Materials Challenges for ITER - CERNcern.ch/lhc-collimation-project/ph2_meeting_files/ICFRM...

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Vladimir Barabash and the ITER International Team, A. Peacock2, S. Fabritsiev3, G. Kalinin4, S. Zinkle5, A. Rowcliffe5,

J.-W. Rensman6, A.A. Tavassoli7, P.J. Karditsas8, F. Gillemot9, M. Akiba10

1 ITER International Team, 85748 Garching, Germany2 EFDA Close Support Unit, 85748 Garching, Germany

3 D.V. Efremov Scientific Research Institute, 196641 St. Petersburg, Russia4 ENES, P.O. Box 788, 101000 Moscow, Russia

5 ORNL, P.O. Box 2008, Oak Ridge, TN 37831-6138, USA6 NRG, P.O. Box 25, 1755 ZG Petten, The Netherlands

7 CEA/Saclay, 91191 Gif sur Yvette Cedex, France8 EURATOM/UKAEA Fusion Association, Abingdon, Oxon, OX14 3DB, UK

9 AEKI Atomic Research Institute, 1121 Budapest, Hungary10 JAEA, Naka-machi, Naka-gun, Ibaraki-ken 311-0193, Japan

Materials Challenges for ITERMaterials Challenges for ITER

-- Current Status and Future ActivitiesCurrent Status and Future Activities

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Outline

ITERITER-- Status of projectStatus of project-- Materials activity strategyMaterials activity strategy

Materials selection for different componentsMaterials selection for different components-- Vacuum vesselVacuum vessel-- InIn--vessel componentsvessel components

-- Stainless steel 316L(N)Stainless steel 316L(N)-- CuCrZr alloyCuCrZr alloy-- Alloy 718Alloy 718-- TiTi--6Al6Al--4V4V-- NiAl bronzeNiAl bronze-- Armour materials (Be, W, CFC)Armour materials (Be, W, CFC)

Test Blanket Modules ActivitiesTest Blanket Modules ActivitiesSummarySummary

V.Barabash et al., ITER Materials, ICFRM-12, December 4 – 9, 2005, Santa Barbara, USA

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Status of ITER

ITER is international fusion energy research project. ITER is international fusion energy research project.

ITER goals:

Show scientific and technological feasibility of fusion energy for peaceful purposes

Integrate and test all essential fusion power reactor technologies and components

Demonstrate safety and environmental acceptability of fusion

ITER activities were started in 1992, Final Design Report was issued in 2001.

Extensive negotiations about construction agreement, selection of site, sharing between Parties, etc. are being carried out during 2001-2005.

Finally, in June 2005, the ITER construction site was selected – Cadarache, France.


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ITER Design Parameters (FDR 2001)

Total Fusion Power 500 (700) MW

Q=Fusion Power/Heating power

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Plasma major radius 6.2 mPlasma minor radius 2.0 mToroidal field, 6.2 m 5.3 TPlasma current 15 (17.4) MAPlasma Volume 837 m3

Plasma Surface 678 m2

Neutron flux Av. 0.57 MW/m2

Max. 0.8 MW/m2

Neutron fluence Av. 0.3MW*a/m2

Max 0.5MW*a/m2

Heat flux, First Wall

0.2 - 0.5 MW/m2

Heat flux, Divertor 10 (20) MW/m2

Number of pulses ~ 30.000Pulse length ~ 400 s

ITER is the first tokamak, with significant neutron fluence. ITER is the first tokamak, with significant neutron fluence. ITER is the first tokamak, with significant neutron fluence.

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Selection of materials for the ITER components was a challenging task, due to unique operational conditions, e.g.:

Selection of Materials

• Magnet – operation at 4K, high stresses;

• In-vessel components – combined effect of n-irradiation, heat flux, hydrogen environment, mechanical loads, complicated design with different type of joints;


Experience from current tokamaks has been taken into account,(plasma facing, diagnostics materials, etc.).

Knowledge from fission neutron irradiation programs was used.

However, in many cases, the choice was not simple, because for anumber of materials the existing data base was not sufficient.

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The material choice for ITER has been orientated toward industrially available materials while taking account of physical and mechanical properties, maintainability, reliability, corrosion performance and safety requirements at the ITER operational conditions.

Selection of Materials – General strategy

The ITER materials categories:The ITER materials categories:

• Standard materials with established manufacturing technologies, (e.g. steel 304, steel 316L(N), alloy 718, W).

• Standard materials, which require some modifications, such as more stringent limits on the alloying elements, etc. (e.g. steel EK1, JJ1, CuCrZr, etc. ).

• Recently developed materials (e.g. CFC).


Materials were selected based on the evaluation of the impact of manufacturing technologies on materials properties and assessment of their performance at the ITER operational conditions.

These selection and justification were supported by the results of the dedicated world-wide R&D program.

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Materials documentation – FDR 2001 (2004)ITER Materials Properties Handbook (MPH-Magnet, Cryo, VV&IC):- A collection of design-relevant data on physical and mechanical properties of a large variety

of the ITER relevant materials.

ITER Materials Assessment Report (MAR):- A description of the rationales for the selection of the specific materials grades for VV and In-

vessel components and assessment of the materials performance.

Appendix A of the ITER Addendum to the RCC-MR:- Design data for materials for the Vacuum Vessel.

Appendix A of the ITER Structural Design Criteria (SDC-IC):- Design data for in-vessel components (including neutron irradiation).

Safety Report:- Collection of safety relevant properties.

ITER Materials Properties Database (MPDB):- Collection of raw experimental data - base for further recommendations

These documents justify the selection and materials performance.These documents justify the selection and materials performance.


Codes, R&D reports,Open sources

Codes, R&D reports,Open sources

Database(MPDB)

Database(MPDB)

ITER MPH,Experts Activity

ITER MPH,Experts Activity

Design DocumentsDesign

Documents

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Materials Activity in current stage

Multidisciplinary task:Multidisciplinary task:

• Preparing of the procurement specifications for materials for various components (in accordance with construction schedule); - Discussion with Industries for implementing the ITER requirements- Discussion of the possible deviations- Schedule for materials procurements

• Consolidation of the materials properties database and modifications of the ITER Materials Documents;- Assessment of the new data from R&D program- Preparing of recommendations based on Design Codes requirements- Completion of MPH and App. A for VV Code, In-vessel SDC and Magnet SDC

• Carrying out the R&D in some still critical areas:- Further optimisation and simplification of design and manufacturing route- Generating of new data needed for the completion of the design evaluation


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No. WBS Element 1. TOKAMAK BASIC MACHINE 1.1 TOROIDAL FIELD (TF) COILS SYSTEM 1.2 POLOIDAL FIELD (PF) COILS SYSTEM 1.3 CENTRAL SOLENOID (CS) SYSTEM 1.5 VACUUM VESSEL 1.6 BLANKET SYSTEM 1.7 DIVERTOR 2. TOKAMAK ANCILLARIES AND CRYOSTAT 2.3 REMOTE HANDLING (RH) EQUIPMENT 2.4 CRYOSTAT 2.6 COOLING WATER SYSTEM 2.7 THERMAL SHIELDS 3. TOKAMAK FLUIDS 4. POWER SUPPLIES - COMMAND CONTROL 5. PORT INTERFACING SYSTEMS 5.1 ION CYCL. HEATING (IC H&CD) SYSTEM 5.2 ELECTRON CYCL. HEATING (EC H&CD) SYSTEM5.3 NEUTRAL BEAM HEATING (NB H&CD) SYSTEM 5.4 LOWER HYBRID HEATING (LH H&CD) SYSTEM 5.5 DIAGNOSTICS 5.6 TEST BLANKETS

Selection of materials for ITER components

Magnet Magnet –– T operationT operation-- 44--77 K77 K

Coating, 5 μm (emissivity)Ag coating

InsulationGlass epoxy G10

Plasma sprayed insulationAl2O3 coatings

BoltsAlloy 718

FastenersSteel grade 660

PlatesTi-6Al-4V

Main structural material: - plates, tubesSteel 304L

Thermal Shield Materials (77Thermal Shield Materials (77--300K)300K)

Vacuum vessel and In-vessel materials: - structural - plasma facing

1.5 VACUUM VESSEL

1.6 BLANKET SYSTEM

1.7 DIVERTOR

5.1 ION CYCL. HEATING (IC H&CD) SYSTEM

5.2 ELECTRON CYCL. HEATING (EC H&CD) SYSTEM

5.3 NEUTRAL BEAM HEATING (NB H&CD) SYSTEM

5.4 LOWER HYBRID HEATING (LH H&CD) SYSTEM

5.5 DIAGNOSTICS


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Materials for Vacuum Vessel:Austenitic steel 316L(N)-IG plates, forgings, rodsBorated steels 304B7 and 304B4 shielding platesFerritic steel 430 platesSteel 660 fasteners, forgingsSteel 304 plates, beamsAlloy 718 boltsWeld filler materials electrodes, wireXM-19 bolts for shieldingSteel 316 bolts (B8M)Pure Cu clad on VV

Materials for Vacuum Vessel support:Steel 304 plates, forgings, rodAlloy 718 boltsSteel 660 forgings, boltsNiAl bronze pinsPTFE for sliding elementsNeoprene rubber

Selection of materials – Vacuum Vessel- Safety Important Component => licensing in France- Code – RCC-MR + ITER addendum- Main structural material – 316L(N)-IG is the Code qualified and main specifications are available- Other structural materials: steel 304, steel grade 660, Alloy 718 are also in the Code- Only functional materials (borated steels, etc.) are not in the Code.


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Selection of materials – Vacuum Vessel

Element

316L

ASME A240

Z2 CND 17-12 controlled nitrogen

RCC-MR 2002

316L(N)-IG ITER grade

min max min max min max C 0.03 0.030 0.03

Mn 2.00 1.60 2.00 1.60 2.00 Si 0.045 0.50 0.50 P 0.030 0.035 0.025 S 0.75 0.025 0.01

Cr 16.00 18.00 17.00 18.00 17.00 18.00 Ni 10.00 14.00 12.00 12.50 12.00 12.50 Mo 2.00 3.00 2.30 2.70 2.30 2.70

Nb+Ta+Ti 0.15* Cu 1.00 0.30 B 0.0020 0.002

(0.001)** Co 0.25 0.05 N 0.1 0.060 0.080 0.060 0.080

Sm, RT

115

147

147 Nb < 0.1, Ta < 0.01

Materials for VV – ongoing activity:Complete App. A to ITER Addendum to RCC-MR

- Introduce required additional properties (e.g. fracture toughness) for structural materials- Introduce properties of functional materials in the ITER Addendum

Non-metallic materials – justification of the properties for licensing- Windows, etc.

Properties of welds- In case of use welds, which are not in Code, – provide the needed justification.

Preparation of the Procurement Specifications- Composition, minimum properties, NDE&QA

Chemical composition of steels

0

20

40

60

80

100

120

140

160

0 50 100 150 200 250 300 350 400 450

Temperature, °C

Des

ign

Stre

ss In

tens

ity S

m, M

Pa

316L(N)-IG

316L, Sec. II, Part D, 2001

316LN, Sec. II, Part D, 2001

316L(N)-IG, RCC-MR, Ed. 2002

316L, ASME Sec. II, Part D, 2004

316LN, ASME Sec. II, Part D, 2004

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General considerations: • Design of structural elements – in accordance with the ITER Structural Design

Criteria for IC (SDC-IC)• Design of some elements (Be/Cu, CFC/Cu, Window/Cu etc.) – by experiment

ITER SDC-IC:• Needs reliable and traceable materials data (including neutron irradiation effect) • Properties of joints (welds, SS/Cu, etc.) have also to be assessed

ITER MPH includes the justification of recommendations.

Properties needed for design analyses:- Physical (incl. neutron irradiation) - average (+deviation)- Tensile properties for design (incl. neutron irradiation) - minimum and average- Design fatigue curves (incl. neutron irradiation) - minimum (∆ε/2; Nf/20)- Fracture toughness (incl. neutron irradiation) - minimum and average

Specifications for materials supply: - Chemical composition,- Minimum properties;- NDE&QA

Selection of materials – In-vessel components


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Selection of materials – In-vessel components

Square waveguides Front mirrors

Windows~ 4 m

Material Material Grade Components • Armour Beryllium S-65C VHP or equivalent • Armour for first wall and

limiter Tungsten Pure W • Armour for divertor CFC SEP NB 31 or equivalent • Armour for divertor

• Structural 316L(N)-IG • Blanket shield modules

• Thin walled pipes • Cooling manifolds • Divertor body

Austenitic steels

304L, 316L • Cooling pipes XM-19 • Divertor support PH steel Steel grade 660 • Divertor support

CuCrZr • PFCs, heating systems, electrical strips, etc.

NiAl bronze • Divertor attachment CuNiBe • Support system

Cu alloys

DS Cu • heating systems,

• Functional Austenitic steel

316 • Fastening components

PH steel Steel grade 660 • Fastening components Ni alloys Alloy 718 • Bolts

• Divertor connections Ti alloy Ti-6Al-4V • Blanket attachment Ceramic Al2O3 or MgAl2O4 • Electrical insulators Special materials

• Pure Cu • Pure Ni • Low C iron • Brazes • Weld fillers


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Stainless steel 316L(N)-IG

ITER recommendation for the reweldability limits, provided by the allowable levels for He production in weld:

– < 1 appm for thick plate welding,– < 3 appm for thin plate or pipe welding.

Rewelding

K. Asano, J. van der Laan, MAR, 2001V.Barabash et al., ITER Materials, ICFRM-12, December 4 – 9, 2005, Santa Barbara, USA

Different types of steels (different treatment, etc.):- Wrought- HIP- Powder HIP - have to be qualified!- Cast- Welds

Welding and joining

100

200

300

400

500

600

700

800

0

10

20

30

40

50

60

0 5 10 15 20

Weld Metal (19-12-2)Irr and test - 400°C

Elon

gatio

n, %

Dose, dpa

U.T.S

Y.S.

T.E.

U.E.

A.A. Tavassoli, this Conf.

• Un-irradiated data for welds are Code qualified;• Irradiated data are being reviewed, the behavior

is similar to base metal; the same safety margin can be used;

• The properties of joints after HIP (solid and powder) joining technique are adequate.

15

0

5

10

15

20

25

30

35

40

45

0 50 100 150 200 250 300 350 400

T test = T irr, oC

Spe

cific

atio

n U

EL

min

, %

unirradiated

2 dpa

5 dpa9 dpa

irradiation effect

Stainless steel 316L(N)-IG

Neutron effect: Tensile properties:-Strengthening and loss of ductility (loss of strain hardening)

Fatigue:– no effect on fatigue

Fracture toughness:– reduction, but material still ductile


0

100

200

300

400

500

600

700

800

0 1 2 3 4 5

D, dpa

Spe

cific

atio

n S

y, m

in, M

Pa

100oC200oC300oC

0.1

1.0

10.0

1.00E+02 1.00E+03 1.00E+04 1.00E+05 1.00E+06 1.00E+07

Fatigue life Nf

Tota

l str

ain

rang

e %

RCC-MR ed. 2002 20°C design curveUnirradiated data Ttest ≤ 250°C[2] Källstrom, Tirr = 35°C, Ttest = 25+250°C, 0.3 dpa[3] Horsten, Tirr = Ttest = 227+327°C, 7.5-11.2 dpa[4] Horsten, Tirr = Ttest = 77°C, 0.58-0.83 dpa[5] Lind, Tirr = Ttest = 290°C, 0.73+2.5 dpa[6] Puzzolante, Tirr = 42°C, Ttest = 25°C, 5.1-5.4 dpa[7] Bergenlid, axial, Tirr = 250°C, Ttest = 200°C, 2.8+9.2 dpa[8] Vandermeulen, Tirr = Ttest = 250°C, 6.2-10 dpa Best fit to unirradiated data (aNf^b)Best fit to irradiated data (aNf^b)Fit irr data displaced by Nf/20Fit irr data displaced by eps/2

0

100

200

300

400

500

600

700

0 100 200 300 400 500

Temperature, °C

Min

imum

toug

hnes

s va

lues

Lower Bound JIC (kJ/m2)

3 dpa = *.76

5 dpa = *.61

J.W. Rensman, MPH 2004&2005

J.W. Rensman, MPH 2005

G. Kalinin, MPH 2005

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Alloy Designation Cu Cr Zr O Other elements

CuCrZr, RF base 0.4-1.0 0.03-0.08 0.1 max CuCrZr, C18150 base 0.5-1.5 0.05-0.25 CEN/TS 13388 base 0.3-1.2 0.03-0.3 0.2 CuCrZr ITER Grade base 0.60-0.90 0.07-0.15 0.002 max 0.01 max

0.05 -Cd

Chemical compositionChemical composition

Reasons for modification of the chemical composition:- smaller scatter of properties, - less coarser Cr particles – better toughness, - better radiation resistance is expected,- welding improved.

CuCrZr alloy

Effect of treatment on properties.

Manufacturing treatments:(SA – 980-1000°C + water quench, Ageing – 450-500°C• SAA (different cooling rate, etc.)• SA + cold work + A• SAA + cold work• cast CuCrZr + SAA

0

100

200

300

400

0 100 200 300 400 500

Temperature, °C

Su m

in a

nd S

m, M

Pa

Su, min, SAA

Su, min SA+560C/2hSm, = 1/3 Su, min (SAA)

Sm=1/3Su min (SA+560C/2h)

Barabash, Zinkle, MPH 2005


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CuCrZr alloy

Fatigue: - Seems that fatigue performance is similar for CuCrZr with different treatment

Creep – fatigue interaction:– very complicated performance:

=> Generally there is reduction of lifetime with hold time; => Higher Sy – higher fatigue lifetime

(see presentation of B. Singh, this Conference)

1.0E-02

1.0E-01

1.0E+00

1.0E+01

1.E+02 1.E+03 1.E+04 1.E+05 1.E+06

Number of cycles, Nf

Tota

l str

ain

rang

e, %

Design fatigue curve, SAA, 20 - 350°CExperimental data, CuCrZr SAA, 20 - 350°CSingh, SA+underageing, 250-350C°, unirradiated

Average limited data, FDR 2001Design curve for pure annealed Cu

Design fatigue curve:- min of average of delta et/2 and Nf/20

M. Li, P. Marmy, B. Singh, MPH 2005

I. Bretherton, 2001, MPH 2005


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CuCrZr alloy

Neutron effect: Tensile properties:- loss of ductility (loss of strain hardening):

saturation at ~ 0.1-0.5 dpa at T ~ 100-300°C. - Some improvements of ductility is observed

after baking treatment (see Fabritsiev, this Conference)

Fracture toughness and fatigue:– effect is small, but data are only up to 0.3 dpa

There is no correlation between loss of ductility at tensile tests and fracture toughness.

0

100

200

300

400

500

0 50 100 150 200 250 300 350 400

Temperature (oC)

valu

e of

J in

itiat

ion

J Q (k

Jm-2

)

SEN(B) B=3mm W=4mm [4-6]

SEN(B) B=3mm W=4mm [7]

Polyn. fit (all)

Polyn. fit (unirrad.)

0.3 dpa

S. Tähtinen, MPH 2005

0

100

200

300

400

500

0.0 1.0 2.0 3.0

Damage dose, dpa

Su ir

r and

Sy

irr,

MPa

Su, irr, T = 150°C

Sy, irr, T = 150°C

Su, irr, T = 300°C

Sy, irr, T = 300°C

V.Barabash & S.Zinkle, MPH 2005

0

5

10

15

20

0.0 1.0 2.0 3.0

Damage dose, dpa

Uni

form

elo

ngat

ion,

%

UA, T = 150°C

UA, T = 300°C

ISDC limit 2 %

Summary for CuCrZr:• Final characterization after selected

heat treatment (s);• There are some missing data:

irradiation creep, fracture toughness and fatigues at high doses;

• Importance of creep-fatigue interaction shall be understood.


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Alloy 718

Key issues:

Stress relaxation -bolts:- ITER goal ~ 0.1 dpa

Loss of strength at low damage dose:– effect is ~ 15%.

J.W. Rensman, MPH 2005T.S. Byun, ORNL

y = e-0.694*x


J.W. Rensman, MPH 2005

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Reasons: low elastic modulus and high strength provide high flexibility.

Ti-6Al-4V alloy

Key issues: Neutron irradiation embrittlement:- ITER goal ~ 0.1 dpa

Hydrogen environment:– it is well known that H affects properties.

0

200

400

600

800

1000

1200

0.0 0.1 0.2 0.3 0.4 0.5

Damage dose, dpa

Stre

ngth

, M

Pa

Su, 50°C Sy, 50°C

Su, 260°C Sy, 260°C

Su, 350°C Sy, 350°C

0

5

10

15

20

25

0.0 0.1 0.2 0.3 0.4 0.5

Damage dose, dpa

TE a

nd U

E, % TE, 50°C UE, 50°C

TE, 260°C UE, 3260°C

TE, 350°C UE, 350°C

MPH 2005: S. Tahtinen, B. Rodchenkov

Hydrogen levels up to 150 wppm have moderate effect on toughness, at a test temperature of 150°C. P. Marmy, 2004

0.12 – 0.15 dpa


However, remaining issues:

• Role of ionizing radiation on possible enhanced hydrogen uptake (MIT launcher, May 2005)

• Fracture toughness and fatigue of irradiated Ti alloy.

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NiAl bronze (NiAl bronze (CuAl10Ni5Fe4)CuAl10Ni5Fe4)::ASTM B150, C63200DIN EN 12163/1216

High strength, low friction and spark resistance are the key properties.

NiAl bronze

0

5

10

15

20

0.0001 0.001 0.01 0.1

Damage dose, dpa

Elon

gatio

n, %

Total elongation, Ref. 1, Fabritsiev (Cu127), Ttest = Tirr = 150°CUniform elongation, Ref. 1, Fabritsiev (Cu127), Ttest = Tirr = 150°CTotal elongation, Ref. 1, Fabritsiev (Cu127), Ttest = Tirr = 300°CUniform elongation, Ref. 1, Fabritsiev (Cu127), Ttest = Tirr = 300°C

Un-irradiated

Trend at 150 C

0

Trend at 300C

0

200

400

600

800

1000

0.0001 0.001 0.01 0.1

Damage dose, dpa

Tens

ile s

tren

gth,

MPa

Ref. 1, Fabritsiev (Cu127), Ttest = Tirr = 150°CRef. 1, Fabritsiev (Cu127), Ttest = Tirr = 300°C

Un-irradiated

Trend

0

V. Barabash, S. Fabritisiev, this Conference

Neutron effect: First data are presented at this Conference.


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Selection of armour materials

Compromise: Plasma Performance Materials Lifetime T retention

~ 680m2 Be first wall low Z compatibility with wide operating range and low T retention

~ 50 m2 CFC Divertor Target No melting under transients (ELMs and Disruptions)Low Z compatibility with wide range of plasma regimes (Te,div ~ 1 – 100 eV)Large T retention (co-deposition)

~ 100m2 Tungsten Baffle/Dome Low Erosion, long Lifetime and low T retention

For PSI issues see:G. Federici, “Plasma Wall Interactions in ITER”, to appear in Physica Scripta.G. Federici, et al., “Plasma material interactions in current tokama and their implications for next step fusion reactors”. Nucl. Fusion 41 (2001 )1967.

23

0

50

100

150

200

250

300

350

400

450

0 500 1000 1500 2000 2500 3000 3500 4000

Temperature, CTh

erm

al c

ondu

ctiv

ity, W

/mK

C42, C32

C30

C27

C11

C61

C62

C59

Carbon Fibre Composites

SEP NB31, CX-2002U, or equivalent

0

100

200

300

400

500

600

Ther

mal

con

duct

ivity

, W/m

K

0 500 1000 1500Temperature, °C

CX 2002U

SEP NS 31SEP NB 31

Dunlop, C 1NIC-01

Key issues: Properties variation at mass production (ITER needs ~ 10 tons):- variation of thermal conductivity (SDRT < 10 W/mK);- stable mechanical properties

Neutron effect:– it is well known neutron irradiation affect thermal conductivity, nevertheless the thermal performance is adequate (V.Barabash, ICFRM-11).

Thermal erosion at transient loads:– is being assessed (EFDA+ITER+TRINITI, RF)

Improving of manufacturing technology is ongoing.

24

Beryllium

S-65C, Brush Wellman USA, or equivalent grade(s)

Key issues: Thermal fatigue/thermal shock resistance- S-65C VHP has the better performance, but others may be also acceptable

Neutron effect:– The behavior of all Be grades with Be0 < 1% seems very similar.- However, the performance of irradiated Be in the irradiated components seems good, based on the results of in-pile tests and test of FW mock-ups after irradiation (see J. Linke, A. Gervash).

0

500

1000

1500

2000

2500

Cyc

les t

o C

rack

Initi

atio

n

0 0.5 1 1.5 2

Side Crack Propagation Depth (mm)

S-65C (L)

DShG-200

S-65C (T)

TShGT TShG-56S-65-H

S-200F (L)98% S-65

Extruded(T) TGP-56

Extruded (L)SR-200

S-200F-H94% S-65 I-400 (T)

Be/60%BeO Be/30%BeO

Grades with best fatigueperformance

S-200F (T)

B. Watson, SNLA

Metallography of S-65C and TR-30after disruption heat load

M. Roedig, FZJ

0

10

20

30

40

50

0

100

200

300

400

500

Yie

ld T

ensi

le S

treng

th, M

Pa

0 100 200 300 400 500 600 700

Temperature, °C

S-65

C/U

TS

Yield Strength, irr, L.Snead, 0.34 dpa~ 0.6 dpa Yield Strength, irr ,R.Chaouadi, ~ 1 -2.5 dpaYield strength, unirr.

Ttest = Tirr

0

5

10

15

20

25

30

35

40

45

50

0

10

20

30

40

50

Tota

l Elo

ngat

ion,

%

0 200 400 600 800Temperature, °C

M.Roedig, 0.35 dpa

L.Snead, 0.34 dpaR. Chaouadi, 1-2.5 dpaAverage, unirradiated

Ttest=Tirrad


25

Delamination cracks

Pure sintered W is selectedData base for pure W is sufficient.

Key issues: Thermal fatigue/thermal shock resistance- Performance of W significantly depends on orientation of grainsand production history (plate, rod, recrystallization, etc.)

0

250

500

750

1000

Duc

tile

to B

rittle

Tra

nsiti

on T

empe

ratu

re, °

C

0 25 50 75 100Fluence, 1024 n/m 2 (E>0.1 MeV), 1 dpa ~ 2 1025 n/m 2

Densimet, bending,Tirr = 250-300°C

W, tensile test, Tirr = 371-382°C

W, tensile test, Tirr = 100°C

W-10Re, bending,Tirr = 250-300°C

W, bending, Tirr = 250-300°C

0.42 1024 n/m 2

ITER Goal

Taniguchi, JAERI, 2000

Neutron effect:– Due to low irradiation temperature (~ 150-500°C, all W grades will be brittle at low dose irradiation- However, the performance of ‘brittle’ W in the irradiated components seems good (see M. Roedig, FZJ).


Tungsten

26

Remaining R&D issues


These issues have to be resolved before the starting of the procurements.

ITER is ready to be build, however, there is still a list of materials issues to be solved.

• Site adaptation: for safety important components a complete information have to be presented for the licensing needs in France;

• In some areas R&D is on-going with goal to simplify design, reduce cost, increase reliability – materials data for supporting of these designs are needed, e.g.:

• US propose cast steel for blanket modules – need neutron irr. data;• RF and China – cheaper Be grades – need to qualify them;• Different technologies used for FW – need qualify CuCrZr properties;• Etc.

• Need to have all missing data, which could impact the performance of materials in ITER

• Finally we need to provide the recommended data for Ultimate assessment and acceptance of the design.

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Test Blanket Module Program in ITER:

ITER is considered as a most suitable test-bed for the breeding blanket of next step.

3 equatorial ports are allocated for TBM. Size of port ~1310 x 1760 mm.

Each port could include 2 types of TBM

The first TBMs must be installed in ITER since the start of the operation:

• Finalisation of the design and supporting R&D must be completed soon.

• All modules before installation in ITER must be qualified for licensing in France:

=> sufficient information about materials should be provided.


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Summary

The design and technical preparation for the construction of ITER are ready for implementation and inter-government negotiations are nearing to completion.

The design of the ITER components is supported by the proper and justified materials selection. This is a results of extensive and successful world-wide ITER materials and technological R&D program.

During construction phase the materials activity will be focused on:Resolving of the urgent remaining issues before starting the procurement of materials, Materials procurement and evaluation of the acceptance of the materials in the final components,Further consolidation of the data, which are needed for the licensing and for justification of the safe and reliable performance of materials during the ITER operation.


Materials Challenges for ITER - CERNcern.ch/lhc-collimation-project/ph2_meeting_files/ICFRM...

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Transcript of Materials Challenges for ITER - CERNcern.ch/lhc-collimation-project/ph2_meeting_files/ICFRM...