ITER full W Divertor Design and
Progress of Technology R&D

Presented by T. Hirai
Tungsten Divertor Section, ITER Organization
On behalf of Tungsten Divertor Task Force at IO
Japan Fusion Energy Forum Sub-cluster MTG DIV-SOL +++, 29-30 Aug 2013,
Tsukuba Japan

Slide 1

Contributors
J. Gunn (CEA)
A. Labusov (Energopul)
P. Stangeby (Toronto Univ)
S. Villari (ENEA)
JADA OVT PA team (JAEA)EUDA IVT PA team (F4E)
RFDA HHF test PA team (NIIEFA)

Tungsten Divertor Task Force:
V. Barabash
S. Carpentier
A. Durocher
F. Escourbiac
A. Fedosov
L. Ferrand
T. Hirai*
T. Jokinen
V. Komarov
M. KocanA. Kukushkin
M. Lehnen
M. Merola
R. Mitteau
R. A. Pitts
W. ShuM. Sugihara* TF leader

Japan Fusion Energy Forum Sub-cluster MTG DIV-SOL +++, 29-30 Aug 2013,
Tsukuba Japan

Slide 2Divertor Cassette assembly
1. 54 Divertor Cassettes + 6 spare Cassettes
2. 3 Plasma-Facing Components (PFCs) on a cassette
3. Outer/Inner Vertical Targets, Dome (Umbrella and Reflector Plates)
4. Plasma-Facing Units (monoblock for VT PFUs and flat tile concept for Dome PFUs)
5. Plasma-facing Unit Attachments (plug and support leg)
6. Plasma-facing Component Attachments (Pin and multilink)
7. Divertor Cassette Attachments (Nose and Knuckle)
8. Diagnostics integration
54 Cassettes Assemblies
9. Remote Handling interface
8.6 t per CA
10. Circulation of cooling water
Dome (W)

Inner verticaltarget (CFC/W)

2.2 m
Outer verticaltarget (CFC/W)

Cassette body (SS)

3.3 m
Japan Fusion Energy Forum Sub-cluster MTG DIV-SOL +++, 29-30 Aug 2013,
Tsukuba Japan

Slide 3

Background of Full-W Divertor Activity
 New ITER divertor strategy− From MAC-12 (Oct. 2011) [ITER_D_6SR5F7]: “[…] for budgeting
purposes, the MAC supports the DG recommendation to de-scope
to a single divertor during construction and initial five years ofoperation”.[…] “As to the specific choice of divertor, MAC
recommends delaying the decision for up to two years.
− These recommendations were adopted by the ITER Council during
its ninth meeting in Nov. 2011 so that IO started to investigate thepossibility of beginning operation with a full-W armoured divertor.

Japan Fusion Energy Forum Sub-cluster MTG DIV-SOL +++, 29-30 Aug 2013,
Tsukuba Japan

Slide 4

Divertor Maintenance strategy

Full-W Divertor

 Extension of Lifetime

Japan Fusion Energy Forum Sub-cluster MTG DIV-SOL +++, 29-30 Aug 2013,
Tsukuba Japan

Slide 5

Outline
1. Design constrain/input
2. Detailed divertor PFC Design
3. Design Validation
4. Technology R&D requirements
5. Progress of technology R&D in EU and JA
6. Conclusion/ScheduleJapan Fusion Energy Forum Sub-cluster MTG DIV-SOL +++, 29-30 Aug 2013,
Tsukuba Japan

Slide 6

W Divertor Design – Requirements/ Input• Structural Design
 Load Specification for the ITER Divertor (C9RF33)
 Structural Design Criteria – In-vessel Component (SDC-IC) (222QGL)• Plasma-Facing Unit Design Heat Load Specification for ITER Tungsten Divertor:(1) Power density and number of cycles on the Vertical Target (7GFMB6)(2) Poloidal conductive/convective surface heat loads (HDB7NZ)• Tolerances Assembly tolerances between Cassettes (20mm gap) : step ±2 mm (Divertor PlasmaFacing Surfaces Position and Shadowing Tolerances Build Up (764MGY), and margin) Assembly tolerances between two halves of VT PFCs (3mm gap) : step ±0.3 mm(in 2D drawings)

 Assembly tolerances between VT PFUs (0.5mm gap): step ±0.3 mm (in 2D drawings)

Japan Fusion Energy Forum Sub-cluster MTG DIV-SOL +++, 29-30 Aug 2013,
Tsukuba Japan

Slide 7

W Divertor Design: Design Constrains/ Assumptions
1. Single coolant tube for a PFU to limit impact on CB
design, diagnostics procurement, assembly plan,other interfaces (splitting target and baffle is notconsidered)
2. Dome design is retained as long as neutronics
analysis confirms
3. Materials for Tungsten Divertor remain same
except (a) exchange of CFC by W and (b) updated
radioprotection requirements
4. Monoblock concept is retained
5. PFU attachment concept (PFU to the steel support
structure) is retained.
6. DW VDE (limiter configuration) contact only at
OVT baffle according to the recent DINA analysis(ITER_D_765S86).

Japan Fusion Energy Forum Sub-cluster MTG DIV-SOL +++, 29-30 Aug 2013,
Tsukuba Japan

Slide 8

Thermal Event Impact at Plasma-Facing Surfaces
IVT

Steady state,
Slow transient load,
ELM, MD, m-ELM,
m-MD, m-VDE

Runaway
ElectronDome“m-” stands for “mitigated”

VDE (limiter like impact)

OVT

Limiter like
plasma impact
Uni-directional
plasma impact
Electron impact

Cassette Body
Japan Fusion Energy Forum Sub-cluster MTG DIV-SOL +++, 29-30 Aug 2013,
Tsukuba Japan

Slide 9Outline
1. Design constrain/input
2. Detailed divertor PFC Design
3. Design Validation
4. Technology R&D requirements
5. Progress of technology R&D in EU and JA
6. Conclusion/ScheduleJapan Fusion Energy Forum Sub-cluster MTG DIV-SOL +++, 29-30 Aug 2013,
Tsukuba Japan

Slide 10

Divertor Shaping Strategy
Objective:
To avoid melting due to leading edge exposure during steady-state andslow transient and to minimize the melting under off-normal events –full edge-shadowing in high heat flux handling areas

1. Tilting of each Divertor Plasma-Facing Component
2. Toroidal-Roof-shaping and Outer Vertical Target chamfer
3. Monoblock toroidal chamfer

Japan Fusion Energy Forum Sub-cluster MTG DIV-SOL +++, 29-30 Aug 2013,
Tsukuba Japan

Slide 11

Tilting Scheme for OVT, IVT and Dome
IVT

OVT

Set-backtilting axis+80º

0.5o
Tilting axis

BaffleTransition

0.74ºTargetTilting - By tilting of OVT
and IVT, a nominal step
between VT is implemented
at the strike point regions.
Components are tilted
when assembled onto theCassette Body (CB), to
ensure no W leading edges
from PFC to PFC

Dome

IPRP

IVT

Umbrella

OPRP

Dome

OVT

Cassette Body

Japan Fusion Energy Forum Sub-cluster MTG DIV-SOL +++, 29-30 Aug 2013,
Tsukuba Japan

Slide 12

IVT design – Monoblock toroidal chamfer at target
R16

R09 R08

R01

PFU variant: R01=R02…R16 (Reference PFUs)
Baffle (A – A)

BaffleSteady State Heat Flux

A

Nominal
step

A
~30

Target (B – B)

TargetSteady State Heat Flux

B

B

Designed
step
Monoblocktoroidalchamfer~30

Higher shoulder: 8 mm
Lower shoulder: 7.5 mm
Toroidal monoblock chamfer: 0.5 mm

Monoblock toroidal chamfer to protect edge from 0.3
mm deviation  0.5 mm toroidal monoblock chamfer
Japan Fusion Energy Forum Sub-cluster MTG DIV-SOL +++, 29-30 Aug 2013,
Tsukuba Japan

Slide 13

OVT design: PFC chamfer + monoblock chamfer
PFU variants: R01=R12; R11=R20;
R02= R03…R10, R13…R19 (reference PFUs)

R21 R19 … R13
R10 … R02
R22 R20
R12 R11
R01

BaffleBaffle (A – A)

Steady State Heat Flux

Nominalstep

Chamfer/PFU
higher

Set-back
LowershoulderA

TargetTransition

A

B

B

Chamfer/PFU~30
Chamfer/PFU: 7 mm
Set-back of 3 PFUs: 21 mm
Higher shoulder: 15 mm
Lower shoulder: 8mm

Target (B – B)
Nominal
step~30

Steady State Heat Flux

Monoblocktoroidalchamfer
Higher shoulder: 8 mm
~30
Lower shoulder: 7.5 mm
Toroidal monoblock chamfer: 0.5 mm

Japan Fusion Energy Forum Sub-cluster MTG DIV-SOL +++, 29-30 Aug 2013,
Tsukuba Japan

Slide 14

OVT design – Transition from target to baffle• Transition: (1) monoblock thickness; (2) monoblock
toroidal chamfer; (3) PFU profile Monoblocks have simple chamfers at plasma facing side
 Location of transition: away from the VDE impact area – steady state loadingarea Poloidal edge: exposed toward bottom direction

BaffleTransition

TargetJapan Fusion Energy Forum Sub-cluster MTG DIV-SOL +++, 29-30 Aug 2013,
Tsukuba Japan

Slide 15

Manufacturing of W monoblock and PFU• W Monoblock Monoblocks consist of simple chamfer or double chamfer at plasma facingside (No complicated 3D profile) Monoblocks can be manufactured by conventional operations such as:• EDM wire cut operation (no need of milling)z• Grinding operation• Conventional cleaning processy Inspection i.e. dimension control, to be defined in NDT protocol

x

• PFU

y

 OVT: only 5 variants
Reference OVT PFU R2 = R3-R10 = R13-R19;
Toroidal roof shaping OVT PFUs R1=R12; R11=R20; R21; R22

 IVT: only 1 variant
Japan Fusion Energy Forum Sub-cluster MTG DIV-SOL +++, 29-30 Aug 2013,
Tsukuba Japan

Slide 16

Dome cooling structure and PFU attachment• Dome design (cooling structure & PFU
attachment) remain the same

W Flat tiles

W TileCopperInterlayer
CuCrZrHeat Sink
316L(N)

Japan Fusion Energy Forum Sub-cluster MTG DIV-SOL +++, 29-30 Aug 2013,
Tsukuba Japan

Slide 17

Outline
1. Design constrain/input
2. Detailed divertor PFC Design
3. Design Validation
4. Technology R&D requirements
5. Progress of technology R&D in EU and JA
6. Conclusion/ScheduleJapan Fusion Energy Forum Sub-cluster MTG DIV-SOL +++, 29-30 Aug 2013,
Tsukuba Japan

Slide 18

Design Validation by analysis
1. Neutronic analysis
2. Design Supporting analysis (thermo-mechanicaland EM analysis)
3. Heat Load distribution Study – PFC andmonoblock level

Japan Fusion Energy Forum Sub-cluster MTG DIV-SOL +++, 29-30 Aug 2013,
Tsukuba Japan

Slide 19

Nuclear Analysis for ITER W Divertor
Model: Neutronic analysis by means of the latest ITER MCNP-5 Monte Carlo
Code for 3D model (MCNP ITER 40° model B-lite v-2)

Assumption: 18% of the cumulated 3x1027 neutrons (End of Life) during first
divertor (up to end of the first full DT campaign)

Aim: Evaluate nuclear heating, radiation damage and helium production

Japan Fusion Energy Forum Sub-cluster MTG DIV-SOL +++, 29-30 Aug 2013,
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Slide 20

Nuclear Analysis Results – damage (dpa)Villari - Neutronic Analysis (ITER_D_HQZHC5)
Maximum dpa at ITER EOL (permanent)
Component

W

Cu/CuCrZrSS

IVT PFU

0.51

2.37

1.28

OVT PFU

0.54

2.50

1.52

Dome Umbrella PFU

0.48

2.11

1.31
0.14

CB plate below Dome
Maximum dpa after 4 years of ITER nuclear phase (first cassette)
Component

W

Cu/CuCrZr SS

IVT PFU

0.089

0.41

0.22

OVT PFU

0.095

0.44

0.27

Dome Umbrella PFU

0.084

0.37

0.23

CB plate below Dome

0.025

Japan Fusion Energy Forum Sub-cluster MTG DIV-SOL +++, 29-30 Aug 2013,
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Slide 21

Design supporting analysis
Input:
Defined Load combinations in Load Specification for the ITER Divertor (C9RF33)

Aim:
Validate structural integrity under the following loads and load combinations by FEM






Inertial loads (associated with gravity and seismic events);
Hydraulic pressure loads,
Electromagnetic loads,
Thermal loads (due to nuclear heating and surface heat fluxes), andAssembly loads (typically due to preloads imposed on the CB during assembly andpreload of the bolts).

Japan Fusion Energy Forum Sub-cluster MTG DIV-SOL +++, 29-30 Aug 2013,
Tsukuba Japan

Slide 22

Design supporting analysis – Load Combinations
Individual events in ITER tokamak

Additional loads specific to divertor

Category

Number of events(1)DW, Hydr. test pressure,Test

1
200 at 240 C 4.4 MPa
200 at 350 C at 1 MPa
Only for thermal loads:
5000 for PFCs and CB @ 10 MW/m2
300 for PFCs and CB @ 20 MW/m2

1

Plasma(3)Testing regime

2

Baking regime(2)Preload, DW, CP, TH

I

3

Normal operation

Preload, DW, CP, TH

I

Seism

Magnet30,000 for rails
4567

SDVDE-II 50-100 ms
FDVDE-II 36 ms
MD-I 50 ms
SL-1(3)8910 SL-1(3)11 SL-1(3)SDVDE-III (>200ms)
FDVDE-III 36 ms
SDVDE-II 50-100 ms
FDVDE-II 36 ms

12 SL-2(3)13
SDVDE-IV slow-fast
14“Rotating Asymmetric
VDEs”
15 SL-2(3 Baking regime
16 SL-2(3 Maintenance

MFD-II
MFD-II

Preload, DW, CP, TH
Preload, DW, CP, TH
Preload, DW, CP, TH
Preload, DW, CP, TH

II
II
II
II

150
150
3000
1

MFD-II
MFD-II

Preload, DW, CP, TH
Preload, DW, CP, TH
Preload, DW, CP, TH
Preload, DW, CP, TH

III
III
III
III

1
1
1
1

Preload, DW, CP, TH
Preload, DW, CP, TH
Preload, DW, CP, TH

IV
IV
IV

1
1
1

Preload, DW, CP, TH
DW

IV
IV

1
1

Load Specification for the ITER Divertor System (C9RF33)
Japan Fusion Energy Forum Sub-cluster MTG DIV-SOL +++, 29-30 Aug 2013,
Tsukuba Japan

Slide 23

Design Supporting Analysis Results
Stress intensity (MPa) at 2nd load combination: CB preload + DW+ CP + Temperature
ITER_D_HR2ZCU

Primary stresses are within allowable limits for the analysed load cases (DW,
Pressure test, Coolant pressure, EML)
Cyclic strength criteria (no ratcheting) are satisfied for the SSS of the PFCs, DomePFUs and CB.
Japan Fusion Energy Forum Sub-cluster MTG DIV-SOL +++, 29-30 Aug 2013,
Tsukuba Japan

Slide 24

Heat Load distribution study – PFC level
SteadyState

Input:DW VDE - Plasma configurations- Heat load specification- Max misalignment at gap- Plasma-facing surface envelop
Tool:
PFCFLUX – 3D field line tracing
code
Aim:
Validate full edge shadowing
during Steady State and VDEs
Key component: OVTJapan Fusion Energy Forum Sub-cluster MTG DIV-SOL +++, 29-30 Aug 2013,
Tsukuba Japan

Slide 25

Heat Load Distribution Analysis Results
 Examin 5 (IP, BT) steady-states No leading edges

 Examine Analysis of ~75 DINA time
sequences  no leading edges‘DW VDE, init li = 0.6, 675 ms’

15MA, 5.3T

einc (MJ/m2)

Wetted area at OVT surface

Energy density at OVT surface

With worst misalignment

With worst misalignment- Demonstrates OVT shaping design could avoid leading edges
Japan Fusion Energy Forum Sub-cluster MTG DIV-SOL +++, 29-30 Aug 2013,
Tsukuba Japan

Slide 26

Heat Load distribution study – monoblock level
Monoblock (axial 12 mm; toroidal 28 mm)edges could be exposed to flux tubes
penetrating through the monoblock gaps
(nominal width 0.5 mm).

Input:- Plasma configurations- Heat load specification- Max misalignment at gap- Monoblock geometryTool:
1. PFCFLUX:
Field tracing analysis (optical)

2. Monte Carlo test particle
simulation:
Ions with random perpendicular velocities(Maxwellian distribution) in magnetic field

Aim:
Validate full edge shadowing at HHF
area at monoblock level
Under final tuning: double chamfer
Japan Fusion Energy Forum Sub-cluster MTG DIV-SOL +++, 29-30 Aug 2013,
Tsukuba Japan

Slide 27

Outline
1. Design constrain/input
2. Detailed divertor PFC Design
3. Design Validation
4. Technology R&D requirements
5. Progress of technology R&D in EU and JA
6. Conclusion/ScheduleJapan Fusion Energy Forum Sub-cluster MTG DIV-SOL +++, 29-30 Aug 2013,
Tsukuba Japan

Slide 28

Full-W Divertor Qualification Program
2 steps in the program(1)Technology Development and Validation : Demonstration of thefitness-for-purpose of the proposed technology, full-W small-scale mockups manufacturing and High Heat Flux testing(2) Full-scale demonstration : Demonstration of the technology via fullscale-prototype PFU manufacturing and testing in IDTF; Compliance with
IO procurement quality requirement~2 m
Full-scale
OVT PFUor IVT PFU

Japan Fusion Energy Forum Sub-cluster MTG DIV-SOL +++, 29-30 Aug 2013,
Tsukuba Japan

Slide 29

1st step: Technology Development and Validation• PFU attachment joint(1) Uniaxial tensile test at RT (at least 5 samples; Load speed 20-60 N/s)
Acceptance Criteria: >8 kN for OVT and >10 kN for IVT (attachment/5monoblocks)(2) Cyclic zero-pull fatigue test at RT cycle number 15k cycles (<1 Hz) 0-8 kN (010kN) OVT (IVT) PFU leg (attachment/5monoblocks)
Loading axis
Acceptance Criteria: deformation along the loading axis < 0.3 mm• W monoblock armor to heat sink joint
HHF test 5000 cycles at 10 MW/m2 and 300 cycles at 20 MW/m2
Acceptance Criteria: No trace of substantial melting by Visual Inspection / No
water leak / No detachment of any armour block / No appearance of any “hotspot” during fatigue cycling / no variation of max Tsurf (oC) exceeding 20%
increase between initial and final thermal mapping

Japan Fusion Energy Forum Sub-cluster MTG DIV-SOL +++, 29-30 Aug 2013,
Tsukuba Japan

Slide 30

2nd step: Full-scale demonstration• Tube to tube joint qualification(1) Welding qualification test in accordance to EN ISO 15614-1
Acceptance Criteria: Quality class B as defined in EN ISO 5817 or EN ISO 13919-1,
and ITER Vacuum Handbook Attachment-1(2) He leak test after rotary bending fatigue test 0.1% strain,10k cycles (<1Hz) at RT
Acceptance Criteria: He leak rate ≤ 10-10 Pa m3/s
(3) Tensile test at 150oC
Acceptance Criteria: > 200 MPa• PFU W monoblock armor to heat sink joint HHF test 5000 cycles at 10 MW/m2 and 300 cycles at 20
MW/m2 at straight part
 HHF test 5000 cycles at 5 MW/m2 at curved part
Acceptance Criteria: No trace of substantial melting by Visual
Inspection / No water leak / No detachment of any armour block /No appearance of any “hot spot” during fatigue cycling / no variationof max Tsurf (oC) exceeding 20% increase between initial and final
thermal mapping~2 m
Full-scale
OVT PFUor IVT
PFU• Compliance with ITER Procurement Quality Requirements
Japan Fusion Energy Forum Sub-cluster MTG DIV-SOL +++, 29-30 Aug 2013,
Tsukuba Japan

Slide 31ITER Divertor Test Facility (IDTF)
IDTF that commissioned in frame of Divertor PA Built within Procurement Arrangement 17P2D
 Location: Efremov Institute, St-Petersburg, RF Electron beam test facility Maximum electron beam power: 800 kW
 Maximum accelerating voltage: 60kV
 Cooling water parameters are ITER divertor
relevant Dedicated system of diagnostics

Japan Fusion Energy Forum Sub-cluster MTG DIV-SOL +++, 29-30 Aug 2013,
Tsukuba Japan

Slide 32

Task Agreements to support technology R&D• Objectives: Deliver results of technology validation timely for the decision of thearmour material (end 2013)
 Support demonstration by the full-scale prototypes activities• 3 TAs with DAs
 With JA DA: for the manufacturing of OVT small-scale mock-ups andfull-scale prototype PFUs
 With EU DA: for the manufacturing of IVT small-scale mock-ups (full-W
full-scale PFU manufacturing already planned in the PA for IVT)
 With RF DA: for the HHF testing of the above component s at the ITER
Divertor Test Facility (IDTF)

Japan Fusion Energy Forum Sub-cluster MTG DIV-SOL +++, 29-30 Aug 2013,
Tsukuba Japan

Slide 33

Outline
1. Design constrain/input
2. Detailed divertor PFC Design
3. Design Validation
4. Technology R&D requirements
5. Progress of technology R&D in EU and JA
6. Conclusion/ScheduleJapan Fusion Energy Forum Sub-cluster MTG DIV-SOL +++, 29-30 Aug 2013,
Tsukuba Japan

Slide 34

EU HHF Test Results 2009-2011
HHF tests were performed on the FE200 facility at AREVA.
Tungsten monoblock mock-ups armouring CuCrZr coolingpipes were successfully tested up to 15 MW/m2 for 1000
cycles and 20 MW/m2 for 300 cycles [1].

Table summarizes the main results on the tested tungsten mock-ups [2]1000 cycles at 10 MW/m2
No significant visible effect1000 cycles at 15
MW/m2
Surface modification
but no melting trace

300 cycles at 20 MW/m2
Self‐castellation ; No debonding oftile ; no substantial melting

Results fulfil the IO acceptance criteria, and indicate that W
monoblock technology is acceptable for the requirements of the
full-W vertical target[1] P. Gavila et al, Fus. Eng. & Des. 86 (2011) 1652-1655[2] P. Lorenzetto et al, 24th IAEA-FEC 2012
Japan Fusion Energy Forum Sub-cluster MTG DIV-SOL +++, 29-30 Aug 2013,
Tsukuba Japan

Slide 35

EU HHF test results in 2013
EU DA campaign: on 12 mock-ups in IDTF in RF
HHF Testing program:• step 1 : 5000 cycles at 10 MW/m²• step 2 : 300 (+700 cycles) at 20 MW/m²• step 3 : critical heat flux
HHF results @ 20MW/m2 obtained on first set of
6 mock-ups.
• No traces of melting on the loaded surface• Cracking developmentwith 7.5 mm of W thickness

Self-castellation was observed

W PLANSEEW PLANSEEwith 6 mm of W thickness

W PLANSEE

Plansee 20MW/m2 after 75 cycles

W PLANSEEW POLOMAW PLANSEE

Ansaldo 20MW/m2 after 75 cycles

Second set of HHF testing
program with 6mm of W thickness
was recently stopped.

Leak occurred after performance of 5000 cycles
at 10 MW/m2 + 529 cycles at 20 MW/m2.

Japan Fusion Energy Forum Sub-cluster MTG DIV-SOL +++, 29-30 Aug 2013,
Tsukuba Japan

Slide 36

JA HHF results in 2012JA DA campaign: OVT CFC/W Full-Scale PFUs
Straight W part of PFUs subjected to the HHF tests -- 5000 cycles at 10 MW/m2 and1000 cycles at 20 MW/m2
20 monoblocks of 12mm (axial) x 28mm (poloidal) x 7.5 mm(W thickness at the top of the tube).

W- ALTM

W- ATM

W- TLWM

W- ATM

Results fulfil the IO acceptance criteria, and indicate that W monoblocktechnology is acceptable for the requirements of the full-W vertical target
Small scale mockups will be tested in IDFT 24-26 Sep 2013
Japan Fusion Energy Forum Sub-cluster MTG DIV-SOL +++, 29-30 Aug 2013,
Tsukuba Japan

Slide 37

Effects of Irradiation on Performance
The effects of irradiation on the performance of the W monoblock have beenassessed by experiments – PARIDE 3 and PARIDE 4 [4].

W monoblocks

Neutron-Irradiation campaign:
PARIDE 3: 0.1 dpa at 200 °C // PARIDE 4: 0.6 dpa at 200 °C

Irradiated W mock-ups (0.6 dpa at 200°C) sustained thermal
fatigue testing at 18 MW/m2 for 1000 cycles
Post irradiation testing of samples from the irradiation experiments PARIDE 3 and PARIDE 4, M. Roedig et al., JNM 329–333 (2004) 766–770

Japan Fusion Energy Forum Sub-cluster MTG DIV-SOL +++, 29-30 Aug 2013,
Tsukuba Japan

Slide 38

Outline
1. Design constrain/input
2. Detailed divertor PFC Design
3. Design Validation
4. Technology R&D requirements
5. Progress of technology R&D in EU and JA
6. Conclusion/ScheduleJapan Fusion Energy Forum Sub-cluster MTG DIV-SOL +++, 29-30 Aug 2013,
Tsukuba Japan

Slide 39

Schedule for full-W Divertor DecisionFinal Design Review of Tungsten Divertor (26-28 June 2013) Full-W Divertor design
 Full-W Divertor technology qualification programme

Physics Assessment
 Report from ITPA Topical Groups to STAC-14 (May 2013) giving aphysics/operational opinion on a full-W divertor start for ITER JET misaligned lamella melt experiment: summer 2013



STAC-15 (Oct 2013) to take note of the Design and Physics outcome and provide
recommendation



ITER Council-13 (Nov 2013) to endorse the STAC/ MAC recommendation on
this topic

Implementation of the decision into the baseline via a PCR (end-2013,
SMP.0116.001032 milestone) if selected
Japan Fusion Energy Forum Sub-cluster MTG DIV-SOL +++, 29-30 Aug 2013,
Tsukuba Japan

Slide 40

Thank you very much for your attention

Japan Fusion Energy Forum Sub-cluster MTG DIV-SOL +++, 29-30 Aug 2013,
Tsukuba Japan

Slide 41