11th AFPA ITER In vessel Components P. Chappuis Guilin 1st to 4th Nov. 2011 Slide 1
Philippe ChappuisIO Blanket Lead Engineer
On Behalf of the ITER Internal Component Division & the Blanket Integrated Product Team
An OVERVIEW of the ITER INVESSEL COMPONENTS
11th AFPA ITER In vessel Components P. Chappuis Guilin 1st to 4th Nov. 2011 Slide 2
Content
Update on the ITER Project
Introduction on In vessel components
The Blanket System
The Divertor
The in vessel Coil System
Conclusion
11th AFPA ITER In vessel Components P. Chappuis Guilin 1st to 4th Nov. 2011 Slide 4
The future ITER site
Staff Offices
Electric Supply
Parking
CoolingTowers
Cryo-plantTritium building – 7 levels @ 25 m x 80 m (~14000 m2) Largest throughput in world (~300 kg/yr).
Tokamak & Assy building – 6 levels @ 166 m x 81 m x 57 m high (~36,000 m2)
Hot cell 60 m x 70 m
39 Buildings, 180 hectares
11th AFPA ITER In vessel Components P. Chappuis Guilin 1st to 4th Nov. 2011 Slide 5
Construction beginning of first buildings on the ITER platform….
On 22nd December 2010 …..
2.5 million m3 of earth levelled
11th AFPA ITER In vessel Components P. Chappuis Guilin 1st to 4th Nov. 2011 Slide 6
On 30th September 2011 …..
The poloidal field coil winding building – well underwayThe building is approximately 257 meters long, 45 meters wide and 18 meters high
11th AFPA ITER In vessel Components P. Chappuis Guilin 1st to 4th Nov. 2011 Slide 7
• Excavation for the Tokamak Complex has been completed in summer 2011 and the associated concrete works have commenced; this work, which includes the seismic isolation plinths and the upper basemat, will continue into 2012;
• The concrete for the lower floor (B2 level) of the Tokamak building will commence during the spring of 2012;
Site Construction Progress: Tokamak Complex
11th AFPA ITER In vessel Components P. Chappuis Guilin 1st to 4th Nov. 2011 Slide 8
On 30th September 2011 …..Construction of office building headquarters well under way – To be delivered summer 2012
11th AFPA ITER In vessel Components P. Chappuis Guilin 1st to 4th Nov. 2011 Slide 9
Toroidal Field CoilNb3Sn, 18, wedged
Central SolenoidNb3Sn, 6 modules
Poloidal Field CoilNbTi, 6
Vacuum Vessel9 sectors
Port Plug heating/current drive, test blanketslimiters/RHdiagnostics
Cryostat24 m high x 28 m dia.
Blanket440 modules
Torus Cryopumps, Major plasma radius 6.2 m
Plasma Volume: 840 m3
Plasma Current: 15 MA
Typical Density: 1020 m-3
Typical Temperature: 20 keV
Fusion Power: 500 MW
Machine mass: 23,350 t (cryostat + VV + magnets)- shielding, divertor and manifolds: 7945 t + 1060 port plugs- magnet systems: 10150 t; cryostat: 820 t
Divertor54 cassettes
Correction CoilsNbTi, 18
FeedersNbTi, 31
Cryostat Thermal shields
The ITER Machine
3/4 PA signed
11th AFPA ITER In vessel Components P. Chappuis Guilin 1st to 4th Nov. 2011 Slide 10
Overall Project Schedule for 2020 First Plasma
11th AFPA ITER In vessel Components P. Chappuis Guilin 1st to 4th Nov. 2011 Slide 11
Introduction on In vessel components
11th AFPA ITER In vessel Components P. Chappuis Guilin 1st to 4th Nov. 2011 Slide 12
ITER In-Vessel Components• Divertor and Blanket directly face the thermonuclear plasma and cover an area of about 210 + 620 m2, respectively. • All these removable components are mechanically attached to the Vacuum Vessel or Vessel Ports.• Max heat released in the PFCs during nominal pulsed operation: 847 MW
– 660 MW nuclear power– 110 MW alpha heating– 77 additional heating
• Removed by three independent water loops (~1200 ks/s each) for the blanket + port plugs and one loop for the divertor (~1000 kg/s), at 3 MPa water pressure, ~70 °C•Max Power to Blanket 704 MW•Max power to Divertor 204 MW•All components design following SDIC
Blanket
Divertor
In Vessel coils
11th AFPA ITER In vessel Components P. Chappuis Guilin 1st to 4th Nov. 2011 Slide 13
The blanket system
Blankets
11th AFPA ITER In vessel Components P. Chappuis Guilin 1st to 4th Nov. 2011 Slide 14
Blanket System Functions
Main functions of ITER Blanket System:
•Exhaust the majority of the plasma power.
•Contribute in providing neutron shielding to superconducting coils.
•Provide limiting surfaces that define the plasma boundary during startup and shutdown.
11th AFPA ITER In vessel Components P. Chappuis Guilin 1st to 4th Nov. 2011 Slide 15
~1240 – 2000 mm
~850
– 1
240
mm
Shield Block (semi-permanent) FW Panel (separable) Blanket Module50% 50% 50% 40%10%
Blanket System
440 modules covering 620 m²
4.5 t
11th AFPA ITER In vessel Components P. Chappuis Guilin 1st to 4th Nov. 2011 Slide 16
Blanket Design
• Major evolution since the ITER design review of 2007- Need to account for large plasma heat fluxes to the first wall
- Replacement of port limiter by first wall poloidal limiters - Shaped first wall
- Need for efficient maintenance of first wall components. - Full replacement of FW at least once over ITER lifetime - Remote Handling Class 1
• Design change presented at the Conceptual Design Review (CDR) in February 2010 and accepted in the ITER baseline in May 2010.
• Post-CDR effort focused on resolving key issues from CDR, particularly on improving the design of the first wall and shield block attachments to better accommodate the anticipated electromagnetic (EM) loads.
• Present Design is mature, to be presented at PDR 29-30 November 2011
11th AFPA ITER In vessel Components P. Chappuis Guilin 1st to 4th Nov. 2011 Slide 17
I-shaped beam to accommodate poloidal torque
Design of First Wall Panel
11th AFPA ITER In vessel Components P. Chappuis Guilin 1st to 4th Nov. 2011 Slide 18
18
First Wall Finger Design
SS Back Plate
CuCrZr AlloySS Pipes
Be tiles
Be tiles
Normal Heat Flux Finger:• q’’ = ~ 1-2 MW/m2 • Steel Cooling Pipes• HIP’ing
Enhanced Heat Flux Finger:• q’’ < ~ 5 MW/m2 • Hypervapotron• Explosion bonding (SS/CuCrZr) + brazing (Be/CuCrZr)
11th AFPA ITER In vessel Components P. Chappuis Guilin 1st to 4th Nov. 2011 Slide 1919
Shield Block Design
260
280
300
320
340
360
380
0.7 0.75 0.8 0.85 0.9 0.95 1
Volume fraction of SS in blanket shield block
Inbo
ard
TF
coil
nucl
ear
heat
(#1
-#14
)W
/leg
New Mix, Water30%- 0.95g/ cc (fendl2.1)
NAR,Water16%- 0.9g/ cc(fen1)Water16%- 0.9g/ cc(fen2)
Water16%- 0.9g/ cc(fen2.1)
• Slits to reduce EM loads and minimize thermal expansion and bowing • Poloidal coolant arrangement• Cooling holes are optimized for Water/SS ratio (Improving nuclear shielding
performance)• Cut-outs at the back to accommodate many interfaces (Manifod, Attachment,
In-Vessel Coils)• Basic fabrication method from either a single or multiple-forged steel blocks
and includes drilling of holes, welding of cover plates of water headers, and final machining of the interfaces.
11th AFPA ITER In vessel Components P. Chappuis Guilin 1st to 4th Nov. 2011 Slide 20
Current reference case with thickened inboard modules17 kW
Straighten Inboard Modules (+2.5 kW ± 1 kW)~19.5 kW
Reduce gaps from 10-14 mmto 8 mm in the inboard(-0.75kW ± 0.25 kW)~18.7 kW
Increase 3 cm inboard Thickness(-4 kW ± 1 kW)
~14.7 kW ±2.25 kW
Blanket Design and Neutronic Shielding Issue
11th AFPA ITER In vessel Components P. Chappuis Guilin 1st to 4th Nov. 2011 Slide 21
21
BLANKETS
DIVERTORS
MANIFOLD
Blanket attachments
VACUUM VESSEL
IN VESSEL COILS
11th AFPA ITER In vessel Components P. Chappuis Guilin 1st to 4th Nov. 2011 Slide 22
Shield Block Attachment
11th AFPA ITER In vessel Components P. Chappuis Guilin 1st to 4th Nov. 2011 Slide 2323
Supporting R&D• A detailed R&D program has been planned in support of the design, covering a range of key topics, including: - Critical heat flux (CHF) tests on FW mock-ups - Experimental determination of the behavior of the attachment and insulating layer under prototypical conditions- Material testing under irradiation - Demonstration of the different remote handling procedures
• A major goal of the R&D effort is to converge on a qualification program for the SB and FW panels - Full-scale SB prototypes (KODA and CNDA) - FW semi-prototypes (EUDA for the NHF FW Panels, and RFDA and CNDA for the EHF First Wall Panels).- The primary objective of the qualification program is to demonstrate
that: - Supplying DA can provide FW and SB components of
acceptable quality.
- Components are capable of successfully passing the formal test
program including heat flux tests in the case of the FW panel.
11th AFPA ITER In vessel Components P. Chappuis Guilin 1st to 4th Nov. 2011 Slide 24
Blanket Remote Handling
On-Rail Module Transporter
- Shield blocks designed for ITER lifetime (semi-permanent component)- First wall panels to be replaced at least once during ITER lifetime (designed
for 15,000 cycles).- Both are designed for remote handling replacement (FW: RH Class 1).- Blanket RH system procured by JA DA.- RH R&D underway.
11th AFPA ITER In vessel Components P. Chappuis Guilin 1st to 4th Nov. 2011 Slide 25
The Divertor
Divertor
11th AFPA ITER In vessel Components P. Chappuis Guilin 1st to 4th Nov. 2011 Slide 26
CFC and W: armourXM-19: all multilinks (lugs and links)C63200: hollow pins of multilinks316L(N)-IG: support structures316L pipe: steel pipesCuCrZr-IG: heat sinkSteel 660: bolts
5MW/M²10MW/M²
5MW/M²
Shield Block AttachmentShield Block Attachment
Divertor Design
11th AFPA ITER In vessel Components P. Chappuis Guilin 1st to 4th Nov. 2011 Slide 27
• First Divertor (CFC/W) is well into procurement phase (5 PAs)
− PFCs: Last PA signed March 2010. Definition of QA for all parties done. Preparation for prototype manufacturing.
− HHF Testing facility in RFDA: PA signed March 2010. Commissioning planned end 2011.
− Cassette Body and integration: PA signature planned early 2012
Divertor Status
HHF testing of Plasma Facing Units
11th AFPA ITER In vessel Components P. Chappuis Guilin 1st to 4th Nov. 2011 Slide 28
400 mm
Tungsten
CFC
All 3 Domestic Agencies have been qualified.
CFC Armoured Areas1000 cycles at 10 MW/m2
1000 cycles at 20 MW/m2
W Armoured Areas1000 cycles at 3 MW/m2
1000 cycles at 5 MW/m2
Divertor Qualification Prototypes
11th AFPA ITER In vessel Components P. Chappuis Guilin 1st to 4th Nov. 2011 Slide 29
Divertor: Proposal to Start with a full-W armour
The original strategy was to start with a CFC/W divertor to be replaced with a full-W divertor before the nuclear phase.
This strategy, which posed the lowest physics risks, requires the start of the construction of the 2nd divertor set already during the construction phase of ITER.
This can not be afforded any more.
Either we extend operation of the CFC/W divertor into the nuclear phase of we start with a full-W divertor.
• The ITER licensing process does not foresee operation with CFC at the divertor target plates during the nuclear phase.
• A very important part of the licensing process, the Public Enquiry, has just been finished. • Even a very limited operational period in DT on a CFC divertor, as has been proposed by STAC for the IO to
consider in several occasions would require a modified safety file and a new public enquiry, with very uncertain results
• A strategy which assumes a simple continuation into DT operation with CFC, even for limited time is therefore impossible as things stand. A very important advantage is the significant cost savings of ~400 MEuro which provides a large proportion of the
budget for investments including deferred procurement during the first 5 years of operation budget
11th AFPA ITER In vessel Components P. Chappuis Guilin 1st to 4th Nov. 2011 Slide 30
Status of W Technology R&D in EU
2000 cycles at 15 MW/m2 on W
Most of all the W repaired monoblocks behaved like not-repaired ones
200°C, 0.1 and 0.5 dpa in tungsten- Successfully tested up to 18 MW/m2
Unirradiated- 1000 cycles x 20 MW/m2 – no failure
11th AFPA ITER In vessel Components P. Chappuis Guilin 1st to 4th Nov. 2011 Slide 31
The In Vessel Coils
In Vessel coils
11th AFPA ITER In vessel Components P. Chappuis Guilin 1st to 4th Nov. 2011 Slide 32
- Coils are placed as close as possible to the VV wall (leaving a gap of 20 mm for in-vessel diagnostic routing).- Outboard poloidal blanket manifold is routed over the ELM coils.- Blankets located over blanket manifold- Blankets possess cut-outs to accommodate the coils and the manifolds
In-Vessel Coils Blanket Manifold
The In Vessel Coils system
11th AFPA ITER In vessel Components P. Chappuis Guilin 1st to 4th Nov. 2011 Slide 33
Design of In-Vessel CoilsStainless steel jacketed mineral insulated cable will used for both the
ELM & VS coils. Both have the same outer jacket but VS coils will use thicker insulation (5 mm instead of 2.5 mm) and Cu instead of CuCrZr.
ELM VSSS OD 59 mm 59 mmSS ID 55 mm 55 mmCu OD 50 mm 45mmCu ID 33.3 mm 30 mmMineral Insulation thickness
2.5 mm 5mm
Copper alloy C18150 C10700
Copper
Mineral insulationCoolant passage
316 L(N) jacket
11th AFPA ITER In vessel Components P. Chappuis Guilin 1st to 4th Nov. 2011 Slide 34
Conclusions
• ITER project is on track following a new optimized schedule (SMP) associated to the Japan earthquake
• Blanket system is ready for PDR with design by Analysis following SDCIC& aiming at PA in 2013
• Divertor is in procurement phase but a full Tungsten design will delay the initial planning
• In vessel Coils design is stable and are being implemented behind the Blankets with improves manifolds
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