Session 4: Plasma facing components - Nucleus Documents... · 15/11/2017 1 Session 4: Plasma facing...

1 15/11/2017

Session 4: Plasma facing components

2nd IAEA Technical Meeting on Divertor Concepts


Eliseo Visca, ENEA

2

Introduction

Outstanding Technical Challenges with Gaps beyond ITER

15/11/2017

Blanket and fuel cycle

FW and PFCs ‘Fusion’ auxiliares

Structural materials

From ITER to DEMO to Fusion Power Plant (FPP)

the DEMO goals are to demonstrate…

• a workable solution for all physics and technology questions

• large scale net electricity production

• T self-sufficient fuel cycle

• high reliability and availability over a reasonable time span


3

Introduction

15/11/2017

Plasma Facing Components

First wall (less than 1MW/m2 )

structure with coolant channels is covered with a plasma-facing material that provides

protection against the particle load from the plasma as well as protection against

damage from transient heat loads. The stationary heat loads from the plasma to the first

wall are moderate and may reach up to 1 MW/m2 in a reactor. Erosion of the plasma-

facing wall material is regarded as the dominating lifetime issue in a reactor

Divertor (up to 20 MW/m2)

complex component with most intense plasma contact. The plasma-facing materials

of the divertor have to be bonded to a high thermal conductivity heat sink material in

order to remove very high stationary heat fluxes of up to 20 MW/m2 and high erosion

resistance for armour



4

Introduction

15/11/2017

ITER DEMO Experimental device with physics and technology

development missions.

Nearer to a commercial power plant but with some

development missions.

• 400 s pulses, long dwell time. Long pulses (> 2h) or steady state. Experimental campaigns. Outages for

maintenance, component replacements. Maximize availability.

Demonstrate effective and efficient maintenance and component replacement technologies.

Cooling system optimized for minimum stresses

and sized for modest heat rejection.

Cooling system optimized for electricity generation

efficiency (e.g. much higher temperature.)

No tritium breeding requirement Tritium breeding needed to achieve self- sufficiency.

Conventional 316 stainless steel structure for in-

vessel components.

Reduced activation materials as structure

Very modest lifetime n-fluence, low dpa and He

production.

High n-fluence, significant in-vessel materials damage.

Licensed as nuclear facility, but like a laboratory,

not a reactor.

Licensing as nuclear reactor. Potential for large tritium

inventory on-site.

Unique one-off design optimized for exptl. goals. Move towards design choices suitable for series

production.



5

Introduction

Near Term DEMO

15/11/2017

Power Exhaust

• Peak heat fluxes near technological limits (>10 MW/m2)

• ITER solution needs validation for DEMO

• Advanced divertor solutions may be needed but integration is very

challenging

• Plans to upgrade Medium Size Toks and/or build a dedicated DTT

Structural and HHF Materials

• n-Embrittlement of W and Cu-alloys at low temp. and loss of strength

at ~ high temp.

• Develop qualified and reliable joining technologies for PFC fabrication

• Need of structural design criteria and design codes

Remote Maintenance

• Strong impact on IVC design (cassette design, PFCs integration)


Conceptual Design Approach

Design features (near-term DEMO):

• 2000 MWth~500 MWe

• Pulses > 2 hrs

• water cooled divertor

• PFC armour: W

• LTSC magnets Nb3Sn (grading)

• Bmax conductor ~12 T (depends on A)

• RAFM (EUROFER) as blanket structure

• VV made of AISI 316

• Blanket vertical RH / divertor cassettes

• Lifetime: starter blanket: 20 dpa (200 appm He); 2nd blanket 50 dpa; divertor: 5 dpa (Cu)

Open Choices: • Operating scenario

• Breeding blanket design concept selection

• Primary Blanket Coolant/ BoP

• Protection strategy first wall (e.g., limiters)

• Divertor configurations (SN, DN, advanced)

• Number of coils

DEMO2

A.J.H. Donné, G. Federici and PPPT Team | IEA-FPCC | Paris | 27-28/01/2016| Page 10

DEMO1

Under revision

ITER DEMO1

(2015) A=3.1

DEMO2

(2015) A=2.6

R0 / a (m) 6.2 / 2.0 9.1 / 2.9 7.5 / 2.9

Κ95 / δ95 1.7 / 0.33 1.6 / 0.33 1.8 / 0.33

A (m2)/ Vol (m3) 683 / 831 1428 / 2502 1253 / 2217

H non-rad-corr / βN (%) 1.0 / 2.0 1.0 / 2.6 1.2 / 3.8

Psep (MW) 104 154 150

PF (MW) / PNET (MW) 500 / 0 2037 / 500 3255 / 953

Ip (MA) / fbs 15 / 0.24 20 / 0.35 22 / 0.61

B at R0 (T) 5.3 5.7 5.6

Bmax,conductor (T) 11.8 12.3 15.6

BB i/b / o/b (m) 0.45 / 0.45 1.1 / 2.1 1.0 / 1.9

Av NWL MW/m2 0.5 1.1 1.9

A.J.H. Donné, G. Federici and PPPT Team - 2016 15/11/2017 2nd IAEA Technical Meeting on Divertor Concepts

15/11/2017 7

DEMO core technical parameters

Rationale?


Introduction

8

Introduction

Rationale of material choice: what are usable elements for DEMO PFC?

15/11/2017

Only elements in green: anything else useful is transmuted by high energy neutrons into very radioactive products

Courtesy Forrest (UKAEA), Stork

Suitable for armour

Suitable for structural (EUROFER, ODS)

SiC-SiC


Melting point < 500°C Radioactivity/Tritium Availability/Cost Strength at 300 °C Ductility at 200 °C Water corrosion

Solid elements at RT with thermal conductivity > 50 W/mK (NB. irradiation)

Ta X

High-heat-flux target : requirements for heat sink material

J-H. You, Nucl. Fusion (2015)

15/11/2017 2nd IAEA Technical Meeting on Divertor Concepts 9

Introduction

15/11/2017 10

Copper means CuCrZr (Cu_OFE not suitable for n-irr) with a K=350 W/mK in the range of 250°C Degradation up to 2 dpa can be neglected for design analysis

ITER SDC-IC: Variation of the tensile strength versus temperature for CuCrZr-IG before and after irradiation (0.3 dpa £ dose £ 5 dpa and Ttest = Tirr).

ITER SDC-IC :Uniform Elongation for CuCrZr-IG before and after irradiation (0.3 dpa £ dose £ 5 dpa and Ttest = Tirr).

We rely on saturation effect of n-dose effect over 5 dpa, recover of ductility at temp range of 200-300 °C, taking into account of loss of tensile strength


Introduction

15/11/2017 11

The reference design concept geometry for divertor water cooled targets is monoblock R&D phase is concluded for ITER and qualification of procurement suppliers is on-going Technologies for fabrication of PFC have reached high level of maturity


Introduction – ITER experience

Full Scale full W Plasma Facing Units of ITER divertor Inner Vertical Target

15/11/2017 12

But considering for DEMO

Different hydraulic condition for water coolant n-irradiation material degradation Component lifetime

1st Issue for structural design of pipe Approxing 20MW/m2 stresses overcome pipe strength After irradiation-> ?


Introduction

J-H. You,M. Li IPP

15/11/2017 13

Assessment of fast fracture risk in pipe wall -> risk of crack growth

JQ of irradiated CuCrZr: 20-50 kJ/m² (~300 °C at 20 MW/m²)

Fracture toughness of CuCrZr alloy

2nd Issue for pipe design Ok when not-irradiated Risk of crack propagation when irradiated


Introduction

15/11/2017 14

Compliant layer W-pipe Pure Copper

It is a weak point in ITER-monoblock reference design

Creep-fatigue?

Under n-irradiation becomes brittle

It is really necessary?

INTERLAYER


Introduction

15/11/2017 15

(I) Helium Cooled Divertor (HCD) – Coolant temperature 700-800 °C

(1) hard design and integration of PFC (2) Low heat removal capability (3) structural material TBD

(II) Water Cooled Divertor (WCD) – Heat sink -> CuCrZr: T>300 °C softening, near to mechanical limits – W Armour

Recrystallization Erosion at high temperature Crack formation over 15 MW/m2 Crack propagation

- Interlayer - Reconsider its function - Different design to decrease thermal load

concentration

(III) Liquid Cooled Divertor

• laminates, particle and fiber reinforced Cu for possible operation at higher temperatures

• for advanced DEMOs

Think about of solution

• grain oriented pure W • WC & SiC reinforced W • W alloy development (PIM) • W-fiber-reinforced-W (WfW)

Topics/Strategies

• W/Cu functionally graded (FGM) • Thermal-break design concept

DEMO is designed by Code Important role plays Thermal Fatigue testing of components to check design concepts performances and to validate design code procedures


• for advanced DEMOs

15/11/2017 16

This session talks for 2nd IAEA technical meeting on Divertor concepts

Session 4: PFC Development Chair: Visca E.

8:30-9:15 It-4: Visca E. Overview of Advanced Water-Cooled Divertor Target Design

Concepts for European DEMO Reactor

9:15-9:45

Iv-14: Greuner H. Strategy and Results of High Heat Flux Testing of European DEMO

Divertor Mock-Ups

9:45-10:15 Iv-15: Neu R. Tungsten Fibre-Reinforced Composites for Advanced Plasma Facing Components

10:15-10:45

Iv-16: Mao Y. Powder Metallurgically Produced Tungsten Fiber-Reinforced

Tungsten Composites

10:45-11:15 Coffee-break

11:15-12:00 Discussion


Presenting: Visca Eliseo, ENEA Thanks to All Task Owners (CCFE, CEA, KIT, IPP, MPG, ENEA)

Overview of advanced water-cooled divertor target design concepts for European DEMO

reactor

15/11/2017 17 2nd IAEA Technical Meeting on Divertor Concepts

Eliseo Viscaa, J.H. Youc, F. Crescenzia, H. Greunerc, T. Barrettd, M. Richoub, A.v. Müllerc a Department of Fusion and Technology for Nuclear Safety and Security – ENEA, Italy b CEA, IRFM, F-13108 Saint-Paul-Lez-Durance, France c Max Planck Institute for Plasma Physics, Boltzmann Str. 2, 85748 Garching, Germany d CCFE, Culham Science Centre, Abingdon OX14 3DB, United Kingdom

18

PFC development approach of PPPT-WPDIV project

15/11/2017

Design concepts under development for target PFCs in the frame of the Eurofusion – PPPT-WPDIV project

Target concepts Coolant Armor Interlayer Heat sink Design logics

ITER-like (baseline) (ENEA)

water W block Thick Cu CuCrZr pipe Reduced size for stress relaxation

Thermal break layer (CCFE)

water W block Perforated thick Cu

CuCrZr pipe Mitigate local heat flux concentration at pipe wall

Composite block (IPP)

water W flat tile Thin Cu W-particle/Cu block

Enhance toughness Reduce thermal stress

Composite pipe (IPP)

water W block None W-wire/Cu pipe

Enhance high-temp. strength

Thin graded interlayer (CEA)

water W block Graded W/Cu film

CuCrZr pipe Enhance joining quality Avoid thick Cu interlayer

W/Cu laminate layer (KIT)

water W flat tile Thin braze Wl/Cu laminate pipe

Enhance high-temp. strength & toughness

Dual pipe jet injection (KIT)

helium W block Thin braze Steel pipe or Wl/Cu pipe

Exploit high-temp strength and irradiation resistance


19

PFC development approach of PPPT-WPDIV project

Helium-cooled Target Development (EUROFER ?) W Laminate Tube concept: W composite laminate tube with W

armour and helium jet cartridge (KIT) Dual pipe jet injection impingement concept + W-PIM (KIT)

15/11/2017

S. Antusch, J. Reisen, . Norajitra al


35

0°

C

550°C

1000 mm

Eurofer

350°C –

550°C

Eurofer

heat load

20

From PPPT-WPDIV Project Management Plan

Water-cooled Target Development steps

1. Define scope and specification (thermal loads, materials, hydraulic)

2. Identify PFC concepts for WCD targets

3. Design studies for each concepts by FEA (failure modelling )

4. R&D of joining technologies for fabrication

5. Mock-up fabrication and acceptance

6. Concept qualification by HHF test

7. Select reference concept(s) for target

15/11/2017 2nd IAEA Technical Meeting on Divertor Concepts

21

1. FEA for thermo-mechanical prediction and optimization

2. Fabrication

3. ENEA for Ultrasonic check

4. SATIR test in CEA

5. HHF in GLADIS facility (Garching DE)

Mock-up R&D route

15/11/2017

1st phase development

1st phase mock-up specifications • steady state surface heat flux load of 10 MW/m2

• pure tungsten armour and a CuCrZr alloy cooling pipe • design thermo-hydraulic conditions: water at 150 °C and 5 MPa flowing at 16 m/s • free geometry of block and pipe for design FE analysis • HHF testing with cool (20°C) and hot (130°C) coolant temperature, for at least 100 cycles


22

Situation

R&D phase 2

15/11/2017

R&D 2nd phase ….on going but changes on geometrical The changes are …..

2nd phase mock-up specifications • steady state surface heat flux of 20 MW/m2

• pure tungsten armour and a CuCrZr alloy cooling pipe • design thermo-hydraulic conditions water at 150 °C and 5 MPa flowing at 16 m/s • defined geometry of block and pipe for design analysis and manufacturing • HHF testing with hot water (130 °C, 4 MPa, 16 m/s, GLADIS)

Le LW

L

Le

Lg

Dim. Value Units Justification

di 12 mm ITER has optimised thermal hydraulics based on 12mm ID

do 15 mm Pipe wall thickness 1.5mm – motivated by corrosion concern

w 23 mm Analysis shows that reduced width reduces likelihood of deep cracking [M.Li]

a 8 mm Plasma sputtering erosion concern

t 12 mm c.f. 4mm, reduced cost of manufacture, feasibility of attachment to cassette

b (side W wall) Variable by concept

e Interlayer - Variable by concept

c (bottom) Variable by concept

Le 75 mm >=75mm for HHF testing [H.Greuner]

LW 49.5 mm >=49.5mm test section length (4 monoblocks per mock-up)

Lg 0.5 mm >=0.5mm, based on ITER design

T. Barrett, CCFE


23

ITER like concept 1st phase

Objective: validate the qualified solution adopted in ITER for the targets PFC at DEMO operative conditions

15/11/2017

Geometrical size of blocks and pipe optimized by FEA EAP verification by CFD with empirical data for heat exchange coefficient

W block axial thickness 4mm 12mm

CuCrZr-Tube inner diameter 12 mm

CuCrZr-Tube thickness 1.5 mm

Cu-Interlayer thickness 0.5 mm

W-Armour side thickness 6 mm

Plasma Facing thickness 7 mm



Cu-Interlayer thickness 0.5 mm



22mmx24mm 28mmx30mm

JAPAN_ALMT


by Di Maio August 2016

October 2016

Inlet Pressure [MPa] 5.0 5.0

Inlet Temperature [°C] 130 130

DT [°C] 6 6

Removed Power [MW] 126 136

G per Cassette [kg/s] 91.37 98.62

24


Objective: validate the qualified solution adopted in ITER for the targets PFC at DEMO operative conditions

15/11/2017

Geometrical size of blocks and pipe optimized by FEA EAP verification by CFD with empirical data for heat exchange coefficient

W block axial thickness 4mm 12mm



Cu-Interlayer thickness 1 mm

W-Armour side thickness 5.5 mm




Cu-Interlayer thickness 1 mm



23mmx25mm 28mmx29,5mm

CHINA_AT&M



October 2016



DT [°C] 6 6



25


Monoblock design: CuCrZr tube 12mm ID, 1.5mm thick Two geometries ITER-like optimized 12 mm axial thickness DEMO design 4 mm axial thickness

W-blocks ALM&T Japan blocks + ENEA Cu-cast interlayer AT&M China blocks (after FEA optimization) + HIP Cu interlayer

#9 #11 #12

#5

#7

#8

Pipe to block joined by HRP


NDT before HHF: Ultrasonic and SATIR

DEMO #007

A. Durocher, “Infrared thermography inspection of the ITER

vertical target qualification prototypes manufactured by

European industry using SATIR”, Fusion Engineering and Design, 2009



SATIR

• The SATIR test applies an alternating hot/cold flow of pipe coolant while monitoring the temperature response of the tungsten surface by infrared imaging

• A slow surface response respect to reference is an indication of defects in manufacture

• With ENEA’s ultrasonic pulse-echo technique the probe is placed in the bore of the CuCrZr pipe with water fill

• Echo is recorded for specific radial penetration

• Surface map created (unwrapped cylindrical map) at joints

• Calibration by reference component with known defect

DEMO #007

DEMO #011

>8oC = Thermal Imperfection

S. Roccella et al., Development of an ultrasonic test method

for the non-destructive examination of ITER divertor components, Fusion Engineering and Design, 84 (2009)

NDT before HHF: Ultrasonic and SATIR

Dtrefmax

of each monoblock (°C)

Monoblock 12 11 10 9 8 7 6 5 4 3 2 1

Mock-up #5 Reference for #8

Mock-up #7 9 (F)

3,9 (R)

4,1 (L)

4 (B)

3,6 (L)

2,9 (L)

3,1 (L)

2,7 (L)

4,5 (F)

Ref. 2 (F)

16 (R)

Mock-up #8 32 (F)

9,1 (R)

5 (F)

4,2 (L)

4,2 (F)

5,1 (F)

5,1 (R)

3,9 (B)

6,1 (R)

5,3 (R)

4,6 (B)

25 (B)



SATIR

Mock-up ID W block Ultrasonic

mm # block defected

DEMO #005 4 OK

DEMO #007 4 #1 #12

DEMO #008 4 #1 #12

ITER-like #009 12 OK



Ultrasonic check

BATCH 3

DtRefmax of each tile (°C)

(Position R=Right, F= Front, L=Left and B=Back)

Monoblock 4 3 2 1

Mock-up #9 5.7 (F) 3.6 (F) 3.8 (L) 3.5 (R)

Mock-up #11 Ref.

Mock-up #12 2.9 (R) 1.2 (L) 1.4 (L) 1.4 (L)

HHF tests in GLADIS

First PHASE HHF -> cold water (20°C, 10 bar, 12 m/s)


Mock-up ID W block Pipe screening Cycling

SATIR Ultrasonic HHF mm ID/OD mm MW/m² n° MW/m²

DEMO #005 4 12/15 20 100 10 OK OK no defect

DEMO #007 4 12/15 20 100 10 OK OK 1 defect

DEMO #008 4 12/15 20 OK OK Hot spots

ITER-like #009 12 12/15 20 100 15 OK OK no defect



SECOND PHASE HHF -> hot water (130°C, 40 bar, 16 m/s)

Mock-up ID W block Pipe screening Cycling

SATIR Ultrasonic HHF mm ID/OD mm MW/m² n° MW/m²

DEMO #005 4 12/15 6-20 300 20 OK OK no defect

ITER-like #011 12 12/15 6-20 300 20 OK OK no defect


Thermal Break Concept

Slide 29

Aims to overcomes problems of differential thermal expansion of joined materials:

Cause: 4:1 Ratio of expansion coefficients. (αw ~4με/˚C αcucrzr~17με/˚C)

Solution: Use interlayer resistance to manipulate temperature so (α.ΔT)w= (α.ΔT)cucrzr

• Also Increased interlayer

thermal resistance deflects heat “flow” around pipe to reduce one sided heating.

0123456789

10111213

0 200 400 600 800 1000

Ra

dia

l d

ista

nce

/m

m

Temperature / degC

ITER-like concept


Ref. Barrett et al. [2]

F.Domptail, M.Fursdon, T.Barrett


http://www.google.co.uk/url?sa=i&rct=j&q=&esrc=s&source=images&cd=&cad=rja&uact=8&ved=0CAcQjRxqFQoTCKmqn6rvvMgCFQTWFAodAjIG-g&url=http://www.euro-fusionscipub.org/&psig=AFQjCNGJth66e6wSFNZQx-UQVp0F89tquQ&ust=1444737115749354

1. Silver plated feltmetal 2. Copper Foam

• Construction

Strong mechanical joint

Soft Thermal break material

Split tungsten relieves pipe stress and anticipates deep cracking “failure” mode.

2mm

Tungsten blocks in 2 halves

8mm ERG Duocel®

Ref. Crescenzi et al. [1]

Studies made to asses the capability of these materials:

• Felt: unable to sustain the high strain expected with required construction method

• Foam: very ductile but available densities of foam were too low resulting in melting at very high heat flux

Design & Development: starting from available materials




Concept Development : 1st phase

Stress reduction in the pipe increases strain in the interlayer… BUT Design considered acceptable for Phase #1 “exploratory” testing

Radial spoke

Strain at 10MW/m2

εmax~1.8%

εtyp =0.8%

Radial spokes

Final design 3Sm reserve factor: 1.4

As-manufactured mock-up:

Model:

Slide 31




1. Tungsten monoblock supplied with copper cast into bore – core of copper is machined out

2. Braze op. #1 : Cu interlayer “sleeve” is brazed to CuCrZr pipe and groves are machined

3. Braze op.#2 : W-Cu block to Cu-CuCrZr pipe/sleeve

4. Precipitation harden CuCrZr during Braze cycle

Manufacturing: Sequence of assembly of mock-ups:

tungsten

Cast Copper

Cu sleeve CuCrZr pipe

Machined groves

Braze #2

Braze #1

Slide 32




Manufacturing: Final products

TBCI-w27-split 4 mm axial thickness

ALM&T Japan blocks

CuCrZr pipe: 10mm ID, 1mm thick


TBCI-w22-split 4 mm axial thickness



Acceptance check: : SATIR + ENEA Ultrasound

Slide 34

Results for Cu-CuCrZr interface

Potential defect

Results for W-Cu interface

S. Roccella, E. Visca, 2016


F. Gallay, M. Richou, 2016



W/Cu composite concept


⇛ Processing of W fibres to obtain ‚pipe‘ preforms by braidings

⇛ W fibre-reinforced Cu: high-strength drawn W fibres used as reinforcements

(a) Cylindrical multi-layered braiding made out of continuous high-strength W fibres with a nominal diameter of 50 μm;

(b) SEM image of the braiding illustrating the arrangement of individual W fibres;

(c) metallographic SEM image of a longitudinal cross section of a drawn K doped W fibre showing the typical elongated grain structure

(a) transversal and (b) axial microsections by means of optical microscopy

Manufacturing of W fibre-

reinforced Cu composite by

means of liquid Cu melt

infiltration

A. Von Muller

W fibre-reinforced Cu

Necking of W fibres plainly visible with ductile failure of both Cu matrix

& W fibrous reinforcements (already at room temperature)

SEM fracture surface images of a W fibre-reinforced Cu pipe specimen (Cu infiltrated

W fibre braiding); after axial tensile test of pipe at room temperature

100 µm

36 15/11/2017 2nd IAEA Technical Meeting on Divertor Concepts

A. Von Muller


Mock-up manufacturing


⇛ HHF testing (GLADIS) of W fibre-reinforced Cu heat sink pipe

Successful HHF testing up to 25 MW/m2

W fibre-reinforced Cu pipe

before HHF testing

W fibre-reinforced Cu pipe

during HHF testing

W fibre-reinforced Cu pipe after

HHF testing

⇛ HHF testing (GLADIS) of monoblock-type mock-up

with W fibre-reinforced Cu heat sink pipe

Brazed joint between W monoblocks and W fibre-

reinforced Cu heat sink pipe (ENEA)

Successful HHF testing up to 100 load cycles at 20

MW/m2 without indication of failure

UT

HHF

A. Von Muller


38

W particle-reinforced Cu composites by melt infiltration

(industry)

⇛ Exemplary property

thermal conductivity

⇛ Range of composition: 60wt.% to 90wt.% W

(~ 40vol.% to 80vol.% W) (a) optical as well as (b) SEM micrographs of a W particle-reinforced

Cu composite with a composition of 60wt.%W–40wt.%Cu

Measured thermal conductivity of melt

infiltrated W particle-reinforced Cu

composites with differing compositions

in comparison to pure W

A. Von Muller


39

W flat tile mock-ups with W particle-reinforced Cu composite

(50/50vol.%) heat sink

Successful HHF tests (GLADIS) up to ~360 load cycles at 20 MW/m2

W armour tiles bonded to composite heat sink during material melt infiltration

W flat tile mock-up with W particle-reinforced Cu

composite (50/50vol.%) heat sink block

Side view of W flat tile mock-up with

W particle-reinforced Cu composite

(50/50vol.%) heat sink

Optical image of W flat tile mock-up

with W particle-reinforced Cu

composite (50/50vol.%) heat sink

block during HHF testing Alexander v. Müller, IPP Garching


FGM Concept- Functionally Graded Material (W/Cu)

40

• Coating in W block Functionally Graded Material (W/Cu) Variation of composition from 100 % of W (0% Cu) to 100% Cu (0% W) Thickness graded layer = 20µm

• CuCrZr pipes (1mm thickness)

• Hot Isostatic Pressing (950°C, 120 MPa, 2h)

150mm

CuCrZr W & FGM

SEM magnification X4000

FGM W


M. Richou

Cu

41

SATIR acceptance check


F. Gallay, M. Richou

Mock-up ID

W block

Ultrasonic

mm # block defected Note

#001 4 OK Internal pipe scratches

#002 4 #1 #12

#003 4 OK

#004 4 OK

#005 4 OK

#006 4 #12

ENEA ultrasonic check



HHF tests in GLADIS: up to 25 MW/m² and under cycling tests (300 cycles

up to 20 MW/m²)

Experiment confirmed relatively low operational temperature at

20 MW/m² (1450°C)

No impact of graded layer thickness heterogeneity on thermal

performance

H. Greuner, next talk


Laminate W – Flat tile concept

The idea of the “Flat tile” concept is to use W only in the hot region of the mockup.

Possibilty to produce mockups (i) without interlayer, (ii) with an interlayer (W-Cu laminate), (iii) without castellation and (iv) with castellation. 16 mockups have been produced.

Castellation

Tested in GLADIS (2016, H. Greuner, IPP)

Without interlayer

With interlayer


2nd Phase Started begin 2017 and on going ….


Extension of HHF testing to higher cycle numbers Cu-interlayer thickness: comparison FGM to ITER-like with 10/100/300 mm thickness Model validation for failure (induced creek in W)

Standardize pipe and W-block dimensions for all concepts under development

Surface W thickness >= 8mm (erosion)

Pipe wall thickness 1.5 mm (corrosion)

W-block axial thickness 12 mm

W-block width 23 mm

4 W-block for mock-up

Gap among W-blocks 0,5 mm

Phase 1 Phase 2

20 MW/m2

2nd Phase


• Material procurement

• Developing of design tools for prediction

• Manufacturing of preliminary samples

• Fabrication of standard mockups

• ND checks by Ultrasonic and SATIR

• HHF testing in GLADIS

KME Elbrodur HF CuCr1Zr tubes (780 kg) 15mm OD, 1,5 mm thick, SAcWA conditions

AT&M (China) pure W blocks

2nd Phase


• Material procurement

• Developing of design tools for prediction

• Manufacturing of preliminary samples

• Fabrication of standard mockups

• ND checks by Ultrasonic and SATIR

• HHF testing in GLADIS

Phase 2 Design for Thermal-Break concept: Optimization with updated 2017 specification (different overall dimension, constraints & heat flux of 20MW/m2)

• Developed a tailored response surface and optimisation process to explore different design efficiently by FEA

• Some examples of designs being explored:

2nd Phase


Thick graded layer for stress-relieving effect, ~500 µm by cold spray

Manufacturing of FGM mockups with required thickness and

composition

14 mm

W-fibre-Cu mockup

Firsts ITER-like mockups to be used for comparison with other concepts

Conclusions


ITER Plasma Facing components qualification strategy can be used as baseline for divertor target design concepts of European DEMO reactor

Water cooled divertor requires using of Copper for heat sink. So operating conditions of the European DEMO divertor must be set as a compromise between keeping sufficient margins w.r.t. the critical heat flux (Tin~150°C) and maintaining sufficient structural integrity during irradiation. Temperature range between 150-200°C and 350°C

Development and characterization of new materials (composites) can be surely improve performances

Innovative approach in design (like Thermal-Break concept that also takes advance of FEA) have been successful tested

Technological improvement of graded material and their joining when applied to plasma facing component fabrication have given very good results

All these manufacturing and concepts developments have to be validated with respect to ‘reactor relevant’ conditions in terms of large scale fabrication, plasma affinity, and reliability also after n-irradiation

Thank you for your attention


50

ITER like concept

Objective: validate the qualified solution adopted by ITER for the targets PFC at DEMO operative conditions 1)From ITER to DEMO divertor:

• Water cooled or He cooled? • Comparison of operational parameter(hydraulic, load

specification, neutronic aspects, materials) • Why copper is considered for DEMO divertor yet? • Results of HHFT from ITER mock-ups and metallographic

images of copper interlayer damaged –> is interlayer needed? • Is available a reinforced copper with lower CTE and higher

neutron resistance?

15/11/2017


October 2016



DT [°C] 6 6




• Representative size: 120mm

• Simple distribution manifold geometry

• „Standardized“ W-slab geometry

• Mock-up 01: W-Cu laminate pipe, 5mm W-slab

• Mock-up 02.1: W-Cu laminate pipe, 12mm PIM W-slab

• Mock-up 02.2: EUROFER pipe

• Testing window: end 2017- begin 2018

500°C (?); 10MPa

In progress

Helium-cooled concept

15/11/2017 51 2nd IAEA Technical Meeting on Divertor

Concepts

15/11/2017 52

An additional critical issue in the design of water-cooled components is the transition from nucleate boiling (two-phase flow) to the boiling crisis or ‘burnout”. The burnout is characterized by the collapse of the heat transfer and the resulting sudden temperature excursion of the heated material. Cu based heat sink materials melt immediately. The threshold heat flux for this to occur is called the critical heat flux (CHF) [55,56]. Depending on the component design, the concentration of the heat-flux from the loaded surface to the inner cooling wall results in a peaked heat flux at the inner cooling wall and CHF could occur for moderate incident heat fluxes. The local CHF strongly depends on the coolant velocity, the coolant pressure, and the local coolant temperature. The coolant velocity is limited either by engineering constraints like erosion by corrosion of cooling tubes or by the pumping power needed to maintain the coolant flow. To achieve a sufficient safety margin against the CHF the inlet temperature has to be in the range of 150°C. [57] Therefore, the operating conditions of the divertor must be set as a compromise between keeping sufficient margins w.r.t. the critical heat flux (Tin~150°C) and maintaining sufficient structural integrity during irradiation. Traditional design rules for the case of Cu-alloy, based on the limited available irradiation data at relevant conditions, require operation in a temperature range between 150-200°C (to minimise radiation embrittlement) and 350°C, above which material properties deteriorate. A modern engineering design approach and design rules, to be further developed, instead of the conservative historical approaches, is required as there are numerous instances where low ductility occurs while the fracture toughness is high (see, e.g., [58]). Analyses should consider the full range of applicable situations including low ductility scenarios as well as crack propagation.


Session 4: Plasma facing components - Nucleus Documents... · 15/11/2017 1 Session 4: Plasma facing...

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