isfnt-11, barcelona, spain, september 16-20, 2013 © 2013, iter organization slide 1 de sign,...

ISFNT-11, Barcelona, Spain, September 16-20, 2013 © 2013, ITER Organization

Slide 1

Design, Fabrication and Testing of the ITER First Wall and Shielding Blanket

Presented by A. René RaffrayBlanket Section Leader; Blanket Integrated Product Team Leader

ITER Organization, Cadarache, France

With contributions from B. Calcagno1, P. Chappuis1, G. Dellopoulos2, R. Eaton1, Zhang Fu1, A. Gervash6, Chen Jiming3, D-H. Kim1, S-W.Kim4, S. Khomiakov5, A. Labusov6, A. Martin1, M. Merola1, R. Mitteau1, M. Ulrickson7, F. Zacchia2

and all BIPT contributors

1ITER Organization; 2F4E, EU ITER Domestic Agency; 3SWIP, China ITER Domestic Agency; 4NFRI, ITER Korea; 5NIKIET, RF ITER Domestic Agency; 6Efremov, RF ITER Domestic Agency; 7SNL , US ITER

Domestic Agency;

11th International Symposium on Fusion Nuclear Technology, Barcelona, Spain, Sept. 16-20, 2013The views and opinions expressed herein do not necessarily reflect those of the ITER Organization


Slide 2

Blanket Effort Conducted within BIPT

Blanket Integrated Product Team

ITER Organization

DA’s-

CN-

EU-

KO-

RF-

US

• Include resources from Domestic Agencies to help in major design and analysis effort.

• Direct involvement of procuring DA’s in design- Sense of design ownership- Would facilitate procurement


Slide 3

Blanket System FunctionsMain functions of ITER Blanket System:

• Contribute in absorbing radiation and particle heat fluxes from the plasma

• Contribute in providing shielding to reduce heat and neutron loads in the vacuum vessel and ex-vessel components

• Provide a plasma-facing surface which is designed for a low influx of impurities to the plasma

• Provide limiting surfaces that define the plasma boundary during startup and shutdown

• Provide passage for the plasma diagnostics


Slide 4

Mod

ules

1-6

Modules 7-10

Mod

ules

11-

18

~850

– 1

240

mm

SB (semi-permanent) FW Panel (separable)

Blanket Module

Blanket System

• The Blanket System consists of Blanket Modules (BM) comprising two major components: a plasma facing First Wall (FW) panel and a Shield Block (SB).

50% 50%50% 40%10%

SB Attachment to VV

100%

http://www.iterkorea.org/

http://fusionforenergy.europa.eu/

http://www.iterrf.ru/

http://www.iterrf.ru/


Slide 5

Blanket System in NumbersNumber of Blanket Modules: 440Max allowable mass per module: 4.5 tonsTotal Mass: 1530 tons

First Wall Coverage: ~600 m2

Materials:- Armor:Beryllium- Heat Sink:CuCrZr- Steel Structure:

316L(N)-IG- Axial attachment (bolt/cartridge) Alloy 718- Pad

Al Bronze- Insulating layer

Alumina

Max total thermal load: 736 MW

Cooling water conditions: 4 MPa and 70°C


Slide 6

Impact of Interface Requirements on Blanket Design

• Interface requirements impose challenging demands on the blanket in particular since the blanket is in its pre-PA phase whereas several major interfacing components are already in procurement.

• Such demands include:- Accommodating plasma heat loads on FW- Maintaining acceptable load transfer to the vacuum vessel- Providing sufficient shielding to the vacuum vessel and TF coils- Accommodating the space allocations for in-vessel coils,

manifolds and plasma heating systems.

• These are highlighted in subsequent slides as part of the blanket design description.


Slide 7

• The blanket is a major contributor to neutron shielding of the coils and vacuum vessel.

• e.g. to enhance shielding, blanket-related modifications were introduced following PDR: - a flat inboard profile - an addition of 4 cm to mid-plane radial thickness - a reduction of the vertical gaps between inboard SB’s from 14 to

10 mm. • This has resulted in lowering the integrated heating in the

coil to from 17 to 15.7 kW (as compared to a target of 14 kW)

Inboard Module Shape and Size Optimized for Neutron Shielding of VV and TF Coil

• An assessment indicates that this could still allow fusion power close to the design value of 500MW for a 1800s rep period without any adjustment of the cryo-plant operating conditions.


Slide 8

I-shaped beam to accommodate poloidal torque

Design of First Wall Panel Impacted by Accommodation of Plasma Interface Requirements


Slide 9

Plasma

RH accessExaggerated shaping

- Allow good access for RH

- Shadow leading edges

Shaping of First Wall Panel• Heat load associated with charged particles is a major component of

heat flux to first wall.• The heat flux is oriented along the field lines.• Thus, the incident heat flux is strongly design-dependent (incidence angle of the field line on the component surface). • Shaping of FW to shadow leading edges and penetrations.


Slide 10

Plasma Loads to the First wall- Combination of steady-state and off-normal loads- Design based on steady-state loads and maximum armor thickness- Beryllium surface damage accepted during off-normal events- Learn from early operation before moving to higher power scenarios to minimize off-

normal events with major impact on FW armor lifetime


Slide 11

First Wall Panels: Design Heat Flux

• 218 Normal heat flux panels EU

• 222 Enhanced heat flux panels RF, CN


Slide 12

Normal Heat Flux (NHF) First Wall

CuCrZr AlloySS PipesBe tiles

SS Back Plate

Normal Heat Flux Semi-PrototypeNormal Heat Flux Finger:• q’’ = 2 MW/m2 • Steel Cooling Pipes• HIP’ing


Slide 13

Enhanced Heat Flux (EHF) First WallEnhanced Heat Flux Finger:• q’’ = 4.7 MW/m2 • Hypervapotron• Explosion bonding (SS/CuCrZr) + brazing (Be/CuCrZr)

Be tilesBimetallic structure

CuCrZr/SS 316L(N)-IG

Finger body (Steel 316L(N)-

IG)

Coolant channel lid (Steel 316L(N)-IG)

Hypervapotron (HVT)

Central weldingExternal welding

Brazing

Laser welding

Laser welding

EHF FW Semi-Prototype


Slide 14

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

• Cooling holes are optimized for Water/SS ratio (Improving nuclear shielding performance).

• Cut-outs at the back to accommodate many interfaces (Manifold, Attachment, In-Vessel Coils).


Slide 15

Shield Block Full-Scale Prototypes

KODA: SB08

CNDA: SB14 (example of some processes)

• Basic fabrication method from a single high-quality forged steel block and includes drilling of holes, welding of cover plates for water headers, and final machining of the interfaces.


Slide 16

Shield Block Attachment• 4 flexible axial supports• Keys to take moments and forces• Electrical straps to conduct current to vacuum vessel

e.g. BM 4 Interfaces

MANIFOLD

FLEXIBLE CARTRIDGE

INTERMODULAR PADS

SHIELD BLOCK

FIRST WALL

ELECTRICAL STRAPS (SB & FW)

CENTERING PADS


Slide 17

Flexible Axial Support

• 4 flexible axial supports located at the rear of SB, where nuclear irradiation is lower.

• Compensate radial positioning of SB on VV wall by means of custom machining.

• Cartridge and bolt made of high strength Alloy-718• Designed for 800 kN preload to take up to 600 kN Category III load.

FSP for testing (NIKIET, RF)


Slide 18

Keys in Inboard and Outboard Modules to Accommodate Moments and Forces from EM Loads

• Each inboard SB has two inter-modular keys and a centering key to react the toroidal forces.

• Each outboard SB has 4 stub keys concentric with the flexible supports.

• Bronze pads are attached to the SB and allow sliding of the module interfaces during relative thermal expansion.

• Key pads are custom-machined to recover manufacturing tolerances of the VV and SB.


Slide 19

Pads and Insulating Layer

• Pads moved to Shield Block to minimize impact on VV interfaces

• Electrical insulation of the pads through ceramic coating on their internal surfaces.

• Cylindrical pads for IMK; rectangular pads for stub keys• Insulating layer performance determined by R&D (8000 cycles at 1.5 MN and 1 run at 2

MN))

Scheme of Testing

Test Samples: Al-bronze, plasma sprayed Al2O3

Inter-Modular Key pads on Inboard


Slide 20

Shield Block and Attachment Designed to Respect Pre-Defined Load from Vacuum Vessel Load Specifications

• Optimizing blanket design (radial thickness and slitting) to reduce EM loads based on the following analysis:- DINA analysis of disruptions and VDEs- Eddy and halo analysis to obtain superposition of wave forms- Dynamic analysis of BM structural response using ANSYS (NIKIET)

• For example, results for BM 1 (one of the most loaded modules) show that the loads on the keys are within the VV LS for Cat. II and Cat. III cases


Slide 21

Interface with Blanket Manifold and In-Vessel Coils• A multi-pipe manifold configuration has been chosen, with

each pipe feeding one or two BM’s - Higher reliability due to minimization of welds and utilization of seamless pipes.

- Superior leak localization capability due to larger segregation of cooling circuits.

- Elimination of drain lines.

• In-Vessel coils: three sets of ELM control coils per outboard sector to control ELMs by applying an asymmetric resonant magnetic perturbation to the plasma surface and 2 sets of vertical stability (VS) coils for plasma stabilization.

• BMs to be designed with cut-outs to accommodate these space reservations.

Multi-pipe Manifold Configuration

Example outboard SB 12 with manifold and ELM coil

cut-outs

ELM coils

VS coils


Slide 22

Interface with Neutral Beam System

• The NB region with the neutron bean injection ports create a particular geometry challenge for the blanket design

• Special modules have been developed to accommodate this as well as the other design constraints

• The NB shine-through also needs to be accommodated by the impacted Blanket Modules


Slide 23

FW 16 to Accommodate Neutral Beam Shine-Through

• Detailed Model done following EHF FW guidelines (central-slot insert not shown here for simplicity).

• Special shaping to accommodate shine-though beam and protect slot.

• Analysis assessment performed to verify stress compliance with SDC-IC (ITER structural design criteria) for both M-type (monotonic) and C-type (cyclic) damage.


Slide 24

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 part of the R&D effort is the 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).- Qualification tests include: He leak test, pressure test, FW heat flux test


Slide 25

Procurement Schedule

Component SMP PA Start/End Date*

Blanket First WallRFDACNDAEUDA

Nov-2013 / Dec-2020Sept-2014 / July-2020Dec-2014 / Feb-2021

Blanket Shield BlockCNDAKODA

Nov-2013 / Feb-2021Nov-2013 / Oct-2020

Blanket Module ConnectorsRFDA July-2014 / Jan-2020

- FDR: April 2013 (successfully held)- FDR Approval: July 2013 (ahead of schedule)

*End date = delivery of last component to ITER site


Slide 26

Summary• The Blanket design is extremely challenging, having to

accommodate high heat fluxes from the plasma, large EM loads during off-normal events and demanding interfaces with many key components (in particular the VV and IVC) and the plasma.

• Substantial re-design following the ITER Design Review of 2007. The Blanket CDR, PDR and FDR have confirmed the correctness of this re-design.

• Effort now focused on finalizing the design work based on input received at the FDR and to prepare for procurement starting at the end of 2013.

• Parallel R&D program and DA pre-qualification process by the manufacturing and testing of full-scale or semi-prototypes.


Slide 27

Latest Blanket Major Milestone

Blanket FDR successfully held in April 2013

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