Overview of the

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Overview of the ITER Plasma Facing
Components Presented by M. Merola with thanks
to R. Pitts, A. Martin, C. Lowry, M. Pick, J.
Palmer, A. Kukushkin, A. Loarte, R. Mitteau And
the ITER Domestic Agencies
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Outline
Divertor First Wall and Blanket
3
Status of PFCs Divertor
4
Divertor System
Main Functions

• Divertor system main functions
• Exhaust the major part of the plasma thermal
  power (including alpha power)
• Minimize the helium and impurities content in
  the plasma

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Divertor System
Overall Layout and Ports
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Cassettes in a circular array held in position by
two concentric radial rails .
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Divertor System
Overall Layout and Ports
Nine equally
spaced divertor level ports, arranged at
40 degrees of each other RH Ports, Diagnostic
Ports Cryopump Ports
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Divertor System
Overall Layout and Ports
Diagnostic
Port 10
RH Port 8
Cryopump Port 12
Cryopump Port 6
RH Port 14
Cryopump Port 4
Diagnostic
Port 16
RH Port 2
Cryopump Port 18
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Divertor System
Outer Toroidal Rail
Central Cassette Rail
Support Pad
Support Pad
Central
Cassette Rail permanently installed in
Diagnostic and Cryopump Ports removable on RH
Ports
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Divertor System
Outer Toroidal Rail
10
Divertor System
Diagnostics
Diagnostic Equipment located inside three RH
Ports and two Diagnostic Ports 16 in total
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ITER Party Responsibilities
Divertor System
Diagnostics
Baseline 2006

Twenty different systems delivered by six Parties
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Divertor System
Diagnostics

• Diagnostic Integration Criteria
• 1) Preserve overall functionality of the divertor
  system, while fulfilling the
• requirements of each specific system.
• 2) Avoid as much as possible creating divertor
  cassette variants, to minimize
• manufacturing costs and maximize the flexibility
  in the management of spare
• Partsgt integrate as many diagnostic interfaces
  into the standard cassette
• design as possible
• Standard PFCs Standard Cassette Body (CB )
  45
• PFCs Variant to integrate Langmuir Probes Stand
  CB 5
• PFCs Variant to integrate Thermocouples
  Stand CB 3
• Standard PFCs CB Variant to integrate Neutron
  Camera 1

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Divertor System
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Divertor System
Cassette body
The divertor CB is reusable to minimise activated
waste it provides neutron shielding, routes the
water coolant and supports the different PFCs
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Divertor System
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Divertor System
Vertical Targets
The inner and outer vertical targets (VTs), are
the PFCs that, in their lower part, intercept the
magnetic field lines, and therefore remove the
heat load coming from plasma via conduction,
convection and radiation.
Inner VT
Outer VT
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Divertor System
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Dome
Divertor System
The inner and outer neutral Particle Reflector
Plates protect the CB from plasma radiation,
allow transient movements of the strike points
and provide improved operational flexibility of
the divertor in terms of magnetic configuration.
The Umbrella, which is located below the
separatrix, baffles neutrals, particularly helium.
Large conductance between the two divertor
channels increases the neutral densities at the
outer target and enhance radiation losses
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Materials Choice
Divertor System
In the current ITER baseline CFC at the strike
points, W on the baffles through the H and D
phases All-W from the start of DT operations

• Rationale
• Carbon easier to learn with
• Lack of melting makes it easier to test ELM and
  disruption mitigation strategies
• T-retention expected to be too high in DT phase
  with CFC targets
• But (limited) DT operations with CFC target still
  in place not excluded

W reflector plates
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CFC W Transition Point
Divertor System

• W-CFC interface moved down 10 cm
• During H D phases strike points can be placed
  on the W part
• Modelling demonstrates that a 7.5 MA plasma
  equilibrium can be established in H-mode on the W
  part of the vertical target
• Heat loads will not exceed the 5 MWm-2 for which
  the W monoblocks are qualified
• At least 2 power e-folding lengths on vert.
  target portion (lp 1.0 cm scaled with 1/Ip from
  0.5 cm _at_15 MA)

Reasonable compromise between easier start with
CFC and gaining experience with W as soon as
possible in preparation for DT (with W).
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Flexibility built-in in the Design
Divertor System
1
2
3

• 1. Radial width of IVT increased2. Position of
  inner dome multi-link3. Position of dome cooling
  pipes
• All modified to maximise flexibility

IVT can be shifted back and forth by ?5? if
required in future designs
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Alignment of PFCs
Divertor System
The PFCs shall be angled to avoid exposing the
leading edges of the armour to the Scrape-Off
Layer (SOL), otherwise the near normal incidence
of the SOL on these edges would cause large
amounts of carbon to be evaporated (or tungsten
melted) with the inherent risk of poisoning of
the plasma and/or inducing a critical heat flux
event in the water coolant. A nominal step in
the toroidal direction between adjacent targets
of 3 mm is taken as a requirement
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Divertor System
Supporting analysis

• Divertor is being subject to final rigorous
  analysis
• Time averaged and transient loads (plasma fluxes
  and radiation (specified by physics) updated and
  being used for final thermal analysis
• Detailed electromagnetic FE modelling (VDE
  forces)
• Hydraulic analysis of whole cassette (cooling
  water pressure drops, draining and dry)

Example for predicted transient heat loads(loads
in red permissible up to 10 s, in yellow up to 2
s)
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Design Values
Divertor System

• The PFCs of the first divertor set shall
  withstand 3000 equivalent pulses of 400 s
  duration at nominal parameters, including 300
  slow transients
• During normal operational conditions
• vertical target has a design surface heat flux up
  to 10 MW/m2 (strike point region) and 5 MW/m2
  (baffle region)
• Under slow transient thermal loading conditions
• lower divertor vertical target geometry has a
  design surface heat flux up to 20 MW/m2 for
  sub-pulses of less than 10 s
• The dome shall sustain heat fluxes of up to 5
  MW/m2
• The umbrella and the particle reflector plates
  shall sustain local heat flux up to 10 MW/m2,
  which can be transiently swept across the surface
  (about 2 s) as the plasma is returned to its
  correct position

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Power Handling
Comparisons
HIGH HEAT FLUX COMPONENTS FOSSILE FIRED BOILER WALL (ABB) FISSION REACTOR (PWR) CORE ITER DIVERTOR
DESIGN 12/15 mm ID/OD
HEAT FLUX average MW/m2 - maximum MW/m2 0.2 0.3 0.7 1.5 3 5 10 20
Max heat load MJ/m2 Lifetime years Nr. of full load cycles Neutron damage dpa Materials - 25 8000 - Ferritic-Martens. steel - 4 10 10 Zircaloy - 4 10 5-8 3000 - 16000 0.2 CuCrZr CFC/W
Coolant - pressure MPa - temperature C - velocity m/s - leak rate g/s Water-Steam 28 280-600 3 lt50 Water 15 285-325 5 lt50(SG) Water 4 100 150 9 11 lt10-7
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Divertor RH
Replacements

• The divertor shall be designed for relatively
  frequent (3 times during the 20 years of ITER
  operation) fully remote assembly and disassembly
• 6 months is the maximum (that is, the target)
  period that shall be allocated for cassette
  refurbishment/replacement (from divertor port
  primary closure plate opening to its closure).
  2-month allowance for machine shutdown and
  start-up time should be added to this period.
• The replacement time of a single faulty cassette
  will depend on its toroidal position and, on the
  average, shall not exceed 2 months
• Two sets of divertor cassettes shall be procured
  to allow the off-line refurbishment

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Maintenance Strategy
Divertor RH

• Main feature
• Four divertor sets ? Three divertor
  replacements

2
1
3
4
PFCs new CBs new (set 2)
PFCs new CBs set 1 5 diagn cass new
PFCs new CBs new (set 1)
PFCs new CBs set 2 5 diagn cass new
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RH Equipments
Divertor RH

• Divertor RH equipment is comprised of
• Two main types of cassette mover
• Cassette Multifunctionl Mover (CMM)
• Cassette Toroidal Mover (CTM)
• Each are to be equipped with a dexterous
  manipulator arm and RH tooling.

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Divertor RH
Divertor Test Platform 2
CMM with Second Cassette End-Effector

• Objective
• To demonstrate in-vessel cassette handling and
  divertor maintenance.
• Details
• To be located in Tampere (Finland) and hosted by
  the TEKES fusion Association.
• Comprises a 27o mock-up of the ITER divertor
  region with one radial access port.
• To initially include
• CMM prototype
• Second cassette mock-up
• Later extensions to include
• Cassette Toroidal Mover (CTM)
• prototype
• standard central cassettes
• divertor cooling pipes

DTP2
 
TECHNICAL RESEARCH CENTRE OF FINLAND
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Divertor Remote Handling
DRP
Divertor Refurbishment Platform
Objective To demonstrate and refine hot-cell
cassette maintenance operations. Details Located
in Brasimone (Italy) and hosted by the ENEA
fusion Association. Comprises a remote work cell
equipped with light dexterous manipulators,
robotic tool / component transporters and
component refurbishment tooling. Visual feedback
via video link (no direct viewing) in keeping
with the ITER hot-cell concept.
 
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Divertor Remote Handling
Main Challenges

• Characteristics
• Very tight cassette-VV clearance often 25mm
• Almost blind operations (for radial transport)
• ITER will be dark !
• There must be a heavy reliance on virtual
  techniques and extremely good correlation of
  virtual models with as-built data.

At 5m, 25mm 0.3 on the lift axis joint angle.
25mm
Hinge
Selected resolvers have a fundamental accuracy of
30 arc-seconds.
x
5m
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Thermo-Hydraulics
Coolant Flow Path
Divertor coolant design parameters Inlet
temperature 100 C Inlet water pressure
4.2 MPa Total pressure drop lt 1.4 MPa CHF
margin gt 1.4 Total flow rate
lt 1000 kg/s
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Experiments
Thermo-Hydraulics
Pressure drop vs.flow rate have been measured on
Outer and Inner Vertical target and Dome (ENEA
Brasimone)
Hydraulic testing of DOME
Hydraulic testing of IVT
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Experiments
Thermo-Hydraulics
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Overall Divertor Pressure Drop (MPa)
Inner VT 0.51
Outer VT 0.32
Dome 0.73
Cassette 0.002
TOTAL 1.56

• Total pressure drop 1. 56 lt 1.6 MPa
• Total flow rate 934 lt 1000 kg/s

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Critical Heat Flux
Critical heat flux tests on monoblocks
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Terminology Flat Tile and Monoblock
HHF Technologies
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HHF Technologies
FT
M
M
FT
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HHF Technologies
Thermal expansion at 300 C
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HHF Technologies
Monoblocks
CFC or W
CuCrZr tube
AMC (CFC) Brazing (CFC) Casting
HIPing Brazing Hot Radial Pressing
Cu interlayer
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HHF Technologies
Flat Tiles
Armour to heat sink joints
Cu casting Brazing
Tungsten
Pure copper interlayer
HIPing Hot Radial Pressing Brazing EB welding
CuCrZr
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Vertical TargetMedium-Scale Prototype
Test results

• W macrobrush
• 15 MW/m2 x 1000 cycles
• CFC monoblock
• 20 MW/m2 x 2000 cycles
• CHF test gt 30 MW/m2

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Vertical TargetFull-Scale Prototype

• W monoblocks
• 10 MW/m2 x 1000 cycles
• CFC monoblock
• 10 MW/m2 x 1000 cycles
• 20 MW/m2 x 1000 cycles
• 23 MW/m2 x 1000 cycles

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• Brazing using TiCuNi and CuNiMn alloys
• 15 MW/m2 x 3000 cycles 20 MW/m2 x 1000 cycles

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Other mockups
Type 2
Type 1
Type 3
Type 4
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HHF test at 20 MW/m2 (19-25 MW/m2) of the type-4
mockup

• Test facility JEBIS
• Flow rate approx. 70 L/min ( 10m/s) at 4 MPa

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600 mm

• W-Cu by casting
• Cu-CuCrZr by CuInSnNi (STEMET 1108) brazing
• 18.5 MW/m2 x 1000 cycles

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The mock-up has survived 1000 thermal cycles at
5 MW/m2 500 thermal cycles at 6
MW/m2 500 thermal cycles at 7 MW/m2 No
overheating, no visible detachments After
testing the mock-up was destructively examined
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Divertor Qualification Prototypes
A qualification is needed for the critical
procurement packages shared by multi-Parties,
including the divertor
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Divertor Qualification Prototypes
Design Value (Incident heat flux) Test results (Absorbed heat flux) Test results (Absorbed heat flux) Test results (Absorbed heat flux)
CFC 300 cycles at 20 MW/m2 1000 cycles 10 MW/m2 987 cycles gt20 MW/m2 1000 cycles 10 MW/m2 1000 cycles 20 MW/m2 N/A
W 1000 cycles at 5 MW/m2 1000 cycles 3 MW/m2 1000 cycles 5 MW/m2 1000 cycles 3 MW/m2 1000 cycles 5 MW/m2 1000 cycles 3 MW/m2 1000 cycles 5 MW/m2 348 cycles 10 MW/m2
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Vertical Targets
Plasma-Facing Components
W monoblock
5 MW / m2
XM-19
316L(N)-IG
10 MW / m2
20 MW / m2 10 sec
XM-19
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Vertical Targets
Plasma-Facing Components
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Dome
Plasma-Facing Components
W Flat Tiles
5 MW / m2
316L(N)
XM-19
316L pipes
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Acceptance Criteria
Plasma-Facing Components
Manufacturing of more than 100 mock-ups with
artificial defects
High heat flux test of mock-ups with artificial
defects
Non-destructive and destructive examinations of
mock-ups
Final definition of the divertor acceptance
criteria
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Acceptance Criteria
Plasma-Facing Components
SamplesThere are 112 samples split in two
batches - 56 HIPing Technology ? Plansee-
56Hot Radial Pressing Technology ? Ansaldo
RicercheEach batch of 56 samples includes -
28 CFC monoblocks 26 short 2 high- 14
W monoblocks- 14 W flat tiles
CFC monoblock (Ansaldo)
W monoblock (Ansaldo)
CFC monoblock (Plansee)
W flat tile (Plansee)
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Status of PFCs First Wall and Blanket
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Caveat Divertor design and plans for
procurement are far more advanced than for the
first wall. It is not possible at this time to
give more than outline indications of FW design
progress
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Functions
FW Blanket

• To limit the neutron fluence to the Vacuum
  Vessel and Coils
• To be remotely installed, repaired and exchanged
• Simple and reliable attachments
• Minimise number of operations
• To exhaust incident plasma exhaust power
• Active cooling technology
• Plasma compatible materials

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Procurement Sharing
FW Blanket
Procurement sharing

• Almost all Parties involved in first wall and
  blanket procurement
• Makes the problem harder from a design and
  organizational point of view
• ? Simplification of the procurement scheme shall
  be urgently attempted !

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Power Handling Capability
FW Blanket
2001 Requirements
2008 Requirements
Power flux and disruption halo current following
field lines Interactions deep into the scrape-off
layer
0.5 MWm-2 Surface flux Disruption halo current
defined as surface current density
No particular shaping requirements Medium heat
flux technology
Attention to shaping required Higher heat flux
technology required
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Thermal Load Specifications
FW Blanket

• First wall power loadings broken down into
  several categories
• Limiter start-up and ramp-down
• Steady state (inter-ELM) heat fluxes
• ELM transient heat fluxes
• Plasma movement onto wall in response to
  confinement transients
• All reported in detailed thermal load
  specification document recently agreed into ITER
  Baseline (PCR-093)
• Derived from extrapolation of results from
  present machines all expressed in terms of
  midplane parallel heat fluxes

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Old Baseline Design
FW Blanket
First wall panel
First Wall leg welded at back of shield module
Vacuum vessel
Shield module
Central support leg to allow for thermal expansion
Key to react radial moment
Flexible cartridge to allow for differential
expansion
4 individual First Wall Panels to reduce induced
currents

• Number of issues identified by 2007 design
  review
• Very small (30 mm) access holes for cutting and
  re-welding of hydraulic connectors ? PFC
  changeout extremely difficult (if not impossible)
  in-vessel
• FW panels toroidally flat on LFS and top, faceted
  on HFS ? no leading edge protection (not required
  on LFS due to favourable plasma curvature)
• Relied on less than 2 mm possible misalignment
  between adjacent panels

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New Design Concept
FW Blanket

• Design driver minimise changes to VV interface
• VV is long lead item manufacture must start as
  early as possible

Shield block semi-permanent item leave
interfaces to VV as fixed as possible
Single FW panel removable in-vessel Reduces RH
operations, better power handling Horizontal
fingers reduce halo loads and eddy
currents Toroidal shaping to protect leading
edges
Poloidal Shaping Allow good access for RH Shadow
leading edges
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New Design Concept
FW Blanket
Complexity arises on LFS due to numerous
asymmetries set by diagnostic ports, heating
systems, test blanket modules, ripple ? maintain
identical FW panels but radially advance poloidal
bands in between ports
Parallel power flux reduced drastically as more
limiters used But start-up and higher potential
heat fluxes near 2nd X-point may impose variable
cooling technology at different locations
3.0 MW/m2
1.5 MW/m2
Poloidal strips of 9 blanket modules advanced by
5 mm to protect port region
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Shaping Calculations
FW Blanket

• Considerable efforts now being expended on FW
  panel shape optimisation
• Semi-analytic work (P. C. Stangeby see PSI
  Conf. 2008)
• Sophisticated numerical field line tracing (R.
  Mitteau)
• Preliminary benchmark of the two methods has been
  successful
• Only toroidal shaping required on LFS to protect
  central hole (curvature of VV provides
  self-shielding for poloidal edges)
• Toroidal poloidal shaping required on HFS
• Design study for upper dump region underway

Port
R. Mitteau
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Technology
FW Blanket
Chen et al., IAEA 2008
Geometry Max Heat Flux(MWm-2) Tsurf _at_ at max Heat Flux Overageing of Cu (400C) Fatigue life (3x105 cycles)
SS tube (1mm) in Cu 1.8 (8kgs-1) 700C 1.3 MWm-2 1 MWm-2
SS tube (0.5mm) in Cu 1.8 (8kgs-1) 590C 2 MWm-2 2 MWm-2
Cu tube in Cu 1.8 (8kgs-1)3 (12kgs-1) 530C(700C _at_ 2.8 MWm-2) 2.5 MWm-2 3.0 MWm-2
Hypervapotron 8.0 (8kgs-1) 700C _at_ 4.5 MWm-2 8 MWm-2
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Conclusions

• Divertor Design is completed
• Extensive integration work carried out on the
  interfacing systems
• Extensive RD has been carried out by EU, JA, RF
  DAs
• All concerned DAs have demonstrated the technical
  capability to manufacture divertor components
  with adequate heat flux performance
• Divertor design process has progressed over the
  years with constant consideration for the
  maintenance process and close interaction with RH
  equipment / process developers. However, RH
  remains a challenge.
• First Wall and Shield Blanket design has been
  modified with respect to the 2001 baseline
• A First Wall shape is being developed which both
  shadows leading edges, and provides for a
  generous RH access aperture
• Different design solutions may be required
  depending from the toroidal and poloidal position
  of the modules
• High heat flux technology is required in some
  regions, but removes the need for start-up
  limiters
• The complex procurement sharing adds a further
  challenge