 

Physics, Montpellier, France, June 1994 (3 Invited Papers and 

41 Posters); 

• 18th Symposium on Fusion Technology (SOFT-18), Karlshrue, 

Germany, August 1994, (2 Invited Papers and 33 Posters); 

• 15th IAEA Conference on Plasma Physics and Controlled Nuclear 

Fusion Research, Seville, Spain, September 1994, (9 Oral 

Contributions and 2 Posters); 

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A D M I N I S T R A T I ON 

• 36th Annual Meeting of the American Physical Society - Division 

of Plasma Physics, Minneapolis, USA, November 1994, (1 Invited 

Paper and 10 Posters). 

In total, the Group prepared 168 Papers and 142 Posters for 

presentations to about twenty different Conferences throughout the 

world. Arrangements were also made by the Group for 1 57 

participants to attend these major meetings during the year. 

Publications 

The Publications Office is responsible for the clearance and production 

of all JET presentations (including Journal Papers, Reports, Conference 

Papers, Poster Contributions, Lectures, etc.). Throughout 1994, over 

360 publications were cleared for external presentation. 

During the year, 218 documents were published from the Project 

and the full list is included as an Appendix to the 1994 JET Progress 

Report. This total included 12 JET Reports, 74 JET Preprints, 7 JET 

Internal Reports, 2 JET Technical Notes and 13 JET Divisional Notes. All 

these documents are produced and disseminated by the Group on a 

wide international distribution. In addition, 81 papers were published in 

scientific Journals. In total, the Group produced 3097 new illustrations 

and figures and took 1998 new photographs for publications and other 

disseminated material during 1994. 

■137-

A P P E N D I C ES 

APPENDIX I 
The JET Council 

Member 

The European Atomic Energy Community (EURATOM) 

The Belgian State, acting for its own part ('Laboratoire de Physique 
des Plasmas de l'École Royale Militaire ­ Laboratorium voor plasma­
physica van de Koninklijke Militaire School') and on behalf of the 
Université Libre de Bruxelles' ('Service de physique statistique, plasmas 
et optique non­linéaire de l'ULB'); and of the 'Centre d'Études 
de l'Energie Nucléaire (CEN)/'Studiecentrum voor Kernenergie' (SCK) 

Representative 

P. Fasella 
C. Maisonnier 

P.E.M. Vandenpias 
T. van Rentergem (to April) 
G. Michaux (from April) 

The Centro de Investigaciones Energéticas 
Medioambientales y Tecnológicas (CIEMAT), Spain 

A. Grau Malonda 

Commissariat à L'Énergie Atomique (CEA), France 

D. Escande 
R. Aymar (to October) 
J­P. Schwartz (from October) 

The 'Ente per le Nuova Tecnologie, l'Energia e l'Ambiente' ('ENEA') 
representing all Italian activities falling within the Euratom Fusion 
Programme including that of the 'Consiglio Nazionale delle Ricerche', 
(CNR); dell'Energia Nucleare e delle Energie Alternative (ENEA), Italy 

R. Andreani 
C. Mancini 

The Hellenic Republic (Greece) 

The Forskningscenter Risø (Risø), Denmark 

The Grand Duchy of Luxembourg (Luxembourg) 

The Junta Nacional de Investigação Científica e 
Tecnológica (JNICT), Portugal 

Ireland 

The Forschungszentrum Jülich GmbH(KFA), 
Federal Republic of Germany 

The Max­Planck­Gesellschaft zur Förderung der 
Wissenschaften e.V. Institut für Plasmaphysik (IPP), 
Federal Republic of Germany 

A. Katsanos 

H. von Bülow (Chairman) 
J. Kjems 

Mrs. S. Lucas 
J. Hoffmann (to October) 
R. Becker (from October) 

C. Varandas 
Mrs. M.E. Manso 

F. Turvey 
M. Benville (to September) 
P. O'Neill (from September) 

G. von Klitzing 

K. Pinkau 

The Swedish Natural Science Research Council (NFR), 
Sweden 

G. Leman 
H. Wilhelmsson 

The Swiss Confederation 

The Stichting voor Fundamenteel Onderzoek der 
Materie (FOM), The Netherlands 

The United Kingdom Atomic Energy 
Authority (UKAEA) 

Secretary: J. McMahon, JET Joint Undertaking 

F. Troyon (Vice­Chairman) 
P. Zinsli 

M.J. van der Wiel 
K.H. Chang 

J.R. Bretherton 
D.R. Sweetman 

■139­

A P P E N D I C ES 

APPENDIX II 
The JET Executive Committee 

Member 

The European Atomic Energy Community (EURATOM) 

The Belgian State, acting for its own part ('Laboratoire de Physique 
des Plasmas de l'École Royale Militaire ­ Laboratorium voor plasma­
physica van de Koninklijke Militaire School') and on behalf of the 
Université Libre de Bruxelles' ('Service de physique statistique, plasmas 
et optique non­linéaire de l'ULB'); and of the 'Centre d'Études 
de l'Energie Nucléaire (CENV'Studiecentrum voor Kernenergie' (SCK) 

Representative 

J.P. Rager 
P.J. Kind 

R. Vanhaelewyn 

The Centro de Investigaciones Energéticas 
Medioambientales y Tecnológicas (CIEMAT), Spain 

F. Mañero 

Commissariat à L'Énergie Atomique (CEA), France 

R. Gravier 
C. Samour (to January) 
Mrs. P. Livanos (from January) 

The 'Ente per le Nuova Tecnologie, l'Energia e l'Ambiente' ('ENEA') 
representing all Italian activities falling within the Euratom Fusion 
Programme including that of the 'Consiglio Nazionale delle Ricerche', 
(CNR); dell'Energia Nucleare e delle Energie Alternative (ENEA), Italy 

A. Coletti 
M. Samuelli (to May) 
F. Pecorella (from May) 

The Hellenic Republic (Greece) 

The Forskningscenter Risø (Risø), Denmark 

The Grand Duchy of Luxembourg (Luxembourg) 

The Junta Nacional de Investigação Científica e 
Tecnológica (JNICT), Portugal 

Ireland 

The Forschungszentrum Jülich GmbH (KFA), 
Federal Republic of Germany 

The Max­Planck­Gesellschaft zur Förderung der 
Wissenschaften e.V. Institut für Plasmaphysik (IPP), 
Federal Republic of Germany 

The Swedish Natural Science Research Council (NFR), 
Sweden 

The Swiss Confederation 

The Stichting voor Fundamenteel Onderzoek der 
Materie (FOM), The Netherlands 

The United Kingdom Atomic Energy 
Authority (UKAEA) 

Secretary: J. McMahon, JET Joint Undertaking 

■140­

N. Chrysochoides 

Mrs. L. Grønberg 
V.O. Jensen 

R. Becker (to October) 
C. Bartocci (from October) 

J. Bonfim (to April) 
J. da Costa Cabral (from April) 
F. Serra 

F. Turvey (Chairman) 
D. Taylor 

V. Hertling 

Mrs. I. Kramer 

G. Leman (Vice­Chairman) 
L. Gidefeldt 

M. Tran 
L. de Faveri (to February) 
S. Berthet (from February) 

A. Verhoeven 

D.C. Robinson 
T.J. Elsworth (to April) 
T. Conlon (from April) 

A P P E N D I C ES 

APPENDIX  III 
The  JET Scientific  Council 

Members  appointed  by  the  JET  Council 

K.  Lackner  (Chairman) 

EURATOM-IPP  Association 

D.  Moreau 

EURATOM-CEA  Association 

Max-Planck-Institut  für  Plasmaphysik 

Département  de  Recherches  sur  la  Fusion  Contrôlée 

D-85748  Garching  bei  München 

Federal  Republic  of  Germany 

R.  Bartiromo 

EURATOM-ENEA  Association 

ENEA  Centro  di  Frascati 

Casella  Postale  65 

1-00044  Frascati/Roma,  Italy 

F.  Engelmann 

NET  ~eam 

Centre  d'Études  Nucléaires  Cadarache 

Boîte  Postale  No.1 

F-13108  St.  Paul  lez  Durance,  France 

D.C.  Robinson 

EURATOM-UKAEA  Association 

UKAEA  Government  Division 

Culham  Science  and  Technology  Centre 

Abingdon,  Oxfordshire,  OX14  3DB 

United  Kingdom 

Max-Planck-Institut  für  Plasmaphysik 

F.C.  Schüller 

D-85748  Garching  bei  München 

Federal  Republic  of  Germany 

M.  Gasparatto 

EURATOM-ENEA  Association 

ENEA  Centro  di  Frascati 

Casella  Postale  65 

1-00044  Frascati/Roma,  Italy 

A.  Grosman 

EURATOM-CEA  Association 

EURATOM-FOM  Association 

FOM  Instituut  voor  Plasmafysica  'Rijnhuizen' 

Postbus  1207  -  Edisonbaan  14 

NL-3430  BE Nieuwegein,  The  Netherlands 

D.R.  Sweetman 

EURATOM-UKAEA  Association 

UKAEA  Government  Division 

Culham  Science  and  Technology  Centre 

Abingdon,  Oxfordshire,  OX14  3DB 

United  Kingdom 

Département  de  Recherches  sur  la  Fusion  Contrôlée 

Centre  d'Études  Nucléaires  Cadarache 

F.  Wagner 

Boîte  Postale  No.1 

EURATOM-IPP  Association 

F-13108  St.  Paul  lez  Durance,  France 

Max-Planck  Institut  für  Plasmaphysik 

T.  He  Isten  (Honorary  Secretary) 

EURATOM-NFR  Association 

Royal  Institute  of  Technology 

Alfven  Laboratory 

D-85748  Garching  bei  München 

Federal  Republic  of  Germany 

R.  Weynants 

EURATOM-EB:BS  Association 

Department  of  Fusion  Plasma  Physics 

Laboratoire  de  Physique  des  Plasmas/ 

S-10C44  Stockholm,  Sweden 

Laboratorium  voor  Plasmafysica 

F.  Hofmann 

EURATOM-SUISSE  Association 

Centre  de  Recherches  en  Physique  des  Plasmas 

Ecole  Polytechnique  Fédérale 

21  Avenue  des  Bains 

CH-1007  Lausanne,  Switzerland 

P.  Lallia 

Ecole  Royale  Militaire/Koninklijke  Militaire  School 

Avenue  de  la  Renaissancelaan,  30 

B-1040  Brussels,  Belgium 

G.  Wolf 

EURA~OM-KFA  Association 

Forschungszentrum  Jülich  GmbH 

Institut  für  Plasmaphysik 

Postfach  1913 

DG-XII,  Commission  of  the  European  Communities 

D-52425  Jülich  1,  Federal  Republic  of  Germany 

200,  Rue  de  la  Loi 

B-1049  Brussels,  Belguim 

Staff  Secretary:  M.L.  Watkins,  JET.Joint  Undertaking