Technical Characteristics 1

1
Technical Characteristics (1)

• Performance
• Fusion power amplification gt 10 with inductive
  current drive (ignition not precluded).
• Fusion power amplification gt 5 using
  non-inductive current drive.
• Typical fusion power level 500 MW
• Testing
• Integrate and test all essential fusion reactor
  technologies and components.
• Design
• Use existing technology and physics database to
  give confidence but be able to access advanced
  operational modes.
• Operation equivalent to a few 10000 inductive
  pulses of 300-500 s.
• Average neutron flux 0.5 MW/m2
• Average fluence 0.3 MWa/m2

2
Technical Characteristics (2)

• Operation
• Address all aspects of plasma dominated by alpha
  particle (helium) heating through burning plasma
  experiments.
• Low fluence functional tests of DEMO-relevant
  blanket modules early high reliability tests
  later.
• Device operation 20 years. Tritium to be
  supplied from external sources.

3
ITER Parameters

• Total fusion power 500 MW (700MW)
• Q fusion power/auxiliary heating power 10
• 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 (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)

4
Inductive Operation (1)

• Experimental basis

The figure shows observed energy confinement time
(s) in various experiments versus value derived
from the scaling law
tE,th 0.0562 HH Ip0.93 BT0.15 P-0.69 ne0.41
M0.19 R1.97 e0.58 Ka0.78
where Ip plasma current (MA) BT on-axis
toroidal field (T) P internal external
heating power (MW) ne electron density
(1019m-3) M atomic mass (AMU) R major
radius (m) e inverse aspect ratio
(a/R) KaSo/pa2, So plasma x-sectional area.
HH, the confinement time enhancement factor,
measures the quality of confinement ( 1 for the
dotted line in the figure).
5
Inductive Operation (2)

• Range of performance

Fusion power (Pfus) versus auxiliary power (Paux)
for a range of currents and for HH 1 and
ne/nGreenwald 0.85. Minimum fusion power is
limited to a factor 1.3 above the expected power
at which transition to L-mode would occur, namely
PLH 0.75 M-1 BT0.82 ne0.58 R1.00 a0.81
6
Inductive Operation (3)

• Flexibility in reaching Q10

The combination of a range of plasma parameters
will allow Q10 to be obtained. The figures show
the operational domain in terms of fusion power
and HH, plus the various limiting boundaries that
are thought to apply. Q10 is maintained
within the shaded region by adjusting auxiliary
power and density.
7
Inductive Operation (4)
The results show the flexibility of the
design, show its capacity to respond to
factors that degrade confinement, show its
ability to maintain the goal of extended burn
Q10 operation, imply the ability to explore
higher Q operation, provided energy confinement
times consistent with the confinement scaling are
maintained.
8
Steady State/Hybrid Operation
Hybrid modes of operation are being evaluated as
a promising route towards establishing true
steady-state modes of operation. There, in
addition to inductively driven current, a
substantial fraction of the plasma current is
driven by external heating and the bootstrap
effect, leading to extension of the burn
duration. This form of operation would be well
suited to systems engineering tests. For a given
value of fusion power (and hence Q), as the
confinement enhancement factor, HH, increases
(simultaneously decreasing plasma density and
increasing bN), the plasma loop voltage falls
towards zero. For example, operation with
Vloop 0.02 V and Ip 12 MA, which corresponds
to a flat-top length of 2500 s, is expected at HH
1, Q 5, ne/nGreenwald 0.7, and bN 2.5.
True steady-state operation at Q 5 can be
achieved with HH 1.2 and bN 2.8.

• Operation space for Ip 12 MA and PCD 100 MW,
  in terms of fusion power versus confinement
  enhancement factor, and the transition from
  hybrid to true steady-state operation.

9
Design - Main Features (1)
Central Solenoid
Blanket Module
Vacuum Vessel
Outer Intercoil Structure
Cryostat
Toroidal Field Coil
Port Plug (EC Heating)
Poloidal Field Coil

• Divertor

Machine Gravity Supports
Torus Cryopump
10
Design - Main Features (2)

• Superconducting toroidal field coils (18 coils)
• Superconductor Nb3Sn in circular stainless
  steel (SS) jacket in grooved radial plates
• Structure Pancake wound, in welded SS case,
  wind, react and transfer technology
• Superconducting Central Solenoid (CS)
• Superconductor Nb3Sn in square Incoloy jacket,
  or in circular Ti/SS jacket inside SS
  U-channels
• Structure Pancake wound, 3 double or 1
  hexa-pancake, wind react and transfer
  technology
• Superconducting poloidal field coils (PF 1-6)
• Superconductor NbTi in square SS conduit
• Structure Double pancakes
• Vacuum Vessel (9 sectors)
• Structure Double-wall welded ribbed shell,
  with internal shield plates and
  ferromagnetic inserts
• Material SS 316 LN structure, SS 304 with 2
  boron

11
Design - Main Features (3)

• First Wall/Blanket (421 modules) (Initial DT
  Phase)
• Structure Single curvature faceted separate FW
  attached to shielding block which is fixed
  to vessel
• Materials Be armour, Cu-alloy heat sink, SS
  316 LN structure
• Divertor (54 cassettes)
• Configuration Single null, cast or welded
  plates, cassettes
• Materials W alloy and C plasma facing
  components, copper alloy heat sink, SS 316
  LN structure
• Cryostat
• Structure Ribbed cylinder with flat ends
• Maximum inner dimensions 28 m diameter, 24 m
  height
• Material SS 304L
• Heat Transfer Systems (water-cooled)
• Heat released in the tokamak duringnominal
  pulsed op. 750 MW at 3 and 4.2 MPa water
  pressure, 120C

12
Design - Main Features (4)

• Cryoplant
• Nominal average He refrig. /liquefac. rate for
  magnets divertor cryopumps (4.5K) 55 kW
  / 0.13 kg/s
• Nominal cooling capacity of the thermal shields
  at 80 K 660 kW
• Additional Heating and Current Drive
• Candidate systems Electron
  Cyclotron, Ion Cyclotron, Lower Hybrid ,
  Negative Ion Neutral Beam
• Electrical Power Supply
• Pulsed Power supply from grid total
  active/reactive power demand 500 MW / 400 MVAr
• Steady-State Power Supply from grid total
  active/reactive power demand 110 MW/ 78 MVAr

13
Design - Magnets and Structures (1)

• The superconducting magnet system has three main
  subsystems
• 18 toroidal field (TF) coils which produce the
  confining/stabilizing toroidal field
• 6 poloidal field (PF) coils which contribute to
  the plasma positioning and shaping
• a central solenoid (CS) coil which provides the
  main contribution to inducing current in the
  plasma.
• Correction coils (located above, outboard of and
  below the TF coils) are also required to correct
  error fields that arise due to imperfections in
  the actual PF and TF coil configuration, and to
  stabilize the plasma against resistive wall mode
  instabilities.
• The magnet system weighs, in total, about 8,700
  t.

14
Design - Magnets and Structures (2)

• The CS and TF coils use Nb3Sn as superconductor,
  and the technology of wind, react and transfer,
  whereas the PF and correction coils use NbTi. All
  coils are cooled by supercritical helium at
  4.5K.
• The TF coil case is the main structural component
  of the magnet system and the machine core. The
  PF coils and vacuum vessel are linked to the TF
  coils such that all interaction forces are
  resisted internally in the system.
• The TF coil inboard legs are wedged all along
  their side walls in operation and they are all
  linked at their two ends to two strong coaxial
  rings which provide toroidal compression and
  resist the local de-wedging of those legs under
  load.
• At the outboard leg, the out-of-plane support is
  provided by intercoil structures integrated with
  the TF coil cases.

15
RD - CS Model Coil (L-1) (1)

• Objectives
• verify conductor performance under
  ITER-relevant conditions
• demonstrate the major steps in manufacturing
  the conductor and ITER CS coil.
• The main coil consists of two modules nested
  inside each other. To test conductor,insert
  coils can be fitted within its bore. Three
  insert coils relevant for ITER are foreseen.
• Significant advances on present superconducting
  coil manufacturing technology were required
• substantial quantities of Nb3Sn strand to a
  uniform quality
• jacketing of a cable of this strand to provide
  structural support against magnetic forces
• accurate conductor bending to the winding
  shape
• heat treatment in a controlled atmosphere,
  insulation in an unspringing process before
  stacking to form the winding, and then
  impregnation with epoxy resin.

16
RD - CS Model Coil (L-1) (2)
Outer module, manufactured by Toshiba, being
placed outside the inner module, which has
already been installed in the vacuum chamber.

• Transfer of a layer onto a coil assembly to form
  the inner module at Lockheed Martin.

17
RD - CS Model Coil (L-1) (3)
In April 2000, the maximum field of 13 T with a
cable current of 46 kA and magnetic stored energy
of 640 MJ were successfully achieved in the test
facility. Pulsed operation has been experienced
under conditions (ramp-up to 13 T at 0.4 T/s,
ramp-down at 1.2 T/s) more severe than for
ITER-FEAT operation. One insert coil has been
tested at 13 T and charged/discharged 10,000
times. More tests are underway.

• Model coil and insert coil installed at the test
  facility in JAERI Naka. In the background is the
  vacuum chamber lid.

18
RD - TF Model Coil (L-2) (1)

• Objective
• validation of design and analysis,
• demonstration of industrial manufacturing
  methods,
• testing of performance of each component
  integrated in the magnet,
• testing and demonstration of reliable
  operation.
• The model consists of a race-track shaped
  sub-size coil, about 4 m high and 3 m wide, and
  two full-size sections of the outer housing. The
  coil includes the key technical features and
  manufacturing approaches foreseen for the actual
  ITER TF coils.
• Although the conductor will not be fully tested
  for superconducting properties (done in L-1) the
  manufacturing defines appropriate tolerance
  targets, procedures and quality control steps.
  The test of the sub-size coil will create
  realistic magnetic loads to demonstrate the
  structural concept.

19
RD - TF Model Coil (L-2) (2)

• Conductor after heat treatment, opened out by
  unspringing to give space to wrap with
  insulation without damaging the superconductor.
  The insulation has been applied to the lower
  turns. (Ansaldo Energia)

Machining of the radial plate which reinforces
the conductor. The conductor is fitted into
grooves in this plate. (Mecachrome/Nöll)
20
RD - TF Model Coil (L-2) (3)

• The top of the groove is closed by a cover plate
  which is laser welded into position. (RTM)

Inner leg case section during forging as a hollow
tube, before cutting into two U sections. (Kind)
21
RD - TF Model Coil (L-2) (4)
The model coil is about to be moved to the TOSKA
facility at Karlsruhe, Germany, which has been
adapted to accommodate the coil and its test
programme.

• The coil with the final surface finish (sand
  blasted and with interface surfaces machined)
  while it is leak tested and the inlet headers are
  preassembled on top of the coil (Alstom).

22
Design - Vessel, Blanket Divertor (1)

• The double-walled vacuum vessel is lined by
  modular removable components, including blanket
  modules composed of a separate first wall mounted
  on a shield block, divertor cassettes, and
  diagnostics sensors, as well as port plugs such
  as the limiter, heating antennae, and test
  blanket modules. All these removable components
  are mechanically attached to the VV.
• These vessel and internal components absorb most
  of the radiated heat from the plasma and protect
  the magnet coils from excessive nuclear
  radiation. This shielding is accomplished by a
  combination of steel and water, the latter
  providing the necessary removal of heat from
  absorbed neutrons. A tight fitting configuration
  of the VV to the plasma aids the passive plasma
  vertical stability, and ferromagnetic material in
  the VV located under the TF coils reduces the TF
  ripple and its associated particle losses.

23
Design - Vessel, Blanket Divertor (2)

• The initial blanket acts solely as a neutron
  shield, and tritium breeding experiments are
  confined to the test blanket modules which can be
  inserted and withdrawn at radial equatorial
  ports. The blanket module design consists of a
  separate faceted first wall (FW) built with a Be
  armour and a water cooled copper heat sink
  attached to a SS shielding block. This minimises
  radioactive waste and simplifies manufacture. The
  blanket is cooled by channels mounted on the
  vessel.
• The divertor is made up of 54 cassettes. The
  target and divertor floor form a V-shape and the
  large opening between the inner and outer
  divertor legs to allow an efficient exchange of
  neutral particles. These choices provide a large
  reduction in the target peak heat load, without
  adversely affecting helium removal.
• The current design uses carbon at the vertical
  target strike points. Tungsten is being
  considered as a backup, and both materials have
  their advantages and disadvantages. The best
  judgement of the relative merits can be made by
  the time of procurement. Carbon has the best
  behaviour to withstand large power density pulses
  (ELMs, disruptions), but gives rise to tritiated
  dust. Procedures for the removal of tritium
  codeposited with carbon by a number of schemes
  are under consideration and need further
  development.

24
RD - Vacuum Vessel (L-3) (1)

• Objective
• To provide input required to complete the design,
  especially regarding critical issues of
  fabrication technology dimensional accuracy,
  welding distortions and achievable tolerances.
• The ITER vessel will be more than twice the
  linear dimensions and over 16 times the mass of
  the largest existing tokamak vessel. The key
  issues can only be properly resolved by building
  a model at full scale.
• The dominant feature of the project is a
  full-scale sector model (sized for the larger
  1998 ITER design), manufactured by the Japanese
  Home Team. Hitachi and Toshiba each built half
  sectors. The distributed manufacturing offered
  opportunities to test and compare different
  candidate weld schemes.
• After the half sectors were fabricated, they were
  leak, pressure and mechanically tested to
  determine their structural characteristics. The
  welding together of the two half sectors
  demonstrated the automatic welding techniques and
  verified the ability to undertake joint
  inspection by ultrasonic testing.
• In parallel, the Russian Federation Home Team
  manufactured a full-scale model equatorial port
  extension, developing and demonstrating
  fabrication technologies to the required
  specifications and tolerances and related
  inspection techniques and procedures. The US Home
  Team developed a fully remotized welding/cutting
  system. This technology has now been transferred
  to the Japanese Home Team and has been used to
  join the port extension to the sector model.

25
RD - Vacuum Vessel (L-3) (2)

• View of full-scale sector model of ITER vacuum
  vessel completed in September 1997 with
  dimensional accuracy of 3 mm

26
RD - Vacuum Vessel (L-3) (3)

• Field joint welding test of VV sector

27
RD - Vacuum Vessel (L-3) (4)

• Equatorial port extension shipped from RF to the
  test site at JAERI Naka for integration test

Inner shell welding demonstration using full
scale sector and port extension
The manufacture of the full-scale sector of the
1998 ITER design gives a sound basis for the
present design. To reduce the vessel fabrication
cost, forging, powder HIPing and/or casting is
being investigated particularly for the blanket
module support housings.
28
RD - Blanket Module (L-4) (1)

• Objectives
• to develop and fabricate prototype components
  for the shielding blanket, in order to assess
  their manfacturing feasibility,
• to assemble them together and develop bolting,
  welding and cutting tools for the remote removal
  of the components,
• demonstrate the performance by testing
  representative parts of the components under
  relevant conditions,
• obtain confirmation of the design choices by
  results from accompanying RD on materials,
  joining techniques and neutronics using a fast
  neutron source.
• Full scale prototypes include multi-layered first
  walls made of stainless steel (as structural
  material), copper alloy (as heat sink) and Be or
  C (as protection material), massive stainless
  steel shields and flexible supports.
• The feasibility of installation and removal of a
  blanket module with mechanical attachments has
  been demonstrated and tested in a prototype
  assembly. A hydraulic, remotely driven bolting
  tool has been developed, which achieves high
  pre-loading using heating rods. High quality
  remotized hydraulic laser-welded connections have
  also been made through a 30 mm penetration hole
  in the front of the module.

29
RD - Blanket Module (L-4) (2)

• Joining Techniques
• Be/Cu joints of high heat flux components (e.g.
  limiter) fast amorphously CuInSiNi-brazed small
  tiles on curved Cu surface (RF), withstood 4500
  cycles at 12 MW/m2.
• Be/Cu joints of lower heat flux components
  (e.g. first wall) hot isostatic pressing (HIP)
  of Be tiles with Ti interlayer at (EU), withstood
  13000 cycles at 0.7 MW/m2.
• Joining of Cu/SS parts with high
  precisionsolid-solid HIP of the first wall (e.g.
  JA) withstood 2500 cycles at up to 7 MW/m2.

30
RD - Blanket Module (L-4) (3)

31
RD - Blanket Module (L-4) (4)

• FW prototype (JA)

Flexible supports (RF)
Shield block prototype - powder HIP (EU)
Assembly test rig (EU)
In the frame of this RD, innovative technologies
have been developed and existing technologies
have been improved, giving confidence in the
feasibility and robustness of the chosen blanket
design.
Module cut for inspection (JA)
Port Limiter (RF)
32
RD - Divertor Cassette (L-5) (1)

• Objective
• To develop the technology needed to construct
  full-scale armoured components capable of
  providing adequate armour, armour-heat sink joint
  (CfC-Cu W-Cu), and heat sink lifetime, and
  sustaining thermo-hydraulic and
  electro-mechanical loads, whilst utilising the
  most cost effective and reliable manufacturing
  processes.
• Major issues include
• the bonding of different plasma facing
  materials on the same component,
• the selection of the heat sink material (CuCrZr
  now preferred), and the
• demonstration that it maintains its properties
  after manufacturing.

33
RD - Divertor Cassette (L-5) (2)

Results of CfC/Cu high heat flux component testing

• Component test results shows that various tile
  geometries can meet the ITER requirements.
  However, the monoblock has proved to be the most
  reliable with no complete detachment of tiles.
  Tungsten brush type armour proved to be a
  solution to having a Cu-W joint able to withstand
  the large difference in thermal expansion of the
  two materials under the high heat flux loads.

34
RD - Divertor Cassette (L-5) (3)

CfC monoblock and W brush armoured vertical
target (EU)

• W and Be armour fast brazing to liner CuCrZr heat
  sink (RF)

Pure Cu-clad DSCu tube armoured vertical target
with saddle block CfC and CVD-W armours (JA).
35
RD - Divertor Cassette (L-5) (4)

An additional aim of the project was to integrate
key plasma facing components together onto a
realistic prototype of the cassette body.
Following the decision of the US to pull out of
ITER, the EU has also constructed an integration
prototype. It is not essential to use all the
real materials for these prototypes, and dummy
components have been made - thermohydraulic
equivalents of the real components.

• Outboard integration mockup prior to installation
  of liner (EU)

Inboard divertor channel integration mockup
undergoing flow tests (US)
Several middle and large scale CfC and W-armoured
divertor mock-ups have been successfully tested
at heat fluxes 20 MW/m2 x 1000 cycles, which is
consistent with ITER operational needs.
36
Design - In-vessel Remote Handling
Systems near the plasma will become radioactive
and will require remote maintenance, with special
remote handling equipment. An in-vessel
transporter system is used for the removal and
reinstallation of blanket modules, multifunction
manipulators for divertor cassette removal, and
specialised manipulators to handle vacuum vessel
port plugs. Special casks, which dock
horizontally to the access ports of the vacuum
vessel, are designed to house such equipment and
to transport radioactive items from the tokamak
to the hot-cell where refurbishment or waste
disposal operations can be carried out. Docking
of these casks to the vessel and the hot cell
flanges is tight, to avoid spreading of
contamination. Hands-on assisted maintenance is
used wherever justifiable, following the ALARA
principles.

• The remote handling strategy for ITER has been
  confirmed by a comprehensive RD programme which
  has successfully demonstrated that key
  maintenance operations such as blanket and
  divertor replacement can be achieved using common
  remote handling technology.
• Several crucial issues such as vacuum vessel
  remote cutting and re-welding, viewing, materials
  and components radiation hardness have been
  addressed and demonstrated.

37
RD - Blanket Remote Handling (L-6) (1)

• Objective
• To develop and demonstrate the ability to
  remotely maintain blanket modules, including
  manipulating a 4 t module at a distance of 8 m
  with an accuracy of 2 mm. A rail-mounted
  vehicle system has been developed to handle the
  heavy blanket module within the limited space and
  with the required precision.
• After development and prototype demonstration of
  the main systems and techniques, full-scale
  testing and verification were successfully
  completed in 1998 on the Blanket Test Platform
  (BTP) constructed at JAERIs Naka laboratory.
• This platform comprises module handling
  equipment, port handling equipment, auxiliary
  remote handling tools and a blanket mock-up
  structure to reproduce the physical environment
  of a 180 ITER in-vessel region. A suppression
  control system to reduce dynamic deflection and
  vibration of the arm to negligible levels has
  subsequently been developed and successfully
  tested.
• The blanket on ITER requires very precise
  positioning ( 0.25 mm) with respect to keys and
  pins. Module insertion tests have therefore been
  carried out to check the ability to handle
  misalignments between modules and keys during
  installation. The module has been successfully
  inserted with a misalignment of 10 mm, using the
  passive compliance of the manipulator, and
  chamfered keyways. Development of a sensor-based
  control system with this positioning accuracy is
  now underway.
• Tests also show that the rail can be deployed 90
  around the torus in about 30 minutes.

38
RD - Blanket Remote Handling (L-6) (2)

• In-vessel transporter design

Rail-mounted vehicle and telescopic arm
39
RD - Blanket Remote Handling (L-6) (3)

40
RD - Divertor Remote Handling (L-7) (1)

• Objective demonstrate that the following
  operations are feasible
• replacing and refurbishing all or individual
  divertor components several times during the
  machine life
• positioning high heat flux components (HHFCs)
  so the maximum step between those on adjacent
  cassettes would be under 4 mm and the maximum
  variation around the whole torus would be within
  10 mm
• locking and securing the supports, making water
  pipe connections, assembling electrical
  connectors, and handling port plugs
• replacing all cassettes in less than 6 months
  and replacing a single cassette in under 8 weeks.
  
• Two full-scale test facilities - the Divertor
  Test Platform (DTP) and the Divertor
  Refurbishment Platform (DRP) - have been set up
  at the ENEA Research Centre of Brasimone (Italy).
  
• The key elements of the divertor maintenance
  procedure are
• radial insertion of the cassettes from the
  chamber,
• toroidal manoeuvring,
• lowering of cassettes into position on wheeled
  and jacking forks,
• remote attachment to rails.

41
RD - Divertor Remote Handling (L-7) (2)

• Tests on the DTP confirm
• the maintenance concept,
• its integration inside the vessel,
• accuracy of cassette positioning,
• adequacy of nominal gaps and tolerances,
• payload capabilities.
• Improvements are being investigated to reduce
  costs and to implement lessons learnt in the
  early tests to improve man-machine interface,
  sensors, and time, as well as to improve sliding
  components and to investigate rescue scenarios if
  components become jammed.
• The DRP is for simulating the most critical
  operations to be undertaken in the hot cell. Only
  those parts of the mock-up which are critical for
  HHFC mounting have been machined accurately.
  Tests so far show that the remote measurement
  system can be operated accurately enough (0.01
  mm) that components can be correctly machined to
  fit. A target mockup has been installed on the
  cassette with the required accuracy, but further
  work is needed to streamline procedures to
  shorten the time taken.

42
RD - Divertor Remote Handling (L-7) (3)

• Divertor Test Platform

43
RD - Divertor Remote Handling (L-7) (4)

• Divertor Refurbishment Platform