Philippe Chappuis IO Blanket Lead Engineer

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


Content

Update on the ITER Project

Introduction on In vessel components

The Blanket System

The Divertor

The in vessel Coil System

Conclusion


Update on ITER


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


Construction beginning of first buildings on the ITER platform….

On 22nd December 2010 …..

2.5 million m3 of earth levelled


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


• 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


On 30th September 2011 …..Construction of office building headquarters well under way – To be delivered summer 2012


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


Overall Project Schedule for 2020 First Plasma


Introduction on In vessel components


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


The blanket system

Blankets


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.


~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


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


I-shaped beam to accommodate poloidal torque

Design of First Wall Panel


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)


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.


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


21

BLANKETS

DIVERTORS

MANIFOLD

Blanket attachments

VACUUM VESSEL

IN VESSEL COILS


Shield Block Attachment


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.


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.


The Divertor

Divertor


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


• 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


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


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


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


The In Vessel Coils

In Vessel coils


- 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


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


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

Philippe Chappuis IO Blanket Lead Engineer

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