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Philippe Chappuis IO Blanket Lead Engineer

An OVERVIEW of the ITER INVESSEL COMPONENTS. Philippe Chappuis IO Blanket Lead Engineer On Behalf of the ITER Internal Component Division & the Blanket Integrated Product Team. Content. Update on the ITER Project Introduction on In vessel components The Blanket System The Divertor

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Philippe Chappuis IO Blanket Lead Engineer

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  1. An OVERVIEW of the ITER INVESSEL COMPONENTS Philippe Chappuis IO Blanket Lead Engineer On Behalf of the ITER Internal Component Division & the Blanket Integrated Product Team

  2. Content • Update on the ITER Project • Introduction on In vessel components • The Blanket System • The Divertor • The in vessel Coil System • Conclusion

  3. Update on ITER

  4. The future ITER site Tokamak & Assy building – 6 levels @ 166 m x 81 m x 57 m high (~36,000 m2) Tritium building – 7 levels @ 25 m x 80 m (~14000 m2) Largest throughput in world (~300 kg/yr). Cryo-plant Electric Supply Staff Offices Cooling Towers Hot cell 60 m x 70 m Parking 39Buildings, 180 hectares

  5. On 22nd December 2010 ….. 2.5 million m3 of earth levelled Construction beginning of first buildings on the ITER platform….

  6. On 30th September 2011 ….. The poloidal field coil winding building – well underway The building is approximately 257 meters long, 45 meters wide and 18 meters high

  7. Site Construction Progress: Tokamak Complex • 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;

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

  9. The ITER Machine Central Solenoid Nb3Sn, 6 modules Cryostat Thermal shields Cryostat 24 m high x 28 m dia. Toroidal Field Coil Nb3Sn, 18, wedged Vacuum Vessel 9 sectors Blanket 440 modules Port Plug heating/current drive, test blankets limiters/RH diagnostics Torus Cryopumps, Divertor 54 cassettes 3/4 PA signed Poloidal Field Coil NbTi, 6 Correction Coils NbTi, 18 Feeders NbTi, 31 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

  10. Overall Project Schedule for 2020 First Plasma

  11. Introduction on In vessel components

  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 In Vessel coils Divertor

  13. Blankets The blanket system

  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.

  15. Blanket System Modules 7-10 Modules 11-18 Modules 1-6 Shield Block (semi-permanent) FW Panel (separable) Blanket Module 440 modules covering 620 m² 50% 50% 50% 40% 10% ~850 – 1240 mm 4.5 t ~1240 – 2000 mm

  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

  17. Design of First Wall Panel I-shaped beam to accommodate poloidal torque

  18. First Wall Finger Design 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) SS Back Plate Be tiles Be tiles SS Pipes CuCrZr Alloy

  19. Shield Block Design • 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.

  20. Blanket Design and Neutronic Shielding Issue Current reference case with thickened inboard modules 17 kW Straighten Inboard Modules (+2.5 kW ± 1 kW) ~19.5 kW Reduce gaps from 10-14 mm to 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

  21. VACUUM VESSEL BLANKETS MANIFOLD DIVERTORS IN VESSEL COILS Blanket attachments

  22. Shield Block Attachment

  23. 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.

  24. Blanket Remote Handling • 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. On-Rail Module Transporter

  25. The Divertor Divertor

  26. Divertor Design 5MW/M² 5MW/M² 10MW/M² Shield Block Attachment Shield Block Attachment CFC and W: armour XM-19: all multilinks (lugs and links)C63200: hollow pins of multilinks316L(N)-IG: support structures316L pipe: steel pipesCuCrZr-IG: heat sinkSteel 660: bolts

  27. Divertor Status • 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 HHF testing of Plasma Facing Units

  28. Divertor Qualification Prototypes • CFC Armoured Areas • 1000 cycles at 10 MW/m2 • 1000 cycles at 20 MW/m2 • W Armoured Areas • 1000 cycles at 3 MW/m2 • 1000 cycles at 5 MW/m2 All 3 Domestic Agencies have been qualified.

  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

  30. Status of W Technology R&D in EU 2000 cycles at 15 MW/m2 on W Unirradiated- 1000 cycles x 20 MW/m2 – no failure 200°C, 0.1 and 0.5 dpa in tungsten - Successfully tested up to 18 MW/m2 Most of all the W repaired monoblocks behaved like not-repaired ones

  31. The In Vessel Coils In Vessel coils

  32. The In Vessel Coils system - 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

  33. Design of In-Vessel Coils Stainless 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. 316 L(N) jacket Copper Mineral insulation Coolant passage

  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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