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Implications of Available Data on Helium Effects in Low-Activation

Ferritic/Martensitic Steels

Roger E. Stoller* and Rick J. Kurtz**many et al.s

*Oak Ridge National Laboratory **Pacific Northwest National Laboratory

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Objective• provide a little background on helium effects• comment on helium effects data relevant to a DT-burning

FNSF, i. e. exposures up to ~10 dpa with ~150 appm He• no data exists in prototypical conditions, available data

from environments that generate “similar” irradiation conditions– fission reactor with tricks employed to get He (neutron energy

spectrum modification, alloy composition, isotopic variation, Heinjector layers

– spallation neutron irradiation– charged particle irradiation with single, dual, and triple beams

• modeling and simulation employed to guide interpretation and implications of data

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Early concerns about helium focused on swelling: mechanistic models used to anticipate impact of He

• Helium effects investigated– experiments with isotopic tailoring of

Ni in austentitic steel used to obtain intermediate and high helium in water moderated reactors (ORR and HFIR)

– experiments confirmed model predictions of possible non-monotonic behavior

Some success, but confidence limited by nature of models and parameters, e.g.

• basic material and defect parameters, point defects and helium

• nucleation model• …

Subsequent progress addressed primary damage formation, fundamental behavior of He, cavity nucleation, He transport

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Swelling can eliminate materials from consideration

0

2

4

6

8

10

12

14

0 50 100 150 200

Volu

met

ric S

wel

ling

(%)

Damage Level (dpa)

Ferritic steel

Ti-modified 316 stainless steel

316 stainless steel

Tirr=400-500ÞC°

F82H (36 appm He) 10B-doped F82H (330 appm He)

Current fusion candidate alloys not immune to He effects:HFIR irradiation at 400˚C to 51 dpa, E. Wakai, et al. JNM (2000)

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“Simulations” Can Get Close, But Each Approach Has Limitations

• energy spectrum of irradiating particle

– pka spectrum– too much or too little He (and H)– other different transmutation

products

• displacement rate• limited irradiated volume• limited irradiation volume

IFMIF

Spallationneutrons

Fusionreactor

ITER

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Relevant results on energy spectrum: MD simulations

10 keV

100 keV5 keV

• energy dependence of defect survival minimizes spectrum effect

• subcascade formation plays prominent role

• reduced parameter variability in mesoscale models

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Latest trick: in situ He Injection

Ni (B, Li. Cu) layer

substrate

nthnth

α

α

58Ni + nth59Ni + γ

59Ni + nth56Fe + 4He (4.76MeV)

• Use (n,α) reactions (Ni, B, Li ..) in mixed spectrum reactors to produce controlled He/dpa for fusion relevant conditions.

• Avoids confounding properties and irradiation effects in doping.

• Applicable to any material - e.g., SiC and a variety of specimens.

• Obtain microstructural information and limited mechanical properties -low-load hardness

Odette, et al., UCSB

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• Yield strength and strain hardening constitutive laws.• Various types of ‘ductility’.• Fatigue crack growth rates.• Fracture toughness.• Irradiation and thermal creep rate.• Thermo-mechanical fatigue limits .• Creep-fatigue interactions.• Environmentally assisted cracking.• Bulk corrosion, oxidation & compatibility.• Void swelling rates.• Creep rupture times and strains.• Creep crack growth rates.• Flaw distributions.

Properties needed for design decisions

Material response involves synergistic interactions among many variables

Unirradiated

Irradiated365 C, 7.4 dpao

Fe-9Cr

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Transmutation products and atomic defects lead to accelerated non-equilibrium mesoscale evolution and damage accumulation over long time

• Voids, bubbles, dislocations and phase instabilities (damage).

• Dimensional instabilities (swelling and irradiation-thermal creep).

• Complete loss of strain hardening capability.• High-low temperature He embrittlement.• Fatigue, creep-fatigue, crack growth.• Corrosion, oxidation and impurity

embrittlement (W, V).

Exposure conditions, neutron irradiation and mechanical loads, limit design choices

Materials Design Window

He embrittlement,Thermal creep,Corrosion

Temperature

Life

time Dimensional

InstabilityHardening, Fracture

N. Ghoniem & B.D. Wirth, 2002

High He may narrow or even close design window

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Impact of He-Rich Environment on Neutron Irradiated Materials

A unique aspect of the DT fusion environment is theproduction of high levels of He and H by transmutationAccumulation of He can have a significant impact on the material design window:

- Potential for loss of ductility and fracture toughness at low temperatures.

- Increased swelling and irradiation creep at intermediate temperatures.

- Loss of high-temperature creep strength.

Grain boundary

1

10

100

1000

Un-implanted 200 appm HeTi

me t

o ru

ptur

e, h

316 SS @ 750°C & 100 MPa

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Impact of He at “low” irradiation temperaturesHe (and H?) has significant potential to exacerbate displacement damage and cause loss of structural integrity:

- High-temperature creep embrittlement.- Intermediate-temperature swelling.- Low-temperature loss of fracture toughness.

Data Y. Dai

0

100

200

KJc

(MPa

¦m)

-200 -100 0 100 200T (°C)

F82H 5dpa @ 300°C

235°CDuctile

Brittle

Potentially very large DBTT shifts with increasing levels of He

Brittle intergranular fracture

High He

Low He

0

200

400

600

800

1000

0 10 20 30 40 50

F82H

Eurofer 97

T91

optimax

optifer

F82H(SP)

T91(SP)

optimax(SP)

F82H, Eurofer (n-irr)

ΔT c (º

C)

dpa

neutron only

100 (appmHe/dpa)

100 (appmHe/dpa) C

c Š 1.1

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Data generated within DOE-JAEA provide dose dependence of transition temperature shift at various irradiation

temperatures for F82H-IEA

F82H-IEA

0

50

100

150

200

250

0 5 10 15 20 25Dose, dpa

Tran

sitio

n Te

mpe

ratu

re S

hift,

o C

250oC

300oC

400oC

500oC

open symbols are from Lucon, SCK-CEN

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Examples of data from:

The First International Workshop on Measuring, Modeling and Managing

Helium-DPA EffectsPaul Scherrer Institut (PSI), Switzerland

15-17 June 2009

Not to be distributed without permission

Cavity microstructure formed by Dual/Triple ion irradiation 

50 nm

Triple-ion irradiation100 appmHe/dpa1000 appmH/dpa

Triple-ion irradiation10 appmHe/dpa40 appmH/dpa

Dual-ion irradiation10 appmHe/dpa0 appmH/dpa

470℃

510℃

600℃

F82H

/ 50

dpa

F82H / 50dpa

Swelling is enhanced by additional H implantation 

0.01

0.1

1

10

460 480 500 520 540 560 580 600

Swel

ling

in F

82H

(%)

Irradiation Temperature (oC)

Fusion(dual)

Fusion(triple)

Spallation(triple)

Dual (damage+He)Triple (Fusion)Triple (Spallation)A

B

Displacement damage [ dpa ]

Void

Sw

ellin

g [ %

]

0.028

%/dp

a0 20 40 60 80 1000

0.5

1.0

1.5

2.0

A larger swelling rate than neutron irradiation?Helium appears accelerate onset of swellingFlux effect should be taken into account.

FFTF/MOTA420°C

DuET470°C

Dose Dependence of Swelling (Dual)

K3-ODS JLF-1

Dua

l Si

ngle

50nm

Microstructure Observation (1)60 dpa, (900appmHe), 773K

0

1

2

3

4

5

6

7

40 60 80 100 120 140 160

JLF-1

K3-ODS

Swel

ling

(%)

Displacement damage (dpa)

Dual-ion irradiated at 773 K15appm He/dpa, 1.0x103dpa/sec

0

2

4

6

8

10

0

1 1023

2 1023

3 1023

4 1023

5 1023

40 60 80 100 120 140 160

K3-diameterJLF-1-diameter

K3-densityJLF-1-densityM

ean

Cav

ity D

iam

eter

(x10

-9m

)

Cavity N

umber D

ensity (m-3)

Diaplacement damage (dpa)

Dual-ion irradiated at 773 K15appm He/dpa, 1.0x10-3dpa/sec

Microstructure Observation (2)60 dpa, (900appmHe), 773K

Dose dependence of the void swelling of JLF-1 reflects that of void size.No bubble growth was observed for the ODS steel.

P.Jung9PSI Helium WorkshopJune 15-17, 2009

3.3/ Tensile results (overview)

[%]0 2 4 6

[M

pa]

0

200

400

600

800

1000

1200

1400cHe[atppm]

2400600

Timp=Ttest=250°C

5000

0

9Cr1MoVNb(T91)strain rate: 8.5 10-5

-4

-3

-2

-1

0

1

2

3

0.025 0.05 0.075 0.1 0.125 0.15 0.175

q (A-1)

ln(I)

100

µm ductilenecking

SANS

250°C

550°C

7 nm

2 nm

100

µm

brittle bubbles

„black dots“

500°C

He bubbles

TEM

SEM

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Peter Jung, ForschungsZentrum Jülich• High-energy cyclotron He

implantation experiments• ~100 μm implanted region• High He/dpa ratio, >1000 appm/dpa

P.Jung5PSI Helium WorkshopJune 15-17, 2009

2.1/ High temperature He embrittlement

[MPa]200 300 400 500

rupt

ure

stra

in [%

]

0,1

1

10 1.4970

316L

1.4914(MANET)

in-beam600°C

102-103atppm He(dashed=reference)

[MPa]200 300 400

rupt

ure

time

[hr]

0.1

1

10

100

1000

1.4970

316L

1.4914(MANET)

in-beam600°C

102-103atppm He(dashed=reference)

creep rupture during He-implantation under stress

HTHE: Reduction of rupture strain and rupture time in austenitic stainless steels.Intergranular brittle fracture, by He bubbles at grain boundaries• lower HTHE in Ti-stabilized steels, trapping of He at coherent precipitates• negligible HTHE in martensitic steels, probably due to soft matrix

H.Schroeder, P.Batfalsky et.al.

P.Jung6PSI Helium WorkshopJune 15-17, 2009

2.2/ HTHE: Fatigue

max.strain rate [1/s]10-3 10-2 10-1

cycl

es to

failu

re

103

104

105

frequency [1/s]0,01 0,1 1 10

600°C

unimplanted

800atppm He

500°C

316L

Ttest=Timpl

A.Leeser, 1988

max.strain rate [1/s]10-5 10-4 10-3

cycl

es to

failu

re

102

103

frequency [1/s]0,0001 0,001 0,01 0,1

5. 10

24 n/m

2

unirradiated

1.4948550°C=1%

M.I.deVries, 1982

at low frequencies:„breathing“ of He-bubbles

reactor irradiationeffect of implanted He

bubbles at grain-boundaries,predominantly vertical to

Experimental Conditions - E. Wakai, JAEA

・Specimens:： 316FR, 316FR+10B (10ppm , 23 ppm)

・Irradiation ： 550CIrradiation ： 550CJRR-3M (0.15 dpa, 2-30 appmHe) for 1200 h JOYO (1.5 dpa, 1 appm) for 800 hJRR-3M + JOYO for 2,000 h JRR‐3M JOYOJ J ,

・Aging:： 550C for1,000h and 2,000h)

・Mechanical Tests： Tensile in VacuumCreep in Vacuum

・He Implantation by Cyclotron (50 MeV-He2+ ions) at 550C1 10, 30, 50 appmHe1.2

]

0.350MeV-He → 316FR

0.4

0.6

0.8

1

ncen

trat

ion

[app

m]

0.1

0.2

amag

e [m

dpa]

0

0.2

0 100 200 300 400

Depth from surface [µm]

He

con

0

D

7( t =0.35 mm )

Creep Properties of 316FR (Present data)

1.2n)

JRR-3材 常陽材 常陽 ->JRR-3材 サイクロトロン材JOYO JOYO JRR‐3M CyclotronJRR‐3M

0.8

1m

e ad

iatio

nLower dpa

0.4

0.6

p Li

fe T

in/

Uni

rra

Higher dpa

0

0.2

Cre

epad

iatio

n Higher dpa

0.1 1.0 10.0 100.0 1000.0

(Irr

a

appm-He/dpa

He/dpa:Th i f lif i f i di i h f i di i

20

The ratio of life time of irradiation case to that of unirradiation case tended to decrease with He/dpa and displacement damage.

Tensile properties and microstructures of 9Cr tempered martensitic steels and ODS-14%Cr ferritic alloy irradiated with high energy protons and neutrons

Tensile

Materials :

- T91 (9Cr1Mo V Nb), N (1 h @1040°C &T (1h @760°C) +

CW 20% + T (1h @760°C); « S » type tensile specimens

- EM10 (9Cr1Mo), N (30 min @950°C) & T (30min @750°C) ;

« S » and « L » tensile specimens

- MA957 ODS (Fe-14Cr-1Ti-0.3 Mo-0.25 Y203), hot

extruded at 1100°C & 25% cold worked, «« SS »» type type

tensiletensile specimensspecimens

Dose up to ~20 dpa (~ 0.17 % He),

(average)T range [115-360°C]

Tests performed at RT, 250°C – 3 10-4/s

Materials, irradiation and testing conditions

S L

« STIP II »irradiation

Tensile properties and microstructures of 9Cr tempered martensitic steels and ODS-14%Cr ferritic alloy irradiated with high energy protons and neutrons

0

200

400

600

800

1000

1200

1400

1600

0 2 4 6 8 10 12 14 16 18 20

strain (%)

stre

ss (M

Pa)

unirradiated7.2 dpa115°C

He : 565

10.7 dpa160°C

He : 850

15.2 dpa230°C

He : 1305

19.6 dpa295°C

He : 174019.9 dpa290°C

He : 1825

Examples of tensile curves (T91 tested at RT)19.6 dpa

Tensile properties and microstructures of 9Cr tempered martensitic steels and ODS-14%Cr ferritic alloy irradiated with high energy protons and neutrons

Evolution of Tensile properties/dpa (T91, 250-350°C)

0

200

400

600

800

1000

1200

1400

0 5 10 15 20 25dose (dpa)

stre

ss (M

Pa)

0

4

8

12

16

20

24

28

%

YSUTSUETE110°C

He : 500

120°C He : 565

150°C He : 795 175°C

He : 910

245°C He : 1205

290°C He : 1370 325°C

He : 1645

365°C He : 1750tested at 350°C

Tensile properties and microstructures of 9Cr tempered martensitic steels and ODS-14%Cr ferritic alloy irradiated with high energy protons and neutrons

Examples of tensile curves (MA957 tested at 250°C)

0

200

400

600

800

1000

1200

1400

0 2 4 6 8 10 12 14 16 18 20

strain (%)

stre

ss (M

Pa)

unirradiated

7 dpa120°CHe:535

11 dpa195°C

He: 850

15 dpa280°C

He: 1305

Tensile properties and microstructures of 9Cr tempered martensitic steels and ODS-14%Cr ferritic alloy irradiated with high energy protons and neutrons

Evolution of Tensile properties/dpa (MA957, 250°C)

0

200

400

600

800

1000

1200

1400

0 5 10 15 20 25dose (dpa)

stre

ss (M

Pa)

0

4

8

12

16

20

24

28

32

36

%

YSUTSUETE

120°CHe: 535

190°CHe: 850

250°CHe: 1200

280°CHe: 1300

300°CHe: 1740

PAUL SCHERRER INSTITUT

Paul Scherrer Institut • 5232 Villigen PSI• Switzerland, Yong DAI 1st He-dpa Workshop, PSI, 15-17.06.2009

Neutron irradiation data: (1) Klueh et al, J. Nucl. Mater. 218 (1995) 151; (2) Hu et al. STP 1046 (ASTM, 1990), p.453; (3) Alamo, Euromat2007.

0 2 4 6 8 10 12 14 16 18 200

50

100

150

200

250

300

DBTTSP = DBTTCVN

DB

TTS

P (°

C)

SP

D

BTT

CV

N (°

C)

Displacement (dpa)

Neutron irradiation0

100

200

300

400

500

600

700F82H T91 Optimax-A/C Optifer-V/-IX Eurofer97

Ti<380°CCVN

27-42 dpaT91 & Eurofer97

FM steels FM steels –– DBTT shiftDBTT shift

PAUL SCHERRER INSTITUT

Paul Scherrer Institut • 5232 Villigen PSI• Switzerland, Yong DAI 1st He-dpa Workshop, PSI, 15-17.06.2009

FM steelsFM steels

0 200 400 600 800 1000 1200 1400 16000

50

100

150

200

250

300

DBTTSP = DBTTCVN

D

BTT

(°C

)

SP

D

BTT C

VN (°

C)

Helium concentration (appm)

0

100

200

300

400

500

600

700F82H T91 Optimax-A/C Optifer-V/-IX Eurofer97

Ti<380°CCVN

k = 0.48

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Materials Design Strategy to Manage High He and Displacement Damage

• Trapping at a high-density of nanometer-scale microstructural features is key strategy for management of He.

NFA

BubblesDislocations

Voids

GB creep cavity

Loops&voids

GB bubbles

TMS

NF

Climb-glide

IG fracture

Ductile Fracturea b

G. R. Odette, M. J. Alinger and B. D. Wirth, “Recent Developments in Irradiation Resistant Steels”, Annual Reviews of Materials Research

V38 (2008) 371-403F82HMA957

NFA Ductile fracture

J. Henry ICFRM13 STIP 320°C ~19 dpa

1700 appm He

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Fracture toughness of ODS steels Comparison of F82H, 12YWT, and 14YWT

TEMPERATURE, oC

-200 -100 0 100 200 300 400 5001-T

FRAC

TUR

E TO

UG

HN

ESS,

MPa

m1/

2

0

100

200

300

400

12YWTT0=102oC

K from JIc

T0=-105oCF82H

K from JIcTo=<-150oC14YWT

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No DBTT shift was observed after irradiation at 300oC to ~ 1.5 dpa

-250 -200 -150 -100 -50 0 50 100 150 200 250 30025

50

75

100

125

150

175

200

14YWT [KJc(1T)] (unirradiated) 14YWT [KJIc] (unirradiated)

14YWT [KJc(1T)] (Tirr= 300oC)

14YWT [KJIc] (Tirr= 300oC)Frac

ture

Tou

ghne

ss [M

Pa√m

]

Temperature [°C]

Irradiated 14YWT Fracture Toughness

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Requirements for neutron sources for fusion reactor materials science

Overcoming radiation damage degradation is the key rate-controlling step in fusion materials development.• Additional factors such as joining are important, but critical

radiation effects data is needed to evaluate feasibility.Evaluation of radiation effects requires simultaneous displacement damage (~200 dpa) and He generation (~2000 appm He).• Data without high fusion relevant dpa and He/dpa of limited value.

Evaluation of mechanical properties for a given material at a given temperature requires a minimum volume of ~10 cm3 with flux gradients < 20%/cm.• Innovative small-volume neutron sources would be useful but do

not replace the need for a moderate-volume intense neutron source.

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Summary• Growing evidence that He effects may have a limit

design/application window for conventional FM steels– considerable uncertainty due to lack of prototypic

irradiation environment• Best alternative is development of ODS variants with

thermally stable, nanometer-sized oxide dispersions– high creep strength– radiation tolerance– acceptable fracture toughness