W.R.Spears

1
Status of ITER

• W.R.Spears
• for the ITER International and Participant Teams
• Forum on the Future of Fusion
• Washington, 19-21st November, 2003

2
Content

• Technical Status and Developments
• Preparations for Construction
• Cultural Improvements
• New Tools
• Negotiations Status
• Site Choice
• Joint Implementation
• Perspective for 2004
• Timescale for Parties
• Activities for Project

3
What is ITER?

• Programmatic
• Show scientific and technological feasibility of
  fusion energy for peaceful purposes.
• Technical
• High plasma fusion power gain (Q 10), extended
  DT burn - steady state ultimate goal.
• Test essential technologies in reactor-relevant
  physics and technology environment.
• Test high-heat-flux and nuclear components -
  average neutron wall load  0.5 MW/m2 and average
  lifetime neutron fluence of  0.3 MWa/m2.
• Demonstrate safety and environmental
  acceptability of fusion.
• Strategic
• Answer, in single device, all feasibility issues
  needed to define a subsequent electricity-generati
  ng fusion power plant - except for material
  developments to provide low activation and larger
  14 MeV neutron resistance for in-vessel
  components

4
ITER Inductive Performance

• For the first time in the history of the
  development of magnetic confinement fusion as an
  energy source, Q gtgt 1.
• The plasma will be dominated by self-heating from
  alpha particles generated in the reaction of
  deuterium and tritium fuel.
• Reference scenario (ELMy H-mode) with inboard
  fuelling and high triangularity would provide
  Qgt10 even at todays most severe experimental
  density limits.

(Operation range shown for various helium
confinement assumptions)
5
ITER Hybrid and Steady State Performance

• Typical inductively driven burn time of ITER 7
  minutes.
• Longer burn desirable to test steam-raising
  tritium-breeding test blankets, - more typical of
  power reactor conditions.

• Performance projections for hybrid operation,
  stretching the burn time up to steady state, are
  for limits (e.g. in plasma density) seen in many
  of todays experiments.
• Some experiments show that if the power entering
  the plasma core is sufficiently high, a second
  transport barrier occurs inside, limiting even
  more the heat conduction across the plasma.
• Control of this barrier would significantly
  increase Q for a given burn time.

6
Design - Main Features
Central Solenoid Nb3Sn, 6 modules
Blanket Module 421 modules
Vacuum Vessel 9 sectors
Outer Intercoil Structure
Cryostat 24 m high x 28 m dia.
Toroidal Field Coil Nb3Sn, 18, wedged
Port Plug (IC Heating) 6 heating 3 test
blankets 2 limiters/RH rem. diagnostics
Poloidal Field Coil Nb-Ti, 6

• Divertor
• 54 cassettes

Machine Gravity Supports (recently remodelled)
Torus Cryopump 8, rearranged
7
Main changes in ITER since July 2001

• Magnets
• increased critical current (from 6 to 800
  A/mm2)
• use of stainless steel jacketing in all
  conductors
• outer intercoil structure uses friction joint of
  welded plates
• Vessel/Blanket
• support arrangement simplified
• nine lower ports
• blanket module has FW supported from welded
  central leg
• improved interlocking of thermal shield
• Building/Services
• introduction of port cells
• relocate gallery equipment - access, e/m loads
• incorporate seismic isolation for
  Cadarache/Rokkasho
• improve site layout

8
Change Control

• Technical Coordination Meetings (TCM)
• Chaired by Interim Project Leader (IPL)
• Decides on change proposals (DCRs)
• Organises and schedules supporting work and
  priorities
• Design Change Requests (DCR)
• Document proposals for changes
• Design Work Orders (DWO)
• Request CAD effort
• Design Work Check (DWC)
• Process to check drawing ofice output
• Design Integration/Drawing Office (DIDO) Meetings
• Reviews ongoing CAD progress, prioritises new CAD
  effort allocation, and schedules detailed design
  reviews

9
ITER Transitional Arrangements

• Due to positive direction of Negotiations, ITER
  Transitional Arrangements (ITA) began in January
  2002 and will continue until the IIFEO is
  established.
• Enable IIFEO to function without delay following
  signature/ratification of the Agreement.
• establish interim key organisational elements
  Interim Project Team and Leader, Preparatory
  Committee of ITER Council
• following site choice, establish provisionally
  staffed interim joint work site there
• coordinate Parties domestic preparations for
  contributing to ITER Joint Implementation
• identify and provisionally assign potential
  senior staff
• elaborate administrative procedures, documents
  and other tools foreseen for managing ITER Joint
  Implementation.
• continue joint technical preparations.

10
Technical Preparations

• Licensing application - close dialogue with
  potential regulators ? review design to ensure
  its quality and completeness and carry out
  additional safety analyses.
• Establish technical specifications for long-lead
  items, i.e. mainly magnets, vacuum vessel, and
  buildings.
• Enhance technical organisation to maintain ITER
  design and prepare for an efficient start of
  construction, and provide reliable basis for
  licensing, including
• Upgrading of CAD system
• Introduction of virtual product data management
  system (VPDM)
• Introduce new document management system (DMS)
• Preparation for later introduction of asset
  tracking system (ATS)
• Improvement of underlying structures and systems
  (mainly IT)
• Preparation for move to construction site

11
Construction Schedule

• 7 year construction
• 1 year integrated commissioning
• IIFEO established 2 years before award of
  construction license
• Long lead item calls for tender sent out and
  procurement started before license awarded
• Success-oriented schedule
• First plasma in 2014

12
Procurement Specifications

• Drafting of detailed technical specifications for
  long lead items
• Magnets
• strand and conductor
• PF and TF coils
• Vessel
• main vessel and ports
• blanket coolant manifolds
• Buildings
• tokamak complex
• cryogenic halls used for PF coil winding
• service tunnels
• Task Forces established with PT/IT membership to
  complete work in necesary detail and with
  industrial realism
• Development of other procurement specifications
  to cover interfaces with long lead items -
  resource limited

13
Document and Model Management Tools

• IDoMS lacks basic commercial product features
• tree/network navigation of linked documents,
• approval workflow tracking,
• document validity (effectivity) according to
  circumstances,
• electronic signature,
• realistic user roles and security.
• These quality control features needed for
  licensing, and controlled worldwide access.
• ITER needs virtual product data management
  (VPDM) software for 3D digital mockup, essential
  for configuration management in the tokamak/ pit.
  
• clash detection, utility routing, and
  increasingly collaborative design with the
  Participants, need better tools to control the
  configuration and available space.
• CATIA 4 soon obsolete. CATIA 5, cannot be used
  without a PDM. Advantageous to upgrade
• nearly all procurement models are in one system
• to be in tune with manufacturers, who have
  already upgraded, and to rehearse conversion
  processes with other manufacturers CAD systems
• to train operators while deadlines are not so
  pressing.
• to rehearse processes and tailor software systems
  essential post-IIFEO.

14
CATIA/VPDM/DMS implementation timescale
2004
2003
DMS Selection
DMS fully productive
CERN EDMS Pilot
VPDM Selection
Smarteam Pilot
VPDM fully productive
VPM Pilot
DMS VPDM Options Study
Expansion as needed
Initial work with CATIA V5
15
Temporary Documentation

• Before introduction of new document management
  system need to define design status and provide
  area for information exchange.

16
Negotiations

• draft Joint Implementation Agreement
• select ITER construction site
• agree how the procurement and costs will be
  shared
• define how the project will be managed
• identify the Director General and senior staff

17
Potential Sites
Spain Vandellòs
France (Cadarache)
Canada (Clarington)
Japan (Rokkasho)
Joint assessment of sites (JASS) has concluded
all sites are feasible
18
Timetable for Consensus on Site Preference and
the JIA for Signature
2002
2003
JULY
JUNE
JULY
AUG
SEPT
OCT
NOV
DEC
JAN
FEB
MAR
MAY
APR
N8 St. Petersburg
N5 TorontoSept. 17/18
N6AomoriOct 29/30
N7BarcelonaDec 10/11
N9
N10
Meetings
Consensus on Preferred Site / Cost Sharing / DG
Final Joint Implementation Agreement
Final JASS Report
Site Evaluation (JASS) Process
Site-Specific Negotiations
Clarington Evaluation
Cadarache Vandellos Evaluation
Rokkasho Evaluation
Process
Development of Scenarios
Higher/Political Level Discussions Leading to
Site Decision
19
ITER International Fusion Energy Organisation
IIFEO
Council
Science and Technology Advisory Committee
Construction Programme Advisory Committee
Auditors
Host Country
Director-General (DG)
DDGs
Organisation/Host Relation
Organisation Personnel (professionals support
staff)
Central Team
Staff of DG
Supporting Services
Support for Project
Management Computer Network
Technical work
Clerical work etc.
Project Manage.
Admin.
Integration Interface
Procure- ment
QA.
Physics
Safety Licensing
Host Relation
Contract
Field Team
Field Team
Field Team
for construction phase
Domestic Agency
Domestic Agency
Domestic Agency
Industries and Other Organisations
Industries and Other Organisations
Industries and Other Organisations
20
Financial Arrangements

• Ensures fair return
• Ensures fair share of value to project
• Allows Parties to get involved in what they view
  as key technologies
• Project needs the power to intevene to control
  procurements of interfacing systems and introduce
  essential design changes which may affect cost

21
Cost

• Cost of ITER established using multi-party
  industrial estimates consolidated using uniform
  material costs and labour rates.
• Cost estimate established as basis of sharing and
  as a valuation to give Parties indicative costs.
  Actual costs depend on Parties individual
  procurement practices.

1988
22
Cost Sharing

• EU TF(0.5), conductors, cassette and outer
  target, vac.pumps, div. RH, casks (0.5), isotope
  sep., IC, EC, diag.
• JA TF(0.5), conductors, inner target, blanket
  RH, EC, diag.
• KO conductors, vessel ports (0.67), blanket
  (0.2), assembly tools, thermal shield, T storage,
  AC/DC (0.65), diag.
• CN magnet supports, correction coils,
  conductors, blanket (0.2), cryostat, gas
  injection, casks (0.5), HV substation, AC/DC
  (0.35), diag.
• RF PF1, conductors, vessel ports (0.33), blanket
  (0.2), port limiters, flexibles, dome and PFC
  tests, Discharge circuits, EC, diag.
• US CS(0.5), conductors, blanket (0.1),
  vac.pumps, pellet inj., vessel/in-vessel cooling,
  tok exh. proc., IC, EC, diag.
• Fund Feeders, Shielding, vac.pumps, misc vac.,
  viewing, NB RH, Hot cell eq., cryodist., diag.,
  CODAC, other sundry items

Canadian Site
EU/JA Site
Remaining allocation depends on site
23
Risk Management

• ITER design development/detailing since July 2001
  has not primarily reduced cost but reduced risk
  of cost overrun and created margins to cover the
  unforeseen. Margins actually realised will not
  be seen by the project for items provided in
  kind.
• Because ITER is a first-of-a-kind design,
  considerable RD was done to minimise technical
  problems of component fabrication, and to avoid
  later delays, design changes, regulatory
  problems, and difficulties in assembly.
• Despite this, suppliers have to demonstrate high
  success rate in production. Where more than one
  Party is involved, staged contracts can be used,
  plus donation of unused funds to the project from
  inadequate producers Party, to minimise overall
  risks to production.
• To cover delivery delays during construction,
  penalty clauses must be included in contracts to
  compensate the other Parties and the project.
  The Project Team needs to be closely involved to
  help recognise and limit such occurrences.
• Such an experiment cannot be built without some
  changes during construction. The strong presence
  of the Project Team on the suppliers premises
  should allow some difficulties to be absorbed by
  adapting interfaces with later manufactured
  items. But in the end, the other Parties may
  have to jointly compensate the manufacturing
  Party for consequent cost increases beyond that
  Partys margins.
• What to do in each of the situations that could
  arise will be documented in the JIA.

24
Preparation toward ITER Organization (Draft)
Decision Making
Technical Work
Legal Work
Site, Cost Senior Staff Decision
IT (Interim Leader)
NSSG
N
P
IPC
Completion of Draft Agreement (Initial)
ITER Preparatory Committee (IPC)
D-G Nominee and Transitional Team (Participant
team)
Working Group With Topical Sub Groups (with T.T.)
Preparatory Tasks
Agreement Entering Into Force
ITER International Fusion Energy Organization
Council
Director General
25
Timetable for Establishing ITER International
Fusion Energy Organisation
2003
2004
NOV
OCT
NOV
DEC
JAN
FEB
MAR
APR
MAY
JUN
JUL
SEP
AUG
Consensus on Preferred Site / Cost Sharing / DG
Establishment of the IIFEO
Domestic Checking Process
JA
Approval by Diet
Sign and submit to Diet
CA
Finalise/Verify JIA
Create Domestic Agency
Cabinet Approval
CN
Signature
Negotiators Initialing of Agreement
EU
Signature
Council Procedure
KO
Approval by National Assy.
Sign and submit to NA
RF
Ratification
Signature
US
Signature
Consultation with Senate
26
Computer Infrastructure Implementation
2005
2004
DMS fully productive
VPDM fully productive
CATIA 5 Expansion
DMS Server Basic Configuration
Establish Local Mirror
VPDM Server Basic Configuration
Remaining Servers Basic Configuration
Establish ITER Site Mirror
Transfer Servers to ITER site
27
Conclusions

• The ITER Transitional Arrangements are being used
  by the project to get many things ready that will
  ease the path once the negotiations are
  successfully completed
• finalisation of long lead time procurement
  packages
• preparation of safety case for licensing bodies
• adoption of tools that are necessary for project
  and quality control
• The inter-governmental negotiations aiming at
  an agreement on construction and operation of
  ITER are nearing completion and, if all goes
  well, can be expected to lead to the start of the
  establishment of the construction team in 2004.

28
ITER Nominal Parameters

• Total fusion power 500 MW (700MW)
• Q fusion power/auxiliary heating power 10
  (inductive)
• Average neutron wall loading 0.57 MW/m2 (0.8
  MW/m2)
• Plasma inductive burn time 300 s
• Plasma major radius 6.2 m
• Plasma minor radius 2.0 m
• Plasma current (inductive, Ip) 15 MA (17.4 MA)
• Vertical elongation _at_95 flux surface/separatrix 1
  .70/1.85
• Triangularity _at_95 flux surface/separatrix 0.33/0
  .49
• Safety factor _at_95 flux surface 3.0
• Toroidal field _at_ 6.2 m radius 5.3 T
• Plasma volume 837 m3
• Plasma surface 678 m2
• Installed auxiliary heating/current drive
  power 73 MW (100 MW)

29
Work on Outer Intercoil Structure

• These function as both shear panels, to provide
  out-of-plane support, and to hold in the TF
  coils. They have insulating joints which must
  transmit tension and shear.
• Two concepts (box and friction), two fabrication
  methods (casting of thick sections and welding of
  plates).
• Stress analysis - problems with box design due to
  fatigue in the bolts, and opening of flanges
  connecting adjacent boxes which overloads the
  shear keys.
• Friction joint design selected, which leads to
  the selection of the welded plate route.

30
Vacuum Vessel Support System

• VV supported independently of magnets at the
  lower ports
• Possible to adjust the VV in the pit after
  welding of the sectors.
• Snubbers used to limit the radial movement during
  earthquake
• Locate parts requiring maintenance outside
  cryostat
• Vertical ropes for upwards loads

31
VV Lower ports

• 9 large VV penetrations at divertor level
• 3 ports for divertor cassette maintenance
• 2 for diagnostic equipment
• 4 ducts (connected to 8 torus pumps)
• Cryostat cryopumps can be made similar to the
  torus pumps and can be located at the same level
  with improved maintainability

.
32
Thermal Shield

• Cryostat thermal shield close to the magnet
  structures and supported in the central region by
  the TF coils
• Most labyrinths eliminated
• Reduced thermal radiation to 4K structures
• Separation of cold volume from the part crossed
  by water pipes
• Reduced total surface (and cost) of the TS.

33
Port Cell Layout

• Second containment barrier moved to port cell
  door
• Reduced number of operations in irradiated areas.

34
Design - Magnets and Structures

• Superconducting.
• Nb3Sn toroidal field (TF) coils produce
  confining/stabilizing toroidal field
• NbTi poloidal field (PF) coils position and shape
  plasma
• modular Nb3Sn central solenoid (CS) coil induces
  current in the plasma.
• correction coils correct error fields due to
  manufacturing/assembly imperfections, and
  stabilize plasma against resistive wall modes.
• TF coil case provides main structure of the
  magnet system and the machine core. PF coils and
  vacuum vessel are linked to it. All interaction
  forces resisted internally.
• TF coil inboard legs wedged together along their
  side walls and linked at top and bottom by two
  strong coaxial rings which provide toroidal
  compression and resist the local de-wedging of
  those legs under load.
• On the outboard leg, out-of-plane support
  provided by intercoil structures integrated with
  TF coil cases.
• Magnet system weighs 8,700 t.

35
Design - Main Vacuum Vessel

• Primary function
• high quality vacuum for the plasma
• first confinement barrier to radioactive
  materials
• second barrier for the separation of air from
  potential sources of in-vessel hydrogen
  generation.
• Decay heat of all in-vessel components can be
  removed by the water in the VV primary heat
  transfer system (PHTS) system, even in conditions
  when the other PHTSs are not functioning.
• 9 x 40 vessel sectors.

36
Design - Blanket

• 421 blanket modules with detachable faceted first
  wall (FW) with Be armour on a water-cooled copper
  substrate, attached to a SS shielding block -
  minimises radioactive waste and simplifies
  manufacture.
• Blanket cooling channels are mounted on the
  vessel.
• Design strongly affected by need to resist
  electromagnetic forces.
• Initial blanket acts solely as a neutron shield,
  and tritium breeding experiments are carried out
  on test blanket modules inserted and withdrawn at
  radial equatorial ports.
• The outboard blanket may later be replaced by a
  tritium breeding blanket.

Vessel
Inlet/outlet manifolds
Flexible supports
First wall panel
Shear key
Gripping hole
Hole to fit flexible support
Electrical strap
Shield block
37
Design - Vessel In-Vessel

• Double-walled vacuum vessel lined by modular
  removable components, including blanket modules,
  divertor cassettes, and diagnostics sensors, as
  well as port plugs for limiters, heating
  antennae, diagnostics and test blanket modules.
  All these removable components are mechanically
  attached to the VV. The total vessel/in-vessel
  mass is 10,000 t.
• Components absorb most of the radiated heat from
  the plasma and protect the magnet coils from
  excessive nuclear radiation. Shielding is steel
  and water. A tight fitting configuration of the
  VV to the plasma aids passive plasma vertical
  stability, and ferromagnetic material inserts
  in the VV located in the shadow of the TF coils
  reduce toroidal field ripple and its associated
  particle losses.

38
Design - Divertor

• 54 cassettes.
• Target and divertor floor form a V which traps
  neutral particles, protecting the target plates,
  without adversely affecting helium removal.
• Large openings between the inner and outer
  divertor balance heat loads in the inboard and
  outboard channels.
• Design uses C at the vertical target strike
  points. W is the backup. C is best able to
  withstand large power density pulses (ELMs,
  disruptions), but produces tritiated dust and T
  co-deposited with C which has to be periodically
  removed. The choice can be made at the time of
  procurement.

Vertical target (W part)
Dome (W)
Vertical target (C part)
Transparent liner for pumping
39
Design - Heating/Current Drive
Waveguides
Windows

• Need to adjust plasma density profiles and heat
  or drive current at various plasma radii - to
  move towards steady state operation and to
  suppress instabilities.
• High energy (1 MeV D-) ion beams radio
  frequency heating tuned to key plasma frequencies
  (ion - 40-56 MHz, electron cyclotron - 170 GHz,
  lower hybrid - 5 GHz)
• RF systems modular and interchangeable in
  equatorial ports EC used in upper ports to
  suppress neoclassical tearing modes
• 2 main beamlines, with room for third height
  adjustable 0.15 to - 0.42 m
• Initial installation 73 MW with room for
  expansion to 130 MW (110 MW PS limit).

Steerable mirror

Front shield
Electron Cyclotron System Equatorial Port Plug
40
Design - Diagnostics

X-ray survey Imaging VUV Spectroscopy
Edge Thomson scattering
X-ray crystal spectroscopy Divertor VUV
spectroscopy X-ray survey Core VUV monitor
Motional Stark effect Toroidal interferometer Elec
tron cyclotron emission Wide-angle viewing/IR
Magnetic diagnostic coils
Divertor reflectometry
X-point LIDAR
About 40 different diagnostic systems will be
used. Selected for endurance and nuclear
compatibility. 3 groups basic control, advanced
control, physics studies. Mostly designed and
operated by Parties researchers.
41
Design - Tokamak Building

• Provides biological shield around cryostat to
  minimise activation and permit human access.
• Additional confinement barrier.
• Controls (with HVAC) contamination spread.
• Provides shielding during remote handling cask
  transport.
• Can be seismically isolated.

Tokamak Building - Equatorial Layout
42
ITER Site Layout
43
Assembly

• Lower cryostat, tokamak gravity supports, lower
  PF and correction coils, TF pre-tensioning rings,
  etc., placed in the pit.
• 40 sectors of vessel 2 TF coils, thermal
  shields, etc., assembled together on-site and
  moved to pit.
• Sectors welded in opposition to minimise
  distortions. TF coil pre-tensioning rings
  installed. Machine datum established. Clean
  conditions in-vessel.
• In-vessel components installed and aligned.
  PF/correction coils and cryostat above equator
  installed.
• External installations proceed in parallel.

44
Technology RD

• Where new technologies or implementations were
  necessary, extensive RD was carried out to prove
  its viability. Seven Large RD Projects were
  established to demonstrate fabrication
  feasibility of major components of ITER.

45
Results of Magnet RD

• Model coil results caused concern - 1.5K
  reduction in max. allowable operating
  temperature, where a margin of about 0.5K allowed
  for in the design. Due to high loading (70t/m)
  and unexpected strand deformation (or breakage?).
• Benefits of low jacket COE material (titanium)
  not realised.
• Better to standardise on stainless steel, and to
  increase strand critical current density from
  500-700 A/mm2 to 700-800 A/mm2.

• Fortunately coincides with increased performance
  strand now produced by industry driven mainly by
  accelerator programme.

46
Max. field 13.5T, max. current 46kA, stored
energy 640MJ (max. in Nb3Sn) Ramp-up 1.2T/s (goal
0.4) and rampdown rates of -1.5T/s (goal -1.2) in
insert coils, and 10,000 cycle test.
47
Welding radial plate cover plate
48
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49
Divertor outer channel undergoing thermohydraulic
tests.
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51
Dimensional accuracy after welding sector halves
3 mm
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55
Positioning 0.5 mm and 0.1, rail deployed 90
around torus in 30 min..
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58
Current 80kA (above value needed in ITER and the
maximum possible in the facility) with a field of
8T. Highest current ever driven in a
superconducting coil.
59
Blanket Option B