Fundamentals of Fusion Materials Radiation Effects

Akihiko Kimura

Institute of Advanced Energy Kyoto University

JAPAN

Fundamentals of

Fusion Materials Radiation Effects

Tutorial Seminar T4

Contents

1. Generals

1) defect structure 2) defect motion 2. Radiation effects

1) hardening/loss of elongation 2) swelling

3) helium effects 3. Radiation tolerance 4. Summary

2

Radiation Induced Defects

3

Vacancy (V) V cluster V-He pair V-He cluster V-solute pair V-S cluster

Interstitial (I) I cluster

I-He pair I-He cluster I-solute pair I-S cluster

Clustering

Precipitation

Segregation

Dissolution

Disordering

4

Examples of Radiation-induced Defects

a) Pure iron, 30 MeV electron, 300C/0.1dpa dislocation loops

b) Fe-0.3Cu alloy, 30 MeV electron, 300C/0.1dpa dislocation loops and Cu clusters

d) 12Cr-Ferritic steel, FFTF, 400C/20 dpa Cr-rich phases

e) 9Cr-Martensitic steel, FFTF, 460C/70 dpa voids

f) 9Cr-Martensitic steel, 6.4 MeV-Fe ions + 1 Mev He ions, 500C/60 dpa

voids + He

The type of the defect depends on the irradiation conditions :

1) particles/energy 2) temperature

3) dose/dose rate

c) 9Cr-Martensitic steel, FFTF, 400C/20 dpa dislocation loops and carbides

5

Examples of Radiation-induced Defects

a) Pure iron, 30 MeV electron, 300C/0.1dpa dislocation loops

b) Fe-0.3Cu alloy, 30 MeV electron, 300C/0.1dpa dislocation loops and Cu clusters

d) 12Cr-Ferritic steel, FFTF, 400C/20 dpa Cr-rich phases

e) 9Cr-Martensitic steel, FFTF, 460C/70 dpa voids

f) 9Cr-Martensitic steel, 6.4 MeV-Fe ions + 1 Mev He ions, 500C/60 dpa

voids + He

The type of the defect depends on the irradiation conditions :

1) particles/energy 2) temperature

3) dose/dose rate

c) 9Cr-Martensitic steel, FFTF, 400C/20 dpa dislocation loops and carbides

6

Muti-scale Observation

Point defects

10^-4

10^-5

10^-6

10^-7

10^-8

10^-11

(m)

10^-10 10^-9 1 μm

100 nm

10 nm

0.01 nm 0.1 nm

1 nm 10 μm 100 μm

Dislocation loops

Voids, Cavities, Bubbles Dislocation recovery

Solute clusters

TEM PAS ^APF

Characteristics of Experimental Techniques

7

Method Main defects Analysis Probe Comments

TEM ・I-loops, -clusters

・ V-loops, -clusters

・ Dislocations

・ Solute clusters

・Defect size

・ Number density

・ 3D distribution

・ Lattice structure

・ Chemical analysis

Electron density

・invisible for vacancies and small defects

PAS ・ Vacancy

・ V-clusters

・ V-S clusters

・ Void size

・ Number density

・ Chemical analysis

Positron lifetime

・ indirect

3D-AP ・ Solute-clusters ・ Cluster size

・ Number density

・ 3D distribution

Ion energy and flight time

・ relative position

Complementary examinations are necessary

Motion of Point Defects

8 A B

Interstitial atom

a) Interstitial mechanism

b) Interstitialcy mechanism B A

B A Vacancy mechanism

Pure Metals

Solute + Interstitials

Mixed dumbbell with rotation Solute +

Vacancy

Vacancy mechanism

Cooperative Motion of Defect Clusters

9 Dislocation loop

Not atoms but strain field moves long distance.

Crowdion 1D-motion

Loop 1-D motion

b direction

Thermal Diffusion

10 The probability of a vacancy existing at the next lattice site (P^F):

P

= n/N = exp (-G

/kT)

G^F: Formation energy of vacancy The probability of an atom moving to the next lattice site (PM):

P

=  exp (-G

/kT)

G_M: Migration energy of vacancy

The jumping frequency (f):

f =  Z exp (-GM/kT) exp (-GF/kT) Z: variant (bcc (6), fcc (8))

The diffusivity (D): b: jump distance

State 1 3

G M

2

G M

Energy, G

Diffusivity depends on vacancy concentration and its mobility.

The diffusivity (D):

Diffusion under Irradiation

11

How does irradiation enhance the diffusion of atoms?

G

: vacancy migration energy

Under irradiation, lattice atoms are knocked on by energetic particles.

Irradiation causes displacement of atoms.  reduction of G

G

: vacancy formation energy

Under irradiation, vacancies as well as interstitial atoms are generated irrespective of temperature.

Irradiations supply vacancies which enhances atomic diffusion. 

reduction of G

{ }

D = fb ¹ ₆ ² = b ¹ ₆ ²  Zexp ^ー (G _M +G _F )/kT

D

1/T

ー

(G

+G

)/k

G

, G

: small

Under irradiation

Thermal

diffusion

Contents

1. Generals

2. Radiation effects

1) hardening/loss of elongation 2) helium effects

3) swelling

3. Radiation tolerance 4. Summary

12

Radiation Effects in This Lecture

1) Irradiation hardening and loss of elongation a) hardening mechanism

b) localized deformation

2) Irradiation and helium embrittlement a) DBTT shift and reduction of USE b) helium effect

3) Void swelling

a) irr. temperature b) helium effect

**Mostly based on research on ferritic steels

/

RAFS (mod.JLF - 1) Tested at RT Cross - head speed

= 0.5 mm/min

0 20 40 60

0 200 400 600 800

- 200 - 600 ° C, FFTF

400 - 430 ° C, FFTF 375 °C, FFTF

(250 - 270 ºC, HFIR, F82H)

350

dpa

Irradiation Hardening, Ds

y

MPa

RAFS (mod.JLF - 1) Tested at RT Cross - head speed

= 0.5 mm/min

0 20 40 60

0 200 400 600 800

- 200 460 -

400 - (250 -

HFIR, F82H)

ºC, HFIR

RAFS(9Cr-2W martensitic steel)

◆ Irradiation hardening depends on irradiation temperature.

Below 430ºC: hardening

Above 460ºC: softening (recovery)

◆ The irradiation response is directly correlated with irradiation-induced microstructure changes.

◆ Mechanism:

Hardening: interstitial disl. Loops

Softening: dislocation recovery

Irradiation Hardening-Temperature (1)

14

Irr. temperature is more critical

than dose for hardening.

Irradiation Hardening-Temperature (2)

RAFS (9Cr-2W), Neutron Irradiation, void images

15

365C, 10dpa Small I-loops

Small voids

420C, 40dpa Large I-loops

Large voids

520C, 40dpa 600C, 33dpa No I-loop, No void

Recovery of martensitic phase

900

Experimental Results

Void Size Swelling Softening

Hardening M 23 C

6 (L)

M ₆ C (M)

Ni addition

Δσｙ

（＋） （－） Swelling,etc Irradiation

Temperature (K) No Voids

No Loops

Voids & Loops

Microvoids Invisible Precipitates and/or Loops (?)

Elemental Processes

Carbon Migration to C-V pair

Break-up of C-V complexes Fe ₃ C precipitation Evaporation of Vacancy Clusters M 23 C

6 (L) M 23 C

6 (S)

Void Swelling Laves-P

Decomposition of C-V pair

Loop Density

800 700 600 500 400

Materials response to irradiation depends markedly on irradiation temperature.

Irradiation effects can be correlated with microstructural evolution.

Irradiation Hardening-Temperature (3)

Void Density

16

Interstitial clusters and carbides are formed by neutron irradiation.

Irradiation Hardening/Microstructure

17

b=a<001>

-- Interstitial dislocation loops --

Irradiation Hardening Mechanism (1)

200

100

50

0 150

As-irr. 400C 500C 600C

Annealing Temperature, C

Irradi atio n Ha rde ni ng , MPa

TEM total

dislocation loop, lines M₂₃C₆

Tensile test results

50 nm FFTF, 370C, 10 dpa

18

Ds _total =  (Ds _loop ) ² + (Ds _prec .) ²

0 20 40 60 80 100

DHV

Mod.JLF-1 60

Intensity /%

Annealing Temperature / C

unirr. asirr. t₁ 0

20 40

0 100 200 300 400 500

100 200 300 400

Lifetime /psec

500 t₂

I₂ Mod.JLF-1+1%Ni

Annealing experiments indicates that there is no good coincidence between the recovery behavior of microvoids and hardening.

0 20 40 60

Intensity , I

/ % 0

20 40 60 80 100 120

Incre ase in HV , D HV

As-irr. 200 300 400 500 600 700 Annealing Temperature / ℃

Hv

He-implanted (580appm)

0.22dpa, <150℃

--Vacancy clusters or microvoids--

Irradiation Hardening Mechanism (2)

19 ATR, 270°C, 2.2 dpa

Irradiation Embrittlement (1)

(1) JMTR/363K, 0.006dpa: RT (2) MOTA/663K, 22dpa: RT (3) MOTA/663K, 35dpa: RT (4) MOTA/733K, 24dpa: RT (5) MOTA/683K, 36dpa: RT (a) JMTR/493K, 0.15dpa: RT (b) EBR-II/663K, 26dpa: RT (c) MOTA/638K, 7dpa: 638K (d) HFIR/323K, 5dpa: RT (e) HFIR/673K, 40dpa: 673K

0 50 100 150 200 250 300

-200 -100 0 100 200 300 400 500

9Cr-2W

9Cr-1Mo

Shift in DBTT / K

Irradiation Hardening / MPa (1) (2) (3)

(b)

(4) (5)

(a)

(c) (d)

(e)

(b)

580appm He 120appm He

9Cr-1Mo

9Cr-2W

0 10 20 30 40 50

dpa 323-673K

363-733K

1) A linear relationship between DDBTT and Ds

.

2) Saturation of DDBTT similar to that of Ds

at 10 dpa.

20

Irradiation Embrittlement (2)

100 150 200 250 300 350 400 Test Temperature / K

0 0.2 0.4 0.6 0.8 1.0

Absorbed Energy / J

0 200 400 600 800 1000

Tensile Stress / MPa

0 5 10 15 20

Elongation / % unirr.

270C, 2.2dpa Tested at R.T.

RAFS (9Cr-2W)

Tensile tests at RT (300K) which is in the USE region revealed the fracture mode is completely ductile mode and accompanied by a large reduction of total

elongation.

Deformation at USE

Unirradiated:

Homogeneous deformation (ductile mode fracture) Irradiated:

Localized deformation (ductile mode fracture)

Irradiation hardening causes localized deformation and accelerates necking or ductile fracture that results in the loss of total elongation.

Deformation at LSE

Irradiated:

Almost no deformation (brittle mode fracture)

USE

LSE

21

Irradiation Embrittlement (3)

“Continuum modeling of localized deformation in irradiated bcc materials”

A Patra, D.L. McDowell, Journal of Nuclear Materials, 432 (2013) 414

Parametric studies of the cross-slip and flow softening (due to annihilation of

irradiation-induced defects) models are performed to study their effects on the localization behavior.

Localized deformation due to the formation of dislocation channels.

22

Contents

1. Generals

2. Radiation effects

1) hardening/loss of elongation 2) swelling

3) helium effects

3. Radiation tolerance

4. Summary

● Void swelling

1 mm V-5Fe alloy

Vacancy Cluster

FFTF irradiation for about 1 year (30 dpa), (Matsui, H.)

Void Swelling (neutron)

0 10 20 30 40 50

0 50 100 150 200 250

Void Swelling / %

dpa

by D.S. Gelles &

A. Kimura 316-Ti, 500C

12Cr Ferritics FFTF 425C 9Cr Ferritics

FFTF 425C

24

Z=<110> Z=<112> Z=<100> Z=<111>

RAFS (9Cr-2W), Neutron Irradiation, 420C, 40 dpa

Void Shape (neutron)

Truncated dodecahedron

25

Void Swelling BCC/FCC (neutron)

F82H 9Cr-2W

0 50 100 150 dpa

AISI316

0.2%/dpa

1%/dpa

600 650 700 750 800 850 900 Temperature (K)

SUS316

Ferritic: 0.2%/dpa, peak-temp. ~425°C Austenitic: 1%/dpa, peak-temp. ~500°C

26

(by Miwa)

(by Garner)

(by Morimura)

(by Mansor)

27

Void Swelling (ion)

Displacement damage [ dpa ]

Void Swelling [ % ]

0 20 40 60 80 100

0 0.5 1.0 1.5 2.0

FFTF/MOTA 420C DuET

470C

JLF-1 (9Cr-2W)

500nm

Dual-Ion Irradiation 470°C ( 6.4MeVFe³⁺ + 1.0MeVHe⁺ )

Distance from Irradiated Surface [ nm ] Helium Injection rate [ appm He/dpa ] Displacement Dose [ dpa ]

dpa Helium 0

500

1000

1500

2000

0 10 20 30 40

0 50 100 150

50nm

Fe-ions + He-ions

 void swelling (20 dpa) Fe-ions

 no swelling (20 dpa)

200 dpa

28

Effects of He & H on Void Swelling (1)

no He 100 appm He

F82HEK181EP450

EP450

no He/H

200 appm He 2000 appm H

S=6.8% S=0.4%

200 dpa

 Separately introduced He and H enhance the nucleation of voids.

 However, above 200 dpa He appears to suppress swelling.

 Strong synergisms exist between the two gases.

by Voyevodin (KIPT) , Garner, Maloy et al.

Effects of He on Void Swelling (2)

void #

void size

Temperature  la rge Sma ll 

He

Sw el ling 

dpa 

He

void nucleation rate

∞ exp 𝐵 𝑘𝑇 He  increase B

void growth rate

∞ exp −𝐸′ 𝑘𝑇

He  decrease E’

Temperature  la rge Smal l 

1) He increases the nucleation rate of voids by increasing thermal stability of voids.

 larger swelling before steady growth 2) He decreases the growth rate of voids by trapping vacancies at the voids.

 delayed appearance of steady growth Strong synergisms exist between the two gases (He & H).

29

Contents

1. Generals

2. Radiation effects

1) hardening/loss of elongation 2) swelling

3) helium effects 3. Radiation tolerance 4. Summary

30

HT9+2Ni T9+2Ni

HT9

He implantation

10

B-neutron

0 100 200 300 400 500 600

He Concentration (appm)

DB TT Sh ift of 10 CVN (K )

0 100 200 300 400 500

Ni-neutron

B-neutron

Effect of Helium

RAFS

isotope tailoring (B, Ni) or He-ions

DBTT shift is larger in isotope tailoring (B, Ni) results.

However, it is rather small in He-ions.

(DDBTTs are normalized to those of standard size specimens)

31

Implanted Area

Un-implanted

Un-implanted 200 240 280 320 360 400 0

0.2 0.4 0.6 0.8

1

Hv

Distance / mm

200 240 280 320 360 400

0 0.5 1 1.5 2 2.5 3 3.5 4

Hv

Distance / mm

500℃,50appm He RT,50appm He

550℃,1000appm He Implanted（This work）

Helium Effect on Hardness

He-implantation (F82H):

- AVF Cyclotron - 50MeV a-particle - 550

C

- 580, 1000appm, - 0.4 dpa

- Depth: 380 mm (with degrader)

32

580appmHe-implantation at <150 ℃ Isochronal annealing to 600 ℃ He bubble formation

50mm 500mm

10mm

Fractured at 77K

No intergranular cracking but cleavage fracture

50nm

Helium Effect on Fracture

33

0 0.2 0.4 0.6 0.8 1 1.2

-150 -120 -90 -60 -30 0

Absorbed Energy / J

Test Temperature / ^oC

as-received He implanted

0 0.2 0.4 0.6 0.8 1 1.2

-150 -120 -90 -60 -30 0

Absorbed Energy / J

Test Temperature / ^oC

as-received He implanted

100 m m _{20 m m}

20 m m

Helium Effect on DBT Behavior

He implanted area Intergranular fracture

No He area

Cleavage fracture

He-implanted

F82H

550°C, 0.4 dpa, 1000 appmHe

He

Fracture mode was changed from CL to IG after He-ion irradiation at

550°C (0.4 dpa, 1000 appmHe).

There is a limit of helium trapping capacity in the martensitic structure above which helium induces grain

boundary embrittlement. ₃₄

FA PR2 PR1 CW Fluence:10 ¹⁸ He/m ²

Desorbed Fraction of He, (Desorbed He)/(Total Retention)

0 200 400 600 800

0 2 4 6 8 10x10^-3

Temperature,T/ ^o C Cold Worked

400°C, 2h 600°C, 12h

800°C, 2h

Desorbed Fraction of Helium ,(Desorbed He)/(Total Retention) Fluence:10¹⁸He/m²

(2000 at.ppmHe) 150 eV

Pure iron (cold worked and annealed) 150eV: no damage, 2000at.ppmHe

● The peak at 500-600°C evolves with increasing dislocation density.

⇒ He trapping by dislocations

● Desorbed He fraction at 500-600 ° C is 35% (700 at.ppmHe) for cold worked iron.

Helium Trapping at Dislocations

35

Helium atoms are trapped at the strain field of the lattice defects, which may be one of the methods to suppress helium induced grain boundary

embrittlement.

Contents

1. Generals

2. Radiation effects

1) hardening/loss of elongation 2) swelling

3) helium effects 3. Radiation tolerance 4. Summary

36

Hardening/Loss of Elongation

37

Phase stability Small

embrittlement

RAFS ： 9Cr-2W F/M steel （ FFTF/MOTA irradiation, PNNL/USA ～ 44dpa ）

ODSS: 9/12Cr-2W steels （ JOYO irradiation ～ 15dpa, JAEA/Japan ）

RAFM Steel

0 5 10 15 20 25

0 200 400 600 800 1000 1200

Eurofer 97

15 dpa, T_test = T_irr

250°C 300°C 350°C 400°C 450°C

Str ess [MP a]

Strain [%]

 Substantial irradiation hardening

 Early strain localization due to

dislocation channeling  A

~0.3%

0 5 10 15 20 25

0 200 400 600 800 1000 1200

ODS-Eurofer HIP 15 dpa, T_test = T_irr

250°C 300°C 350°C 400°C 450°C

Stress [MPa]

Strain [%]

RAFM-ODS Steel

 Still work hardening  almost no loss of uniform elongation (A

~7%)

E. Materna-Morris et al. JNM, 2011

Hardening/Loss of Elongation

38

No-loss-of-Elongation

39 従来のフェライト鋼（RAFｓ）

伸びの低下を伴なう照射硬化

ODS鋼

伸びの低下を伴わない照射硬化

Ｓ1

Dislocation

source Slip plane

Irradiation defects Dislocation

Ｓ2

Slip plane

Ｓ1

Dislocation

source Slip plane

Irradiation defects Dislocation

Defect absorption

Ｓ2

Slip plane Defect

absorption

Oxide particle

Oxide particle

Ferrite/Martensite steels

Hardening and loss of elongation ODS steels

Hardening and no loss of elongation

1) Absorption of defects 2) Softening

3) Dislocation multiplication 4) Channel deformation 5) Accelerate necking 6) Loss of elongation

1) Absorption of defects

2) Dislocation pile-up at oxide particles 3) Inhibit the source 1

4) Activate another source 2 5) Homogeneous deformation 6) No loss of elongation

Ｓ3

ODSS-1 ODSS-2 Non-ODSS

Oxide particles suppress grain

growth and supply a number of grain

boundaries that absorb vacancies.

by Voyevodin (KIPT) etal.

-- steady growth rate in BCC --

-- remarkable effect of microstructures --

Resistance to Void Swelling (ion)

40

0 1 2 3 4 5 6 7

40 60 80 100 120 140 160

JLF-1 K3-ODS

Swelling (%)

Displacement damage (dpa)

Dual-ion irradiated at 773 K 15appm 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-diameter JLF-1-diameter

K3-density JLF-1-density

Mean Cavity Diameter (x10-9 m) Cavity Number Density (m -3)

Diaplacement damage (dpa)

Dual-ion irradiated at 773 K 15appm He/dpa, 1.0x10^-3dpa/sec

Dose dependence of the void swelling of JLF-1 reflects that of void size.

No bubble growth was observed for the ODS steel.

Swelling Tolerance in ODSS (2)

Dual ion beam, 60 dpa, (900appmHe), 500°C

41

K-ODS (16Cr-ODS) JLF-1 (9Cr-2W)

Dual Fe

+ 9 00 appmHe Si n g le 6.4 MeV Fe

50nm

-- Helium enhances void swelling --

-- ODS steel is highly resistant to void swelling –

Swelling Tolerance in ODSS (1)

500 °C, 60 dpa (DuET)

S = 0.18-0.65% S = 3.08%

42

K-ODS (16Cr-ODS) JLF-1 (9Cr-2W)

Dual Fe

+ 9 00 appmHe Si n g le 6.4 MeV Fe

50nm

-- Helium enhances void swelling --

-- ODS steel is highly resistant to void swelling –

Swelling Tolerance in ODSS (1)

500 °C, 60 dpa (DuET)

S = 0.18-0.65% S = 3.08%

20nm

43

● XRD 結果と酸化物粒子の結晶構造

ox id e m at rix

ox id e m at rix

Strain field at interface Trapping sites

Trapping Sites at Interphases

44

Fe-15Cr-4Al ODS steel with 0.3%Zr

0 0.2 0.4 0.6 0.8 1 1.2 1.4

-200 -150 -100 -50 0 50

Test Temperature / ^oC

Absorbed Energy / J .

F82H as-received F82H He implanted 9Cr-ODS as-received 9Cr-ODS He implanted 14Cr-ODS as-received 14Cr-ODS He implanted

F82H

9Cr-ODS 14Cr-ODS

0 0.2 0.4 0.6 0.8 1 1.2 1.4

-200 -150 -100 -50 0 50

Test Temperature / ^oC

Absorbed Energy / J .

F82H as-received F82H He implanted 9Cr-ODS as-received 9Cr-ODS He implanted 14Cr-ODS as-received 14Cr-ODS He implanted

F82H

9Cr-ODS 14Cr-ODS

Intergranular

Cleavage

100μ m fracture

fracture

Bottom of Notch He Implanted Area

Cleavage fracture

a) F82H b) 9Cr-ODS

Intergranular

Cleavage

100μ m fracture

fracture

Bottom of Notch He Implanted Area

Cleavage fracture

a) F82H b) 9Cr-ODS

50nm 50nm 50nm

c) TEM image of 9Cr-ODS

Bottom of Notch He-implanted Area

a) F82H b) 9Cr-ODSS

-- ODS steel is highly resistant to He-G.B embrittlement --

c) GB image of 9Cr-ODSS

Tolerance to He-G.B. Embrittlement

45

AVF Cyclotron (50MeV a-particle)

- 550

C, 1000appm, 0.4 dpa

RAFS/ODSS Comparison

Irradiation

Temp. (°C) ^{Steel 100 200 300 400 500 600 700}

Interstitial (I) RAFS mobile

ODSS mobile

I-cluster RAFS observed not observed

ODSS observed not observed

Vacancy (V) _{RAFS immobile mobile}

ODSS immobile mobile

V-cluster _{RAFS stable growth} evapolation

ODSS stable growth evapolation

Irradiation

Effects ^{RAFS HARDENING SWELLING RECOVERY}

ODSS HARDENING SWELLING

V-He cluster _{RAFS stable growth} evapolation

ODSS stable ^growth

Irradiation

Effects (He) ^{RAFS HARDENING SWELLING RECOVERY}

ODSS HARDENING SWELLING

46

6. Summary

1. Radiation effects are caused by radiation induced defect clusters which undergo the temperature dependent nucleation and growth processes.

2. The radiation effects depend on irradiation temperature which affects the mobility and stabilities of defects and consequently the type of defect

clusters.

3. Irradiation hardening could be due to I-loops that annihilate with vacancies after the decomposition of microvoids above ~420°C.

4. Some microvoids grow to show the peak of void selling at ~460°C.

5. Helium stabilizes vacancy clusters, which accelerates the nucleation and retards the growth of defect clusters.

6. Radiation tolerance is due to trapping point defects and helium atoms.

There is a trapping capacity limit for the other defects, such as dislocations and grain boundaries, and oxide particles.

47

* These are based on the experimental results of RAFS and ODSS.