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Contributions of US Fusion Materials Programto ITER Test Blanket Modules

S.J. Zinkle, on behalf of the US fusionmaterials researchers

ITER Test Blanket Module meeting

UCLA, March 2, 2005

Assessment of structural materials issues for ITERTBMs•New versions of the FM steels F82H and Eurofer with

improved toughness and radiation performance are evolving

•Multiple fabrication stages and thermo-mechanical treatmentsrequired to construct TBMs will affect final microstucture andpossibly impact in-service mechanical behavior

•The planned US/Japan 15J/16J HFIR irradiation experimentprovides a good approximation to TBM irradiation conditions(300C,400C, 2.5-5dpa)

•Possibility of including advanced versions of the FM steels andspecimens derived from prototypic fabrication cycles areunder discussion with JAERI

FY05 R&D activities• ITER materials R&D

–High strength Cu alloy assessment and neutron irradiation R&D–Complete critical assessment of ITER materials properties

handbook data–Support US PFC activities–Support ITER US plasma diagnostics activities–Initiate plasma heating materials R&D (loss tangent of ceramic

insulators)–Participate in plasma diagnostics ITPC workshops (watching

brief on radiation effects and experimental artifacts)• ITER TBM materials R&D

–Structural materials (ferritic/martensitic steels)–Functional materials (SiC flow insert, barrier coatings)

• US/Japan (Jupiter-II and DOE-JAERI) collaboration• Cross-cutting theory and modeling

Dose at start of loading: 8.6 x 10-4

dpa


dpa


dpa


dpa

CuC

rZr

OFH

C C

uOptical Microstructure of neutron-irradiated Cufollowing in-situ straining

Specimen fracture occurs without ductile necking(irradiation assisted stress corrosion cracking?)

EU-US collaborationB.N. Singh, D.J. Edwards, S. Tähtinen, P. Moilanen, P. Jacquet and J.Dekeyser, Risø-R-1481(EN), to be submitted, Nov. 2004

Overview of Radiation Effects Experiments

Dec. 06Jul. 057 dpa600, 900, 1100˚C

SiC andSiC/SiC

RB18J“

Apr. 06Apr. 045 dpa425, 600, 700˚C

V alloys &MHD insul.

RB17J (Eushield)

Jupiter-II

Jan. 08Apr. 066 dpa300, 400, 500˚C

Ferritic steelRB16J“

Jan. 08Apr. 066 dpa300, 400, 500˚C

Ferritic steelRB15J (Eushield)

“

Mar. 09Oct. 0450 dpa300, 400, 500˚C

Ferritic steelJP29 (HFIRtarget)

“

Mar. 09Oct. 0450 dpa300, 400, 500˚C


“

Sept. 06Apr. 0423 dpa300, 400, 500˚C


“

Nov. 04Dec. 039 dpa300, 400, 500˚C


JAERI

Irrad.finish

Irrad.start

Irrad. conditionsMaterialsIrrad.Capsule

Program

Low Temperature Radiation Hardening is Important inFusion RAFMs up to ~400˚C

0

200

400

600

800

1000

1200

0 0.05 0.1 0.15 0.2 0.25 0.3 0.35

USJF82Hss2

En

gin

ee

rin

g S

tre

ss,

MP

a

Engineering Strain, mm/mm

Representative USDOE/JAERI F82H Data:200-600°C, 3-34 dpa

Unirradiated YS

200°C/10 dpa

250°C/3 dpa

300°C/8 dpa

400°C/10 dpa

400°C/34 dpa

500°C/8 dpa

500°C/34 dpa

600°C/8 dpa

Tirr

=Ttest

J.P. Robertson et al.,(DOE/JAERI collaboration)

Engineering stress-strain curves of 9Cr-2WVTa ferritic/martensitic steel after irradiation in spallation environments

Low uniform elongation occurs after irradiation to 0.6 dpa at low temperatures

(a)

(b)

0

200

400

600

800

1000

0 5 10 15 20 25 30

F82H IEA Std.

En

gin

eeri

ng

Str

ess

(MP

a)

Engineering Strain (%)

Irrad. at 573KTested at RT

Irrad. 773KTested at RT

Irrad. 773KTested at 773K

Irrad. at 573KTested at 573K

0

200

400

600

800

1000

0 5 10 15 20 25 30

F82H IEA TIG

En

gin

eeri

ng

Str

ess

(MP

a)

Engineering Strain (%)





Fig. 1 Stress-strain curves of F82H BM (a) and TIG (b)

irradiated at 573K and 773K in tests at RT

F82H BM

F82H TIG

Deformation microstructures in neutron-irradiatedFe-8Cr-2WVTa ferritic/martensitic steel (F82H)

Slip plane: (110) and (011)Slip direction: [111] and [111]

Dislocation channels

Deformation band

N. Hashimoto et al., Fus.Sci.Tech. 44 (2003)

B ≈ [111]g = 110

110

500nm100nm

110

Irradiated weld metal (lower radiation hardening) did notexhibit dislocation channeling after deformation

5 dpa

5 dpa

F82H base metal

F82H TIG weld

Dislocation channel interactions in Fe deformed followingneutron irradiation at 70˚C to 0.8 dpa

g.b.

Need well-engineered materials tomitigate neutron radiation effects

Clearedslipchannel

Master Curve Shifts (ΔTo) and He Effects• Modeled irradiation hardening (Δσy)

induced ΔTo ≈ 0.6°C/MPa• Peak hardening up to ≈ 600 MPa =>

large ΔTo => To ≥ 250°C.• Spallation proton data suggests at > 600-

800 appm He weakens grain boundariesproducing very brittle intergranularfacture that interacts synergistically withΔσy.

• Estimates of combined effects suggestTo > 500°C possible - clearly a showstopper

• High concentrations of H may also bedamaging

0

100

200

300

400

KJc

(M

Pa!

m)

-200 -100 0 100 200

T (°C)

F82H - SINEXT

Unirr. MC

Model

Unirr. - Cor.

Unirr.

Irr. MC

Irr. - 5 mm

Irr. - 10 mm

235°C

0

100

200

300

!T

o (

°C)

0 100 200 300 400

!"y (MPa)

!T/!"y = 0.57 (°C/MPa)

F82H !"y est.

T91

F82H

RPV

b.

Model

≈ 0.6°C/MPaRPVmodel

lowerstrain

hardening

model

0

100

200

300

KJc

(T)

(MP

a!m

)-250 0 250 500 750

T (°C)

!"y He/IGF

operatingrange

J. Henry et al., JNM 318 p.215J. Henry et al., JNM 318 p.249 G.R. Odette, UCSB

Overview of Improved Steels• Steels can exhibit a wide range of properties depending on detailed

composition and thermomechanical treatment• Four generations of ferritic steels based on materials science

principles have been commercialized (R. Viswanathan, Adv. Mater.& Processes 162, No. 8 (2004) 73)– 1st generation 1960-1970; 2nd generation 1970-1985; 3rd generation 1985-

1995; 4th generation currently emerging

• Fusion 9Cr ferritic/martensitic steels are based on “2nd Generation”steels developed around 1985; fusion steels are comparable to 3rdgeneration commercial steels– Fusion substitution of W for Mo (reduced activation) was also pursued for “3rd

generation” commercial steels

• Future steel development options will likely be based on evolutionary(ingot metallurgy/ classical precipitation) and revolutionary(nanoscale oxide dispersion strengthening) approaches

Ferritic/martensitic Steels with Reduced Radioactivity andSuperior Properties Compared to Commercial Steels havebeen Developed by Fusion

Developmentalreducedactivation steels

IEA fusionreducedactivationsteel

Commercialferritic steel(HT9)

Fusion-developed steels also have superiortensile strength, irradiated fracturetoughness, and thermal conductivity

Comparison of thermalcreep-rupture strengths

10-7

10-6

10-5

10-4

10-3

10-2

10-1

100

1 10 100 1000 104

Comparison of Fission and Fusion Radioactivity after Shutdown

Cu

rie

s/W

att

(T

he

rma

l P

ow

er)

Years After Shutdown

Fission: Light Water Reactor

Fusion: Conventional Ferritic steel

Fusion: Reduced Activation

Ferritic Steel

Coal AshBelow Regulatory Concern

Modified Thermomechanical Treatment Procedure forNew 9Cr Ferritic/Martensitic Steel Produced HighStrength

• Strength and ductility intensile test are comparableto high-strengthexperimental ODS steel

R.L. Klueh, to be published

Low-Cr Bainitic Steels Originally Developed by Fusion asReplacement for Commercial 2 1/4 Cr-1Mo Steel

• Low-chromium steels have advantages– Better weldability for easier plant fabrication—an important consideration

• May not require post-weld heat treatment– Cheaper because of less chromium

• Previous studies at ORNL discovered effect of heat treatment ontype of bainite formed

• Mechanical properties determined by type of bainite formed• Discovery was used to develop 3Cr steels

– 50% increase in tensile strength compared to conventional 2 1/4 Cr-1Mo steel– Improved fracture toughness behavior

0

200

400

600

800

1000

0 100 200 300 400 500 600 700

Comparison of the Yield Tensile Strength

of New 3 Cr Steel and 2 1/4 Cr Steel

0.2

% Y

ield

Str

en

gth

(M

Pa

)

Temperature (˚C)

New alloy

Current alloy

Creep Properties of 3 Cr Steels haveAdvantages Over 2 1/4Cr and 12Cr Steels

• Creep resistance improved overJapanese 2.25Cr (T23,T24) steels

•Properties better than for HT9 andas good or better than for modified9Cr-1Mo steel

•Long-term creep behavior (>5000 h)still needs to be determined

•ASME code case qualification beingpursued by industry on two 50 tonheats of fusion-developed 3 Cr steel

Creep Properties of 3 Cr Steels haveAdvantages Over 2 1/4Cr and 12Cr Steels

• Creep resistance improved overJapanese 2.25Cr (T23,T24) steels

•Properties better than for HT9 andas good or better than for modified9Cr-1Mo steel

•Long-term creep behavior (>5000h) still needs to be determined

•Two 50 ton heats have beenfabricated to initiate ASME codeapproval

Oxide dispersion strengthened Steels• There are two main options for ODS steels

– Ferritic ODS steel (typically 12-16%Cr)– Ferritic/martensitic ODS steel (typically ~9%Cr)

Limited to temperature below~700 CMarginal oxidation resistance athigh temperatures

Nearly isotropic propertiesafter heat treatmentBetter fracture toughness

9% ODSferritic/martensiticsteel

Anisotropic mechanicalpropertiesLower fracture toughness

Higher temperature capabilityBetter oxidation resistance

12-16% ODSferritic steel

DisadvantagesAdvantagesSteel

0

200

400

600

800

1000

1200

400 600 800 1000 1200

Temperature (K)

12YWT

MA956

ODS (SUMITOMO)

9Cr-2WVTa

ODS(ZrO2)

ODS(TiO2)

ODS(MgO)ODS(Al

2O

3)

Yie

ld S

tress

(M

Pa)

New 12YWT Nanocomposited Ferritic Steel hasSuperior Strength compared to conventional ODS steels

O Y Ti10 nm

• Atom Probe reveals nanoscale clustersto be source of superior strength

– Enriched in O(24 at%), Ti(20%), Y (9%)– Size : rg = 2.0 ± 0.8 nm– Number Density : nv = 1.4 x 1024/m3

• Original Y2O3 particles convert tothermally stable nanoscale (Ti,Y,Cr,O)particles during processing

• Nanoclusters not present in ODS Fe-13Cr + 0.25Y2O3 alloy

• Thermal creep time to failure is increased byseveral orders of magnitude at 800˚Ccompared to ferritic/martensitic steels

• Potential for increasing the upper operatingtemperature of iron based alloys by ~200°C

• Acceptable fracture toughness near roomtemperature

Nanocomposited ODS Steel Summary• Fundamental understanding of the formation of nanoclusters (NCs)

and uniform grain sizes in Fe-14Cr ferritic alloys has been achieved– Advanced characterization tools utilized to study effects of composition,

processing conditions at key steps in fabrication process• Small Angle Neutron Scattering (SANS)• Energy Filtered Transmission Electron Microscopy (EFTEM)• Local Electrode Atom Probe (LEAP) field ion microscope

• The thermal stability of nanoclusters in several ODS steels has beenstudied– Results suggest remarkable nanocluster stability (e.g., particles are resistant to

coarsening at 1000ºC for up to t > 105 h)• Preliminary results showed essentially no changes in the size or

structure of the nanoclusters in 12YWT ODS steel following 6.8MeVFe+3 irradiation to 20 dpa at 700ºC

• PIE of ODS ferritic alloys in HFIR JP26 (5-8 dpa, 300-500˚C) will bestarting– Variety of ODS ferritic alloys (MA957, PM2000 and 12YWT, etc.)– Microstructural effects of fusion-relevant He injection during irradiation

Local Electrode AP Revealed High a Number Densityof Nanoclusters in the As-Extruded 14YWT Alloy

• The nanoclusters are consistent with Y-, Ti- and O-enricments• LEAP also shows C associated with the NC’s

• A possible grain boundary shown in the atom map is enriched with Y,Ti, and O atoms

• Segregation of Cr and W to the grain boundary was also observed• Annealing the as-extruded alloy at 1000 º C for 1 h showed no

changes in the NC’s

SiC/Pb-Li Compatibility Study• Good compatibility was observed between monolithic SiC and Pb-17Li at 800° and

1100°C in static capsule tests up to 1000h.• No Si detected in chemical composition measurement of the Pb-Li after 800 and

1100 °C capsule test– (<5 ppm detection limit)

• Additional testing at 800-1200˚C up to 2000 h is in progress

Irradiation Effect (1)Thermal / Electrical Conductivity

• Irradiated thermal conductivity of CVD SiCat 200 – 800 C has extensively beenstudied.

101

102

103

104

0 100 200 300 400 500 600

Ele

ctr

ical R

esis

tivit

y (

oh

m-c

m)

Isochronal Anneal Temp. (°C)

1.1 dpa irradiated

non-irradiated

IrradiationTemperature

• Irradiated electrical conductivity of CVD SiC (andits annealing) is routinely measured for passivethermometry.

• Measurements are mostly limited to roomtemperature.

Physical Properties Below Crystallization Temperature- Amorphized Morton CVD SiC -

102

103

104

105

106

0.2 0.22 0.24 0.26 0.28 0.3 0.32 0.34

Res

isti

vit

y (

oh

m-c

m)

1/T^0.25

Amorphous

NonIrradiated

0°C -100°C -200°C-150°C200°C

Low-T res s-10

0 200 400 600 800 1000

Anneal Temperature (°C)

885°CDisintegrated

Anneal Figure ResistivityTirr

• Temperature dependence follows T-1/4 dependence indicating hopping conduction.• Reducing the density of states increases the conductivity up to the point of crystallization• Limited data obtained during irradiation (radiation induced conductivity) also available

10-10

10-9

10-8

10-7

10-6

10-5

10-4

10-3

10-2

10-1

100

101

10-2 10-1 100 101

Elec

trica

l Con

ducti

vity

(S/m

)

Dose Rate (Gy/s)

x x x x x x x

Kyocera ! SiC, 22°C

Kyocera ! SiC, 200°C

Hexoloy SiC, 22°C

CVD SiC, 22°C

Hexoloy non-irradiated

+ +Nicalon

Nicalon non-irradiated

+++++σ=σo+KRδ

Radiation Induced Conductivity for different grades of SiC:Electrical Conductivity varies by 8 orders of magnitude!

• Ionizing radiationexcites electrons intothe conduction bandenhancing conductivity.

• If unirradiatedconductivity is high, RICis insignificant

Doserate

L.L. Snead, J. Nucl. Mater.329-333 (2004) 524

All values measured to date are <2 S/m

HFIR-18J will examine transport properties of SiC andSiC/SiC composites for FCI applications

(Planned)~103Tyranno™-SA Grade-3

PlannedPlannedCVI Composites, Type-S orSA3, varied PyC interphases

(Planned)~103Hi-Nicalon™ Type-S

Planned0.05<0.005CVD SiC Low Conductivity

PlannedPlanned10~1CVD SiC Standard

PIE, ThermalConductivity

PIE, ElectricalConductivity

ElectricalConductivity

(S/m)

NitrogenConcentration

(ppm)

• The 18J experiment enables a reliable prediction of the irradiated electrical andthermal conductivities for various SiC-based materials by the constitutivemodeling approach.

US/Japan Jupiter-II collaboration

10

100

0 100 200 300 400 500 600 700 800

Th

erm

al C

on

du

ctiv

ity

(W

/m-K

)

Measurement Temperature (°C)

Rohm Haas CVD SiCORNL Data

Non-irradiated

4.5 dpa/800°C

4.0 dpa/500*C

4.6 dpa/300°C

0.001

0.01

0.1

1

0.0001 0.001 0.01 0.1 1 10

dpa

Th

erm

al D

efec

t R

esis

tan

ce (

W/m

-K)-1

ORNL Data

200°CComposite

200°CCVD SiC

800°CComposite

800°CCVD SiC

!

K (T )[ ]"1

=1

Ku(T )+

1

Kgb(T )+1

Kd 0

+1

Krd

#

$ %

&

' (

umklapp boundaries intrinsicdefects

radiationdefects

Thermal Defect Resistance Model Enables ReliablePrediction of Irradiated Thermal Conductivity ofComposites

Thermal Conductivity of 2D SiC/SiC Composites

0

20

40

60

80

100

0 200 400 600 800 1000 1200

Th

erm

al

co

nd

uc

tiv

ity

(W

/mK

)

Temperature (°C)

2D-SA Tyrannohex

(unirradiated)

(50% parallel fibersand 50% interfaces)

(100% transverse fibersand 100% interfaces)

0

10

20

30

40

50

0 200 400 600 800 1000 1200

Temperature (oC )

Th

erm

al

Co

nd

uc

tiv

ity

(W

/mK

) fiber: Tyranno SAmatrix: CVI-SiC

Vf = 0.4

(enhanced f/m bonding)

(degraded interface)

(typical interface)

5000

500

50

Interface

conductance(W/cm

2K)

0

10

20

30

40

50

400 600 800 1000 1200

Temperature, K

Th

erm

al

Co

nd

uc

tiv

ity

, W

/mK fiber: Tyranno SA

matrix: CVI-SiCV

f = 0.4, t = 200 nm

(anisotropic PyC)

(isotropic PyC)

(degraded PyC)

20 W/mK

0.2 W/mK

2.0 W/mK

Coating

Conductivity

• Number and quality of interfacesimportant

• Fibers parallel to conduction pathuseful (3D architecture)

• Samples currently being irradiated

Experimental:

NERI Program

• Strong bonding due to fibersurface treatment useful

• Degraded interface due toirradiation or fiber/matrixmismatch

Model 1: Thin Interface Conductance• Preferred orientation of

graphitic crystallites• Strong interface cohesion• Degraded interphase due to

irradiation or oxidation

Model 2: Interphase Conductivity

Fusion Materials Program

U.S. Department of EnergyPacific Northwest National Laboratory

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