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OAK RIDGE NATIONAL LABORATORYU. S. DEPARTMENT OF ENERGY
Mechanical Property Changes inMetals due to Irradiation
Steven J. Zinkle
Materials Science & Technology Division,Oak Ridge National Laboratory
Materials for Generation IV Nuclear ReactorsNATO Advanced Study Institute
Cargese, Corsica, FranceSept. 24-Oct. 6, 2007
OAK RIDGE NATIONAL LABORATORYU. S. DEPARTMENT OF ENERGY
Outline•Effect of low temperature (<0.3 TM) irradiation on the tensile
properties of metals• Dose and temperature dependence• FCC vs. BCC metal behavior
•Fracture toughness embrittlement in irradiated metals•High temperature He embrittlement of grain boundaries•Overview of deformation mechanisms in irradiated metals
(restricted to radiation hardening/embrittlement regime)• Microscopic flow localization observations (dislocation channeling)• Similarities and differences between flow localization phenomena in
unirradiated and irradiated metals• Practical consequences: e.g., structural design rules for uniform elongation <2%
• Not covered: hardness, fatigue, irradiation creep
OAK RIDGE NATIONAL LABORATORYU. S. DEPARTMENT OF ENERGY
Radiation Hardening Mechanisms
Radiation damage creates obstacles to dislocation motion- point defects ( ≡ solute strengthening in alloys)- small clusters and precipitates ( ≡ precipitates in alloys)- impenetrable clusters ( ≡ Orowan strengthening in alloys)
- all give ΔσY proportional to (dose)0.3-0.5
obstacles
dislocation
σOrowan loop
OAK RIDGE NATIONAL LABORATORYU. S. DEPARTMENT OF ENERGY
Dose Dependence of Radiation Hardening in Copper
• Fission and fusion neutron radiation hardening behavior are in good agreement• Hardening “saturation” occurs for doses above ~0.1 dpa (agrees with defect
cluster density behavior)
50
100
150
200
250
300
350
0.0001 0.001 0.01 0.1 1 10 100
Neutron Radiation Hardening of Copper
Kruglov et al (1969)El-Shanshoury (1972)Mohamed et al (1982)Vandermeulen (1986)Heinisch (1988)Fabritsiev et al (1994)Singh et al (1996)Zinkle&Gibson (1999)Singh et al. 1999
0.2
% Y
ield
Str
en
gth
(M
Pa
)
Damage Level (dpa)
Tirr=30 - 200˚C
50
100
150
200
250
300
350
0.00010.001 0.01 0.1 1 10 100
Neutron Radiation Hardening of Copper
0.2
% Y
ield
Str
en
gth
(M
Pa
)
Damage Level (dpa)
Tirr=30 - 200˚C
14 MeV neutrons: filled symbolsfission neutrons: open symbols
OAK RIDGE NATIONAL LABORATORYU. S. DEPARTMENT OF ENERGY
Radiation Hardening in Copper: Seeger vs.Friedel relationships
• Two general models are available to describe radiationhardening (∆σ) in metals:–Dispersed barrier model (Seeger, 1958)--valid for strong obstacles
Where M=Taylor factorα=defect cluster barrier strengthµ=shear modulusb=Burgers vector of glide dislocationN, d=defect cluster density, diameter
–Friedel 1963 (also Kroupa and Hirsch 1964) weak barrier model:
!
"# = M$µb Nd
!
"# =1
8MµbdN 2/3
OAK RIDGE NATIONAL LABORATORYU. S. DEPARTMENT OF ENERGY
!
"# = M$µb Nd
!
"# =1
8MµbdN 2/3vs.
0.1
1
10
1016 1017 1018 1019 1020 1021
Shear Strength of Cu Single Crystals Irradiated
and Tested Near Room Temperature
Blewitt et al 1960
Shear
Str
ength
(kg
f/m
m2)
Neutron Fluence (n/cm2)
n~0.33
0
2
4
6
8
10
0 0.2 0.4 0.6 0.8 1 1.2
Shear Strength of Cu Single Crystals Irradiated
and Tested Near Room Temperature
Blewitt et al 1960
Shear
Str
ength
(kg
f/m
m2)
(!t)1/2 (1020 n/cm2)1/2
slope change
@3x1018/cm2
(3x10-4 dpa)?
0
2
4
6
8
10
0 0.2 0.4 0.6 0.8 1 1.2
Shear Strength of Cu Single Crystals Irradiated
and Tested Near Room Temperature
Blewitt et al 1960
Shear
Str
ength
(kg
f/m
m2)
(!t)1/4 (1020 n/cm2)1/4
0
2
4
6
8
10
0 1 2 3 4 5
Shear Strength of Cu Single Crystals Irradiated
and Tested Near Room Temperature
Blewitt et al 1960
Shear
Str
ength
(kg
f/m
m2)
(!t)1/3 (1018 n/cm2)1/3
Seeger; N~(φt)n, n=1–>1/2 Friedel; N~(φt)n, n=1/2
OAK RIDGE NATIONAL LABORATORYU. S. DEPARTMENT OF ENERGY
Effect of test temperature on irradiated strength• What is the effect of irradiation on the yield strength test
temperature dependence (athermal and thermal components)?
Flo
w S
tre
ss
Test Temperature
!athermal
!thermal
BCC metals: large τthermal component
FCC metals: small τthermal component
Hardening models (Seeger, Fleisher, etc.):Thermal component of flow stress iscontrolled by radiation-induced hardeningcenters
OAK RIDGE NATIONAL LABORATORYU. S. DEPARTMENT OF ENERGY
Early Work on BCC Metals such as Fe led to Conclusions that theTest Temperature Dependence was not Affected by NeutronIrradiation
• These early studies were limited to low doses <1019 n/cm2), where the magnitude ofradiation hardening (and radiation-induced defect sizes) is relatively small
200
300
400
500
600
700
800
900
1000
0 100 200 300 400 500 600
Effect of Test Temperature on the Yield Strength of
Mild Steel Irradiated with Fission Neutrons at 60 C
Lo
wer
Yie
ld S
tren
gth
(M
Pa)
Test Temperature (K)
Unirrad.
2.3e18 n/cm2
Little (1970)
OAK RIDGE NATIONAL LABORATORYU. S. DEPARTMENT OF ENERGY
Comparison of the Yield Strength Behavior ofAnnealed and (low-dose) Irradiated Iron
10
20
30
40
50
60
70
50 100 150 200 250 300
Effect of Test Temperature on the Yield Strength of
Polycrystalline Iron Irradiated with Fission Neutrons at 60 CL
ow
er
Yie
ld S
tren
gth
(kg
/mm
2)
Test Temperature (K)
Unirrad.1.2e16 n/cm
2
Ohr (1969)
OAK RIDGE NATIONAL LABORATORYU. S. DEPARTMENT OF ENERGY
Comparison of the Yield Strength Behavior ofAnnealed and Irradiated Iron at Higher Doses
0
10
20
30
40
50
60
50 100 150 200 250 300
Effect of Test Temperature on the Yield Strength of
Single Crystal Iron Irradiated with Fission Neutrons at 75 CL
ow
er
Yie
ld S
tren
gth
(kg
/mm
2)
Test Temperature (K)
Unirrad.
1.8x1017
Kitajima (1968)1.3x1018
1x1019
n/cm2 (E>1 MeV)
} 60 MPa
200 MPa
OAK RIDGE NATIONAL LABORATORYU. S. DEPARTMENT OF ENERGY
Comparison of the Yield Strength Behavior ofAnnealed and Irradiated Iron at Higher Doses
0
50
100
150
200
250
300
350
100 150 200 250 300 350 400 450
Lo
we
r Y
ield
Str
en
gth
(M
Pa
)
Test Temperature (K)
unirradiated
irradiated
2.8x1021
/m2
J. Diehl et al. (1968)Single crystal Fe
Tirr
~350 K
!"
=2
10
MP
a
!"=60 MPa
OAK RIDGE NATIONAL LABORATORYU. S. DEPARTMENT OF ENERGY
Radiation Hardening in Eurofer97 Steel irradiated at300oC
400
500
600
700
800
900
1000
1100
0 50 100 150 200 250 300 350
Yie
ld S
tren
gth
(M
Pa)
Test Temperature (oC)
!"=540 MPa
!"=460 MPa
J. Rensman (2005)
9 dpa, Tirr
=300oC
25 mm plate
OAK RIDGE NATIONAL LABORATORYU. S. DEPARTMENT OF ENERGY
Effect of Test Temperature on the Yield Strength ofPolycrystalline Copper, Neutron Irradiated at 60˚C
0
100
200
300
400
500
0 100 200 300 400 500
Yie
ld S
tre
ng
th (
MP
a)
Test Temperature (K)
Unirrad.9.5e16
4.1e17
3.6e18
5.9e18
1.3e19
2.3e20n/cm2
(E>1 MeV)
Makin (1967)
• Change in test temperature dependence from unirradiated behavior occurs for fluencesabove ~0.5 to 1x1018 n/cm2, E>1 MeV (Makin 1967, Koppenaal 1965)
• Similar range of test temperature dependence occurs in as-irradiated vs. postirradiationannealed Cu (Rühle, Makin, Howe, Koppenaal, etc.)
180 MPa
410 MPa
Radiationhardeningcontrolled primarilyby SFTs
OAK RIDGE NATIONAL LABORATORYU. S. DEPARTMENT OF ENERGY
Effect of Electron Irradiation on the YieldStrength Behavior of Single Crystal Copper
1
1.5
2
2.5
3
3.5
4
4.5
0 50 100 150 200 250 300 350
Re
so
lve
d S
he
ar
Str
en
gth
(M
Pa
)
Test Temperature (K)
unirradiated
1.0x1020
e/m2
3.4x1021
e/m2
2 MeV electron irrad. Cu
Tirr
=300 K
K. Kamada et al. (1973)
Radiation hardening controlledprimarily by dislocation loops
OAK RIDGE NATIONAL LABORATORYU. S. DEPARTMENT OF ENERGY
Comparison of the Yield Strength Behavior ofAnnealed and Quenched Aluminum
0
5
10
15
20
25
30
35
40
50 100 150 200 250 300
Yie
ld S
tre
ng
th (
MP
a)
Test Temperature (K)
slow cooled from 600oC
quenched from 607oC
n=4x1020
/m3
Loop d=23 nm
AluminumK.H. Westmacott (1966)
Radiation hardening controlledprimarily by dislocation loops
OAK RIDGE NATIONAL LABORATORYU. S. DEPARTMENT OF ENERGY
Effect of Strain Rate and Temperature onYield Strength Behavior of Unirradiated V-4Cr-4Ti
180
200
220
240
260
280
300
320
340
360
20°C300°C400°C500°C600°C700°C750°C800°C850°C
10-6 10-5 10-4 10-3 10-2 10-1 100
Lo
we
r Y
ield
Str
ess (
MP
a)
Strain Rate (s-1)
"10 MPa
Negative strain rate exponent correspondsto dynamic strain aging regime
180
200
220
240
260
280
300
320
340
360
380
400
420
440
0 100 200 300 400 500 600 700 800 900
Lower Yield StressFlow Stress
Str
es
s (
MP
a)
Temperature (°C)
O mobile (in V-5Ti alloy)
Ti mobile
V mobile
HomogeneousDeformation
InhomogeneousDeformation
CreepDeformation
Engin
eering S
tress (
MP
a)
Normalized Crosshead Displacement
10-3/s
20°C
200°C
300°C
500°C
600°C
400°C
700°C
750°C800°C
850°C
100
0.05
OAK RIDGE NATIONAL LABORATORYU. S. DEPARTMENT OF ENERGY
Effect of Strain Rate and Temperature onYield Strength Behavior of Irradiated V-4Cr-4Ti
Strain rate exponent becomes negative in irradiated alloys at Ttest>300oC even whendynamic strain aging serrations are not visible in stress-strain curves
500
550
600
650
700
750
10-5
0.0001 0.001 0.01 0.1
Lo
we
r Y
ield
Po
int
(MP
a)
Strain rate (s-1
)
V-5Cr-5Ti400˚C, 4 dpa, EBR-II
V-4Cr-4Ti (50%CW+1000˚C,2h)200˚C, 0.5 dpa, HFBR
Ttest
~Tirr
CR+1050˚C TWCA/1050˚C
OAK RIDGE NATIONAL LABORATORYU. S. DEPARTMENT OF ENERGY
What are the consequences of radiationhardening?
• Increased strength (good!)•Decreased tensile elongation (bad!)
– Practical impact/consequences: need to use more conservative structural designrules for uniform elongation <2%
• For BCC metals, increase in the ductile-brittle transition temperatureand decrease of toughness in the “ductile” regime (can becatastrophic!)– Radiation hardening also tends to reduce the fracture toughness of FCC metals
OAK RIDGE NATIONAL LABORATORYU. S. DEPARTMENT OF ENERGY
Examples of tensile curves for pure metals irradiated at ~330 K
Al0.35TM
Ni0.19TM
V0.15TM
Mo0.11TM
OAK RIDGE NATIONAL LABORATORYU. S. DEPARTMENT OF ENERGY
V
0
200
400
600
800
1E-05 0.0001 0.001 0.01 0.1 1 10
dpa
Str
ength
, M
Pa
YS
UTS
0.0
0
10
20
30
40
50
60
1E-05 0.0001 0.001 0.01 0.1 1 10
dpa
Elo
ng
atio
n,
%
UE
TE
0.0
Mo
0
200
400
600
800
1E-05 0.0001 0.001 0.01 0.1 1 10
dpa
Str
ength
, M
Pa
YS
UTS
0.0
0
10
20
30
40
50
60
1E-05 0.0001 0.001 0.01 0.1 1 10
dpa
Elo
ng
atio
n,
%
UE
TE
0.0
Ni
0
200
400
600
800
1E-05 0.0001 0.001 0.01 0.1 1 10
dpa
Str
ength
, M
Pa
YS
UTS
0.0
0
10
20
30
40
50
60
1E-05 0.0001 0.001 0.01 0.1 1 10
dpa
Elo
ng
atio
n,
%
UE
TE
0.0
Al
0
200
400
600
800
1E-05 0.0001 0.001 0.01 0.1 1 10
dpa
Str
ength
, M
Pa
YS
UTS
0.0
0
10
20
30
40
50
60
1E-05 0.0001 0.001 0.01 0.1 1 10
dpa
Elo
ng
atio
n,
%
UE
TE
0.0
Effect of neutron irradiation near 70˚C on tensile properties
K. Farrell et al.
Ni V Mo
Ductile FCC metal Ductile BCC metal Low-ductility BCC metalReduction of uniform elongation to <2% typically occurs within 0.001 to 1 dpa
OAK RIDGE NATIONAL LABORATORYU. S. DEPARTMENT OF ENERGY
Low uniform elongations occur in many BCC and FCCmetals after low-dose irradiation at low temperature
Uniform elongation of neutron-irradiated V and Nb
-2
0
2
4
6
8
10
12
0 100 200 300 400 500 600 700
Un
ifo
rm E
lon
ga
tio
n (
%)
Test Temperature (˚C)
0.5 dpa
0.1 dpa
Test Temperature isIrradiation Temperature
0
2
4
6
8
10
0.0001 0.001 0.01 0.1 1 10
Fabritsiev et al (1996)Heinisch et al (1992)Singh et al. (1995)
Un
ifo
rm E
lon
ga
tio
n (
%)
Damage Level (dpa)
Unirradiatedelongations
Tirr=36-90˚C
Uniform elongation of neutron-irradiated GlidCop Al25 and CuCrZr
0
5
10
15
20
25
30
0 50 100 150 200 250 300 350 400
Un
ifo
rm E
lon
ga
tio
n (
%)
Irradiation and Test Temperature (˚C)
CuCrZr
ODS Cu
GlidCop Al25
V-4Cr-4Ti
0.3TM
0.3TM
0
10
20
30
40
50
0.0001 0.001 0.01 0.1 1 10 100
Te
nsile
Elo
ng
atio
n (
%)
Displacement Damage (DPA) 0
1 dpa = 2x1021 n/cm2 (E>0.1 MeV)
Wiffen (1978)
UniformElongation
TotalElongation
Nb-1Zr Tirr=40-60˚C
Cu alloys
Low uniform elongation is induced aftervery small doses at low temperatures
0.4TM
OAK RIDGE NATIONAL LABORATORYU. S. DEPARTMENT OF ENERGY
Radiation hardening in V-4Cr-4Ti
0
200
400
600
800
1000
0 100 200 300 400 500 600 700
Effect of Dose and Irradiation Temperature onthe Yield Strength of V-(4-5%)Cr-(4-5%)Ti Alloys
Yie
ld S
tre
ng
th (
MP
a)
Irradiation Temperature (˚C)
Ttest~Tirr
unirradiated
0.1 dpa
0.5 dpa5-50 dpa
V-4Cr-4Ti
0.3 TM
OAK RIDGE NATIONAL LABORATORYU. S. DEPARTMENT OF ENERGY
Neutron Irradiation Data Show Low Ductility and HighHardening of Mo-Re alloys up to 1070 K (0.37 TM)
0
500
1000
1500
2000
0
1
2
3
4
5
6
7
8
600 700 800 900 1000 1100
Summary of Tensile Properties of Fast Neutron-Irradiated Mo-Re alloys
Hasegawa et al (1995)Fabritsiev&Pokrovsky (1998)
Ult
ima
te T
en
sil
e S
tre
ng
th (
MP
a)
Un
ifo
rm e
lon
ga
tio
n (
%)
Temperature (K)
Unirrad. UTS
Irrad. UTS
Irrad. eUIrrad. UTS
(brittle fracture)
0.3 TM
OAK RIDGE NATIONAL LABORATORYU. S. DEPARTMENT OF ENERGY
Summary of Tensile Properties of Neutron-irradiated Nb-1Zr in Li-bonded capsules
0
100
200
300
400
500
600
700
0
5
10
15
20
25
30
35
200 400 600 800 1000 1200 1400 1600
Summary of Tensile Properties of Neutron-Irradiated Nb1Zr in Li-bonded CapsulesU
ltim
ate
Te
ns
ile
Str
en
gth
(M
Pa
)
Un
ifo
rm e
lon
ga
tio
n (
%)
Temperature (K)
Unirrad. UTS
Irrad. UTS
Irrad. eU
0.3 TM
OAK RIDGE NATIONAL LABORATORYU. S. DEPARTMENT OF ENERGY
Fracture surface of Irradiated Nb-1Zr shows ductilebehavior, despite low uniform elongation value
F.W. Wiffen, unpublished results
OAK RIDGE NATIONAL LABORATORYU. S. DEPARTMENT OF ENERGY
The issue of localized flow (planar slip) has beenthe subject of numerous materials scienceinvestigations (unirradiated metals)
• Three general parameters may be considered to influence planarslip (e.g., Gerold & Karnthaler Acta Met. 39 (1989) 2177; Basinskiet al. Phil Mag. A 76 (1997) 743):– Stacking fault energy (weak effect)– Value of yield stress (weak effect)– Occurrence of short range ordering (solid solution alloys) or ordered
precipitates that are shearable (generally dominant effect)
OAK RIDGE NATIONAL LABORATORYU. S. DEPARTMENT OF ENERGY
Plastic deformation mechanisms for dispersion hardenedmaterials containing work-softening obstacles
Macro
scopic
pla
sti
c s
train
, ! p
/(1
-f)
cluster annihilation resistance, {2E"/f(1-f)µA}0.5
Uniform slip
Uniform slip
T. Mori and T. Mura, Mat Sci Eng. 26 (1976) 89
Hetero
gene
ous s
lip
Ee=elastic energystored by one obstacleρ=obstacle densityf=vol. fraction ofobstacles
OAK RIDGE NATIONAL LABORATORYU. S. DEPARTMENT OF ENERGY
0
100
200
300
400
500
600
700
0 0.05 0.1 0.15 0.2 0.25 0.3
Load-Elongation Curves for V-4Cr-4Ti Irradiated in HFBR to 0.5 dpa
Eng
inee
ring
Str
ess,
MP
a
Normalized Crosshead Displacement, mm/mm
110˚C 270˚C
325˚C
420˚C
Tt e s t
~Tirr
100 nm100 nm
TTtesttest=T=Tirrirr=270=270˚̊CC
g=011g=011
Irradiated Materials Suffer Plastic Instability(due to Dislocation Channeling?)
OAK RIDGE NATIONAL LABORATORYU. S. DEPARTMENT OF ENERGY
Overview of dislocation channeling
• Dislocation channeling is a viable mechanism to locally worksoften the matrix– Shearable obstacles
• Channeling involves localized flow and therefore inhibitsdislocation multiplication– Limited interaction between dislocation sources
However, …………….• It is not generally established that the catastophic reduction in
tensile elongation is directly due to dislocation channeling– High tensile elongations and significant work hardening rates can occur in
irradiated metals that exhibit dislocation channeling
OAK RIDGE NATIONAL LABORATORYU. S. DEPARTMENT OF ENERGY
Localized deformation (and dislocationchanneling) occurs in many irradiated materialsystems
Vanadium
316 SS316 SS
A533BA533B
Zircaloy-4
MolybdenumMolybdenum
CopperCopper
OAK RIDGE NATIONAL LABORATORYU. S. DEPARTMENT OF ENERGY
Dislocation channel interactions in Fe deformedfollowing neutron irradiation at 70˚C to 0.8 dpa
g.b.
Need well-engineered materials tomitigate neutron radiation effects
Clearedslipchannel
~200 nm~200 nmoffsetoffset
}}~80 nm~80 nmoffsetoffset}}
Zinkle & Singh, J. Nucl. Mater. 351 (2006) 269
OAK RIDGE NATIONAL LABORATORYU. S. DEPARTMENT OF ENERGY
Summary of Dislocation Channeling ParametersMaterial Dose
(dpa)Irrad.Temperature
TestTemperature
Channelwidth (nm)
Channelspacing
Reference
Cu ~0.001-0.5 313-323 K 295 K ~50-200 0.5-2.3 µm 9 research groups
Cu ~0.5 ~50˚C 50 K77 K295 K
5070100
0.8-1.2 µm Howe 1974
CuCrZr 0.3 323 K 323 K 100-300 Singh & Stubbins fatigue
Cu-0.8%Co ~0.001 ~50˚C 295 K 160 1.2-2.3 µm Sharp 1974
Cu-0.05%Al ~0.001 ~50˚C 295 K 240 1.2-2.3 µm Sharp 1974 (channels notobserved in Cu-4%Al)
Au ~0.003 20˚C 295 K ~100 Okada 1989
Ni ~50˚C 295 K ~300 Noda 1977
Pd 0.3 ~50-100 Victoria et al 2000
304L SS 5 (ions)“
500˚C“
563 K295 K
1550-200 (twins)
Brimhall 1995
Fe-10%Cr-30%Mn
~0.005 300˚C ~100? F. Abe 1992
!-Fe ~0.38 323 K 323 K 50-200 Singh et al.;Victoria et al 2000
Nb ~0.002 ~400 Tucker 1969
V 0.1-0.8 330 K 295 K 20-80 0.5-2.5 µm Arsenault 1977; Farrell 2002
V-4Cr-4Ti 0.5-5 500-673 K 295-673 K ~50-100 Rice&Zinkle 1998, Gazda 1998
Mo ~0.2 323 K 323 K ~500 Luft 1991; Singh et al.
TZM ~0.2 373 K 373 K 100-200 Singh et al.
Re Pitt 1980
Zircaloy-2,4 425-563 K 293-573 K 40-100 Coleman 1972, Onchi 1980
Au quenched 160-500 Yoshida 1968, Bapna 1974
Al quenched 650-1000 Mori 1969; Tokuno 1987
Dislocation channel width is ~100 nm for a wide range of experimental conditions
OAK RIDGE NATIONAL LABORATORYU. S. DEPARTMENT OF ENERGY
Investigation of dislocation interactions with SFTsSFT annihilation by a single dislocation (remnant apex)
Y. Matsukawa, ORNL
OAK RIDGE NATIONAL LABORATORYU. S. DEPARTMENT OF ENERGY
Dislocation interaction with SFTs in quenched Au
• Type 1 interaction (Frank loop formation) at room temperatureMatsukawa, Stoller, Osetsky & ZinkleJ. Nucl. Mater. 351 (2006) 285
OAK RIDGE NATIONAL LABORATORYU. S. DEPARTMENT OF ENERGY
Dislocation interaction with SFTs in quenched Au
• Type 2 interaction at room temperature (superjog creation with noSFT remnant)
Matsukawa, Stoller, Osetsky & ZinkleJ. Nucl. Mater. 351 (2006) 285
OAK RIDGE NATIONAL LABORATORYU. S. DEPARTMENT OF ENERGY
Dislocation interaction with SFTs in quenched Au
• Types 1(a) & 2(b) interactions also occur at 100 K
Matsukawa, Stoller, Osetsky & ZinkleJ. Nucl. Mater. 351 (2006) 285
OAK RIDGE NATIONAL LABORATORYU. S. DEPARTMENT OF ENERGY
Dislocation interaction with SFTs in quenched Au
• Type 3 interactions at room temperature (SFT apex remains);not observed at 100 K
Matsukawa, Stoller, Osetsky & ZinkleJ. Nucl. Mater. 351 (2006) 285
OAK RIDGE NATIONAL LABORATORYU. S. DEPARTMENT OF ENERGY
Interaction of a screw dislocation with 78-vacancy SFT and 91-intersitital cluster in Cu thin foil
300 K
Cooperative effects may be important for annihilation of sessile defectclusters by gliding dislocations during deformation
Yu.N. Osetsky
OAK RIDGE NATIONAL LABORATORYU. S. DEPARTMENT OF ENERGY
Effect of Irradiation Dose and Strain onDeformation Modes of FCC Metals
0
5
10
15
20
0.0001 0.001 0.01
Deformation Mode Diagram for Neutron-irradiated Au
Elo
ng
ati
on
(%
)
Displacement Damage (dpa)
Uniform elongation limit
Dislocationchannels
Cell Structure
Au14 MeV n irrad.
Tirr=T test=295 K
~0.5 µm
~0.3 µm
~0.3 µm
0
2
4
6
8
10
12
14
16
0.0001 0.001 0.01
Deformation Mode Diagram for Neutron-irradiated Ni
Elo
ng
ati
on
(%
)
Displacement Damage (dpa)
Uniform elongation limit
Dislocationchannels
Cell Structure
Ni14 MeV n irrad.
Tirr=T test=295 K
~0.5 µm
~0.5 µm
~2 µm
~2 µm
A. Okada et al. Mater. Trans. JIM 30 (1989) 265
OAK RIDGE NATIONAL LABORATORYU. S. DEPARTMENT OF ENERGY
Tensile Deformation Behavior in Irradiated Cu
40
60
80
100
120
140
160
0 100 200 300 400 500
Tensile Deformation Behavior in Irradiated Cu
Cri
tic
al
Re
so
lve
d S
he
ar
Str
es
s (
MP
a)
Temperature (K)
No channels
!t=4x1024/m2, E>1 MeV (~0.5 dpa)L.M Howe, Rad. Eff. 23 (1974) 181
As-irradiated
0.25 h, 723 K1 h, 723 K
5 h, 723 K
Similar behavior observed in quenched Au(S. Yoshida, Trans. JIM 9 suppl. (1968)83)
Engineering and true stress-strain tensile curves for stainlesssteel before and after spallation irradiation at ~100˚C
HTUPS316 Stainless Steel
Unirradiated
1.2
2.8
4.0 dpa
0.40.6
1.1
0
200
400
600
800
1000
0 10 20 30 40 50 60 70
Elongation, %
Engin
eering S
tress, M
Pa
HTUPS316 Stainless Steel
Unirr.
1.22.8
4.0 dpa0.4
0
200
400
600
800
1000
1200
1400
0.0 0.1 0.2 0.3 0.4 0.5 0.6
True strain
Tru
e s
tre
ss,
MP
a
EC316LN Stainless Steel 10.7 dpa
3.6
2.5
1.1
0.5 Unirr.
0
200
400
600
800
1000
0 10 20 30 40 50 60 70
Elongation, %
En
gin
ee
rin
g S
tre
ss,
MP
a
EC316LN Stainless Steel
Unirr.10.7 dpa
1.12.5
3.6 0.5
0
200
400
600
800
1000
1200
1400
0.0 0.1 0.2 0.3 0.4 0.5 0.6
True strain
Tru
e s
tress, M
Pa
T.S. Byun et al., J. Nucl. Mater. 298 (2001) 269
OAK RIDGE NATIONAL LABORATORYU. S. DEPARTMENT OF ENERGY
Radiation Effect in True Stress-True Strain Curve (FCC)
EC316LN
unirradia
ted
10.7 dpa
1.12.5 3.6
0.5
0
200
400
600
800
1000
1200
0 0.1 0.2 0.3 0.4 0.5
True strain
Tru
e s
tre
ss
True stress-true straincurves for EC316LNstainless steel; the curves ofirradiated specimens areshifted in the positivedirection by strains of 0.14,0.18, 0.23, 0.28, and 0.385,respectively, to superimposeon the curve of unirradiatedmaterial. Irradiation-induced increases in yieldstress were 305, 358, 421,485, and 587 MPa,respectively.
T.S. Byun & K. Farrell, Acta Mater. 52 (2004) 1597
Further work is need to determine importance of dislocation channeling in observedexhaustion of initial strain hardening capacity
OAK RIDGE NATIONAL LABORATORYU. S. DEPARTMENT OF ENERGY
A533B
unirr
adia
ted
0.003 0.015 dpa
0
200
400
600
800
1000
0 0.03 0.06 0.09 0.12 0.15
True strain
Tru
e s
tre
ss,
MP
aRadiation Effect on True Stress-True Strain Curve (BCC)
True stress-true straincurves for A533B steel; thecurves of irradiatedspecimens are shifted inthe positive direction bystrains of 0.035 and 0.09,respectively, tosuperimpose on the curveof unirradiated material.Irradiation-inducedincreases in yield stresswere 132 and 235 MPa,respectively.
T.S. Byun & K. Farrell, Acta Mater. 52 (2004) 1597
• True stress- true strain curve for irradiated material coincides withunirradiated curve, after translation to account for radiation hardening– Suggests main effect of radiation hardening is to partially exhaust strain
hardening capacity
132 MPa235 MPa
OAK RIDGE NATIONAL LABORATORYU. S. DEPARTMENT OF ENERGY
Plastic Instability Stress (σPI) of BCC Metals
• Plastic Instability Stress (σPI) = the true stress version of Ultimate Tensile Stress• Plastic Instability Stress is independent of dose when yield stress < σPI.• Yield stress can be > σPI, which is defined only when uniform deformation exists.• σPI is considered to be a material constant, independent of initial cold-work or radiation-
induced defect clusters
σML=true stress at maximum load
0
200
400
600
800
1000
1200
1400
1E-05 0.0001 0.001 0.01 0.1 1 10
dpa
Str
ess,
MP
a
A533B (HFIR)a A533B (HFIR)a
A533B (HFIR)b A533B (HFIR)b
A533B (HFIR)c A533B (HFIR)c
? ML ? YS
? ML????IS ?ML????YS
?YS
?ML????IS = ? YS
Prompt
necking
at yield
Uniform
deformation
exists before
necking DC
0
σML=σUTS =σPI σML=σYS
σYS σML=σIS =σYS
σYSσML
T.S. Byun & K. Farrell, Acta Mater. 52 (2004) 1597
400
600
800
1000
1200
1400
1E-05 0.0001 0.001 0.01 0.1 1 10 100
dpa
Str
ess, M
Pa
3Cr-3WV (HFIR) 3Cr-3WV (HFIR)
9Cr-1MoVNb (HFIR) 9Cr-1MoVNb (HFIR)
9Cr-2WV (HFIR) 9Cr-2WV (HFIR)
9Cr-2WVTa (HFIR) 9Cr-2WVTa (HFIR)
9Cr-1MoVNb (LANSCE) 9Cr-1MoVNb (LANSCE)
9Cr-2WVTa (LANSCE) 9Cr-2WVTa (LANSCE)
?ML ? YS
0
σML σYS
A533B pressure vessel steel Ferritic/martensitic steels
OAK RIDGE NATIONAL LABORATORYU. S. DEPARTMENT OF ENERGY
Plastic Instability Stress (σPI) ofFCC Metals irradiated near 70˚C
T.S. Byun & K. Farrell, Acta Mater. 52 (2004) 1597
0
200
400
600
800
1000
1200
1400
1600
1800
1E-05 0.0001 0.001 0.01 0.1 1 10 100
dpa
Str
ess,
MP
a
EC316LN (LANSCE) EC316LN (LANSCE)
HTUPS316 (LANSCE) HTUPS316 (LANSCE)
AL6XN (LANSCE) AL6XN (LANSCE)
316 (HFIR)a 316 (HFIR)a
316LN (HFIR) 316LN (HFIR)
316 (HFIR)b 316 (HFIR)b
?ML ?
YS
0
σML σYS
0
200
400
600
800
1000
1E-05 0.0001 0.001 0.01 0.1 1 10
dpa
Str
ess,
MP
a
Ni (HFIR) Ni (HFIR)
Cu (HFIR) Cu (HFIR)
? ML ? YS
0
σML σYS
Type 316 austenitic stainless steel Pure Cu, Ni
Ni
Cu
OAK RIDGE NATIONAL LABORATORYU. S. DEPARTMENT OF ENERGY
Plastic Instability Criterion for BCCmetals irradiated near 70˚C
V
0.00120.0001
0
100
200
300
400
500
600
700
0 10 20 30 40
Elongation, %
En
gin
ee
rin
g s
tre
ss,
MP
a
0.012
0.12 dpa
0.69
PIS = 390 MPa
A533B
Unirradiated
0.015
0.054
1.2 dpa
0.003
0
200
400
600
800
1000
1200
0 5 10 15 20 25
Elongation, %
En
gin
ee
rin
g s
tre
ss,
MP
a
PIS =670 MPa
• Prompt plastic instability at yield occurs when yield stress > σPI.• σPI is constant for unirradiated and irradiated conditions; implies that σPI is acriterion for plastic instability
σPI=670 MPaσPI=390 MPa
OAK RIDGE NATIONAL LABORATORYU. S. DEPARTMENT OF ENERGY
Plastic Instability Criterion (FCC & HCP) irradiated at ~70˚C
EC316LN
10.7 dpa
3.6
2.5
1.1
0.5 Unirradiated
0
200
400
600
800
1000
1200
0 10 20 30 40 50 60
Elongation, %
En
gin
ee
rin
g S
tre
ss,
MP
a
PIS = 890 MPa
Zircaloy-4
unirradiated0.001
0.1
0.80 dpa
0.01
0
200
400
600
800
0 10 20 30 40
Elongation, %
En
gin
ee
rin
g s
tre
ss,
MP
a
PIS=510 MPaσPI=510 MPa
σPI=890 MPa
OAK RIDGE NATIONAL LABORATORYU. S. DEPARTMENT OF ENERGY
Deformation mechanisms in stainless steel
10-6
10-5
10-4
10-3
10-2
10-1
0 0.2 0.4 0.6 0.8 1
Deformation Map for 316 Stainless SteelIrradiated to 1 dpa
No
rma
lize
d S
he
ar
Str
es
s, !/µ
(20˚C
)
Normalized Temperature, T/TM
Elastic regime
(d"/dt<10-8 s-1)
Coble creep
Dislocation creep
Dislocation glide
Theoretical shear stress limit
20˚C 300˚C 650˚C
24
240
Un
iax
ial
Te
ns
ile
Str
es
s,
MP
a
2.4
0.24
2400
10- 8 s- 1, L=50 µm
N-H creep
Twinning
10-6
10-5
10-4
10-3
10-2
10-1
0 0.2 0.4 0.6 0.8 1
Deformation Map for 316 Stainless Steel
No
rma
lize
d S
he
ar
Str
es
s, !/µ
(20˚C
)
Normalized Temperature, T/TM
Elastic regime
(d"/dt<10-8 s-1)
Coble creep
Dislocation creep
Dislocation glide
Theoretical shear stress limit
20˚C 300˚C 650˚C
24
240
Un
iax
ial
Te
ns
ile
Str
es
s,
MP
a
2.4
0.24
2400
10- 8 s- 1, L=50 µm
N-H creep
Twinning
Irradiation induces changes incontrolling deformation mechanisms
Channeling (Disln glide) occurs at highertemperatures (~300˚C)
Twinning occursat lower
temperatures(<200˚C) and high
strain rates
OAK RIDGE NATIONAL LABORATORYU. S. DEPARTMENT OF ENERGY
Calculated irradiated Ashby deformation mapfor V-4%Cr-4%Ti
10-6
10-5
10-4
10-3
10-2
10-1
0 0.2 0.4 0.6 0.8 1No
rma
lize
d S
he
ar
Str
es
s, !/
µ(2
0˚C
)
Normalized Temperature, T/TM
Elastic regime
(d"/dt<10-10 s-1)
Coble creep
Dislocation creep
Dislocation glide
Theoretical shear stress limit
14
141
Un
iax
ial
Te
ns
ile
Str
es
s,
MP
a
1.4
0.14
1410
10-10 s-1, L=30 µm
N-H creep
20˚C 400˚C 700˚C
Irrad. creep
Damage rate = 10-6 dpa/s
OAK RIDGE NATIONAL LABORATORYU. S. DEPARTMENT OF ENERGY
Tensile toughness is not a reliable indicator offracture toughness in BCC metals
• Magnitude of radiation hardening (overall matrix strength) is most importantparameter for fracture toughness in BCC metals
0
200
400
600
800
1000
0 2 4 6 8 10
Effect of Test Temperature on the Tensile Stress-Strain Curve of V-4Cr-4Ti Irradiated to 4 dpa at 400˚C
En
gin
ee
rin
g S
tre
ss (
MP
a)
Engineering Strain (%)
Ttest
=20˚C
Ttest
=390˚C
! ! d"=26 MPa
! ! d"=66 MPa
0
5
10
15
20
-200 0 200 400
Effect of Neutron Irradiation at ~400˚C on the Charpy Impact Properties of V-4%Cr-4%Ti
Ab
so
rbed
En
erg
y (
J)
Test Temperature (˚C)
4 dpa,Tirr=400˚C
V-4Cr-4Ti
unirradiated
0.5 dpa,Tirr=420˚C
Charpy V-notch
Fracture Toughness of BCC Metals:Master Curve Approach
• Ludwig-Davidenkov relation providesa rough estimation of embrittlementdue to radiation hardening
• The measured DBTT depends on numerous parameters,including strain rate and amount of physical constraint incracked sample
• Important parameters such as constraint andstrain rate are included in recent advancedapproaches (Master curve)
0
200
400
600
800
1000
-200 0 200 400 600
Yie
ld S
tre
ng
th,
MP
a
Temperature, °C
unirradiated
irradiated
!*/M
UCSB
0
100
200
300
400
500
Ke (
MP
a!m
)
-250-150 -50 50 150 250 350 450
T (°C)
!T ! 145°C
OAK RIDGE NATIONAL LABORATORYU. S. DEPARTMENT OF ENERGY
Low temperature radiation hardening causesfracture toughness embrittlement in BCC metals
0
200
400
600
800
1000
1200
-200 -100 0 100 200 300 400 500 600
Effect of Irradiation Temperature and Doseon the Yield Strength of V-4%Cr-4%Ti Alloys
Yie
ld S
trength
(M
Pa)
Irradiation and Test Temperature (˚C)
~0.5 dpa
~4-15 dpa
!f*
DBTT0.5 dpa DBTT4 dpaDBTTunirr.
unirradiated
0
5
10
15
20
-200 0 200 400
Effect of Neutron Irradiation at ~400˚C on the Charpy Impact Properties of V-4%Cr-4%Ti
Ab
so
rbe
d E
ne
rgy
(J
)
Test Temperature (˚C)
4 dpa,Tirr=400˚C
V-4Cr-4Ti
unirradiated
0.5 dpa,Tirr=420˚C
Charpy V-notch
OAK RIDGE NATIONAL LABORATORYU. S. DEPARTMENT OF ENERGY
•Macroscopic Mode I fracture is composed ofcoordinated Mode III shear displacements at the crack tip
TEM in-situ deformation studies can be used to provideinsight on fundamental fracture processes
Atomic resolution imaging of ductile crack propagation (plane stress)
2 nmY. Matsukawa, ORNL
OAK RIDGE NATIONAL LABORATORYU. S. DEPARTMENT OF ENERGY
Summary and Conclusions•Low temperature (<0.3 TM) irradiation causes rapid
hardening and reduction in uniform elongation of metals•When radiation hardening increment becomes a significant fraction of the
unirradiated strength, new dependencies on test temperature and strainrate may occur
•Uniform elongation is typically reduced to <1% after doses of ~0.1 dpa•Responsible mechanisms for loss of tensile ductility (dislocation
channeling, radiation hardening) are currently being investigated•Radiation hardening causes fracture toughness embrittlement
and increases in the ductile-to-brittle-transition temperature inirradiated BCC metals•Can be mitigated by alloying additions that reduce radiation hardening or
increase σ*