consequences of irradiation on behavior of structural...

OAK RIDGE NATIONAL LABORATORYU. S. DEPARTMENT OF ENERGY

Consequences of Irradiation onBehavior of Structural Materials

Steven J. Zinkle

Materials Science & Technology Division, OakRidge National Laboratory

Materials for Generation IV Nuclear ReactorsNATO Advanced Study Institute

Cargese, Corsica, FranceSept. 24-Oct. 6, 2007


Outline•Effect of low temperature (<0.3 TM) irradiation on the mechanical

properties of austenitic and ferritic/martensitic steels• Dose and temperature dependence

•Deformation mechanism maps for irradiated materials•Overview of irradiation creep•High temperature He embrittlement of grain boundaries•Prospects for developing new high-performance structural materials

for Gen IV reactor applications


Overview of Gen IV and current (Gen II+)Structural Materials Environments

All Gen IV concepts introduce substantialincreases in dose and temperature

S.J. Zinkle ,OECD NEA Workshop on StructuralMaterials for Innovative Nuclear Energy Systems,

Karlsruhe, Germany, June 2007, in press


Room Temperature Radiation Hardening in 9Cr FMSteels

Tirr=300oC

Tirr=60-100oC

J. Rensman, 2005


Room Temperature Radiation Hardening in 9Cr FMSteels

J. Rensman, 2005

Tirr=60-100oC

Tirr=300oC

Low Temperature Radiation Hardening is Importantin Ferritic/martensitic steel 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)


Radiation hardening in Fe-(8-9%Cr) steels

0

200

400

600

800

1000

1200

0 100 200 300 400 500 600 700

Yie

ld S

tre

ng

th,

MP

a

Temperature, °C

Tirr=T

test

8-9Cr Steels: Yield Strength as Functionof Temperature, 0.1 - 94 dpa

unirradiated

0.3 TM


0

200

400

600

800

1000

1200

0 100 200 300 400 500 600 700

12CrYS/temp2

HT-9 controlfrom LiteratureUS/J Program

Yie

ld S

trength

, M

Pa

Temperature, °C

HT-9: Yield Strength as Functionof Temperature

Tirr

=Ttest

Ttest~Tirrad

Radiation embrittlement is of concern in HT-9 forirradiation temperatures up to ~400oC


Radiation hardening in Fe-(8-9%Cr) steels after 3 dpa

0

200

400

600

800

1000

1200

0 0.05 0.1 0.15 0.2 0.25 0.3 0.35

Engin

eering S

tress, M

Pa

Engineering Strain, mm/mm

Irrad., Ttest

=250˚C

Irrad., T

test=90˚C

F82HUnirradiated, T

test=250˚C

a)


(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)






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


Tensile curves for F82H ferritic/martensitic steelirradiated at 300˚C to 5 dpa

0

200

400

600

800

1000

0 5 10 15 20 25 30

Irr. 300°C

En

gin

ee

rin

g S

tre

ss

(M

Pa

)


Tested at RT with 1x10- 3/s



Unirr. Tested at RT

with 1x10- 3/sTested at 300°C

with 1x10- 3/s Unirr.Tested at 300°C

with 1x10- 3/s

N. Hashimoto et al., MRS vol. 650 (2001)

m = (1/σ)(∂σ/∂lnε)Strain hardening exponent @RT:m = 0.041 (unirradiated)m = 0.028 (irrad. 300˚C)


Irradiation of Austenitic Stainless Steel in Mixed SpectrumReactors produces significant hardening up to 500˚C, with

peak hardening at ~300˚C

0

200

400

600

800

1000

0 100 200 300 400 500

Yie

ld S

trength

, M

Pa

Temperature, °C

3 - 20 dpa

316 and PCA steels

Unirradiated

0.3 TM

Hardening at 290-400˚C is dominated by He bubbles


Austenitic Stainless Steel exhibits low uniform elongationafter irradiation at temperatures up to ~400˚C

Reduction in uniform elongation requires higher dosesthan in simple metals (e.g. Cu, Ni)

0.01

0.1

1

10

100

0.0001 0.001 0.01 0.1 1 10 100

literature data!100°Cliterature data 200-230°Cliterature data 250°Cliterature data 270-350°CJ.P. Robertson et al 60°CJ.P. Robertson et al 200°CJ.P. Robertson et al 330°C

Un

ifo

rm E

lon

ga

tio

n,

%

Neutron Dose, dpa


Plastic Instability Stress (σPI) of BCC Steels

• 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

670 MPa


Plastic Instability Stress (σPI) of austeniticstainless steel 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

Type 316 austenitic stainless steel


Deformation mode map for 316 SS neutron-irradiated at 65-100oC and tested at 25˚C

K. Farrell et al. ORNL/TM-2002/66 (2002)


Deformation Mechanism Map for UnirradiatedV-4Cr-4Ti at Strain Rates of 10-3-10-5 s-1:

Comparison of experimental data and revised calculated map

0.0 0.2 0.4 0.6 0.8 1.010

-6

10-5

10-4

10-3

10-2

10-1

YS,10-3s

-1

YS,10-4s

-1

YS,10-5s

-1

10-3s

-1

10-4s

-1

10-5s

-1

Dislocation creep

Dislocation glide

Theoretical shear stress limit

No

rmalized

Sh

ear

Str

ess,

!/µ

Normalized Temperature, T/TM

V-4Cr-4Ti, d"/dt=10-3-10

-5s

-1, L=20 µm

1410

141

14

1.4

Un

iaxia

l Ten

sile

Stre

ss (M

Pa)

0.14

Coble creep

Elastic regime

d"/dt<10-8s

-1

M. Li and S.J. Zinkle,J. ASTM Intern. 2(2005)12462


The Operating Window for BCC metals can be Dividedinto Four Regimes (red values are relevant for Nb1Zr)

I, II: Low Temperature Radiation Embrittlement Regimes– Fracture toughness (KJ) embrittlement: high radiation hardening causes low

resistance to crack propagation (occurs when SU>500-700 MPa)• Regimes which cause KJ<30 MPa-m1/2 should be avoided (Tirr< ~600 K ?)

– Loss of ductility: localized plastic deformation requires use of moreconservative engineering design rules for primary+secondary stress (Se)

III: Ductile Yield and Ultimate Tensile Strength Regime (eU>0.02)– Sets allowable stress at intermediate temperature (very small regime for Nb-1Zr)

IV: High Temperature Thermal Creep Regime (T>~1050 K)– Deformation limit depends on engineering application (common metrics are 1%

deformation and complete rupture)

!

Se

=

1

3Su

1

3Su+E("

U#0.02)

8

$

% & '

( )

*

+ , ,

- , ,

εU<0.02

εU>0.02

where εU is uniform elongation, SU is ultimate tensile strength, E is elastic modulus(additional design rules also need to be considered)

(Tirr< ~900-1270 K)


Stress-Temperature Design Window for Nb-1Zr

0

20

40

60

80

100

200 400 600 800 1000 1200 1400 1600

Design Window for Nb-1ZrD

es

ign

Str

es

s (

MP

a)

Temperature (K)

Sm (1/3 UTS)

St (10 5 h, 2/3

creep rupture !)radiation

embrittlement regime

(" t>1x10 20 n/cm2)

Space reactor temperature regime

Low Low ductility ductility regimeregime

((φφt>1x10t>1x102020 n/cm n/cm22))


Stress-Temperature Design Window forUnirradiated Type 316 Stainless Steel

0

50

100

150

200

300 400 500 600 700 800 900 1000 1100

De

sig

n S

tre

ss

(M

Pa

)

Temperature (K)

Sm (1/3 UTS)

St (105 h, 2/3

creep rupture !)

Smt (316LN SS)

Low Low ductility ductility regimeregime

(>1-5 dpa)(>1-5 dpa)


Can we break the shackles that limit conventionalstructural materials to ~300oC temperature window?

Structural Material Operating Temperature Windows: 10-50 dpa

Low temperature radiation embrittlement typicallyoccurs for damage levels ~0.1 dpa (0.01 MW-yr/m2)Zinkle and Ghoniem, Fusion Engr.

Des. 51-52 (2000) 55

Thermal creep

Radiationembrittlement

ηCarnot=1-Treject/Thigh


Can we break the shackles that limit conventionalstructural materials to ~300oC temperature window?

Structural Material Operating Temperature Windows: 10-50 dpa

Low temperature radiation embrittlement typicallyoccurs for damage levels ~0.1 dpa (0.01 MW-yr/m2)Zinkle and Ghoniem, Fusion Engr.

Des. 51-52 (2000) 55

ηCarnot=1-Treject/Thigh

ThermalcreepregimeRadiation

embrittlementregime

SCWR

VHTR


Consideration of Chemical Compatibility can Result inDramatic Reductions in Temperature Window

Estimated Structural Material Operating Temperature Windows:Moderately-pure He coolant, 10-50 dpa

Zinkle and Ghoniem, Fusion Engr.Des. 51-52 (2000) 55

Group V refractory alloys are particularly sensitive to embrittlement from O,C,N solute


Irradiation Creep of Austenitic Stainless Steel willGenerally be of Concern only for High Fluence (>20

dpa), High Stress Environments

0

0.5

1

1.5

2

0 100 200 300 400 500

25% CW, 19 dpaSA, 19 dpa25% CW, 7.4 dpaSA, 7.4 dpa

Eff

ective

Str

ain

, %

Effective Stress, MPa

PCA330°C


Calculated irradiated Ashby deformation mapfor Type 316 stainless steel at low strain rates

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

20˚C 300˚C 650˚C

24

240

Un

iax

ial

Te

ns

ile

Str

es

s,

MP

a

2.4

0.24

2400

10-10 s-1, L=50 µm

N-H creep

Twinning

Irrad. creep

Damage rate = 10-6 dpa/s


Effect of applied stress on the rupture lifetime ofaustenitic stainless steel annealed at 750˚C

• 100 appm He reduces rupture lifetime of stainless steel at 750˚C by 10-100X due toformation of grain boundary He cavities

10-1

100

101

102

103

104

40

50

60

70

80

90

10

0

20

0

unimplanted

pre-implanted, 300K

pre-implanted, 1023K

"in-beam"

Tim

e t

o r

up

ture

, h

Stress, MPa

Ttest

= 1023 K

cHe

= 100 appm

dcHe

/dt = 100 appm/h

Schroeder and Batflasky, 1983


Fine-scale matrix precipitates can trap He and providegreatly improved He embrittlement resistance

Yamamoto & Schroeder J.Nucl. Mat. 155-157 (1988)

Type 316stainless

steelImproved

stainless steelcontaining fine-

scale precipitates


Austentic steels are typically more sensitive tohelium embrittlement effects thanferritic/martensitic steels

Schroeder & StammASTM STP 1046 (1989) p.244


Cracking is observed in welded metals containingHe (from nuclear transmutations)

Heat affectedzone (HAZ)

Weld bead

Occurs in heat affected zone for low He concentrationsAt high He concentrations, cracking also occurs in the fusion zone (weld bead)

HAZ

Effect of transmutant He on welding behavior of metals

• Irradiated materials with He contents above ~1 appm cannot befusion-welded due to cracking associated with He bubble growth;the lower temperatures associated with FSW may allow repairjoining of irradiated materials

0

0.2

0.4

0.6

0.8

1

1.2

1.4

0 0.5 1 1.5 2 2.5 3

Calculated size of He bubbles at grain boundaries in 316 SS

Dia

me

ter

(µm

)

Time (s)

1400 K

1500 K1600 K

Threshold He bubble size for cracking in 316 SS


Irradiation of Austenitic Stainless Steel in Mixed SpectrumReactors causes Pronounced Loss in Elongation and

Significant Reduction in Fracture Toughness

0

10

20

30

40

50

0

100

200

300

400

500

0 100 200 300 400 500

Un

ifo

rm e

lon

ga

tio

n (

%)

Fra

ctu

re T

ou

gh

ne

ss

KJ

Q (

MP

a-m

1/2

)

Test Temperature (˚C)

Type 316 SS, ~5 dpaHFIR & HFR data

eU,

irradiated

eU, unirr.

KJQ

, unirr.

KJQ

,

irradiated


Fracture Toughness of Irradiated BCC Structural Alloys• There are two general approaches to mitigate DBTT increases

•Reduce radiation hardening by alloying modifications (e.g., low-Cu RPV steels)•Increase σ*

• Significant improvements in resistance to lowtemperature radiation embrittlement can beachieved by selective alloying (e.g., reduced Cu inreactor pressure vessel steels)

0

200

400

600

800

1000

-200 0 200 400 600

Yie

ld S

tre

ng

th,

MP

a

Temperature, °C

unirradiated

irradiated

!*/M

300300

Cha

rpy

Ener

gy (J

)

Submerged-Arc WeldsIrradiation: 1 x 1023 n/m2, 288°CNormalized To Unirradiated Curve

Unirradiated

TransitionRegion

Shift

0.06% CuIrradiated

Loss

0.30% CuIrradiated

Upper Shelf150150

100100

5050

00-200-200 -100-100 00 100100 200200Temperature (°C)Temperature (°C)


Comparison of the fracture toughness of 9Cr and12Cr steels

0

50

100

150

200

250

300

350

400

0 100 200 300 400

ORNL UnirradECN UnirradUCSB UnirradORNL Irr HT-9ORNL Irr F82-HECN Irr HT-9

Fra

ctu

re T

oughness, M

Pa!m

Test Temperature, °C

UnirradiatedHT-9, F82-H

HT-9T

irr=80-90°C

HT-9T

irr=250-300°C

F82HT

irr=250°C

c)

A.F. Rowcliffe et al., J. Nucl. Mater. 258-263

Fusion Reduced Activation Fusion Reduced Activation Ferritic/martensiticFerritic/martensiticSteels Exhibit Superior Irradiation PerformanceSteels Exhibit Superior Irradiation Performance

Compared to Conventional 9Cr-1Mo and 12Cr SteelsCompared to Conventional 9Cr-1Mo and 12Cr Steels● Suppression of irradiation embrittlement

(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(f) ATR/543K, 2.2dpa: RT

0

50

100

150

200

250

300

-200 -100 0 100 200 300 400 500

9Cr-2W

9Cr-1Mo 12Cr/2.25Cr-2W

Shift

in D

BTT

/ K

Irradiation Hardening / MPa

(1)(2)

(3)

(b)

(4)

(5)

(a)

(c)

(d)

(e)

(b)

580appm He120appm He

9Cr-1Mo

9Cr-2W

0 10 20 30 40 50dpa

Ni doped 9Cr-2W

w/o Ni

(f)

(f)

A. Kohyama, ANS TOFE (Madison, WI, 2004)


Effect of Neutron Irradiation on DBTT ofFerritic/martensitic Steels


8-9%Cr RAFM Steels Exhibit LessEmbrittlement Per Unit of Hardening ThanLow-Alloyed RPV Steels

!To = 0.3!"

y RAFM

!To = 0.7!"

y RPV

YIELD STRENGTH INCREASE, MPa

0 100 200 300 400 500 600

!T

o

SH

IFT

, o

C

0

50

100

150

200

250

F82H-IEA

F82H-HT2

9Cr-2WVTa

Eurofer97(Lucon, SCK-CEN)

Eurofer97(Rensman, NRG)

RPV Steels

(Sokolov, ASTM STP 1325)

M.A. Sokolov, ICFRM12, J. Nucl. Mater.367-370 (2007)


Effect of Cr Content on the Ductile-BrittleTransition Temperature of IrradiatedFerritic/martensitic Steels

Klueh and Harries 2001Caution: thermomechanical treatments not optimized!


( )df!"

µ#$

%&=

1

1206.3

2002

2

2

n

PSD

MGssr

DDkT

bA

Ct

!!"

#$$%

& '

()

(*+

(,

(-.

!!"

#$$%

& '+=

=

µ

//

µ

//µ0

0

&

&

( ) )(3.06 lattice Tirr

ysobstaclef !""! #++=

10

)(TEideal

!"

2

r

dr22

a

r

4

2

00!

""#

$%%&

'+

()

=

r

rD

kTat

gb

r

*+,

M. Li and S. Zinkle


Ni Base Alloys Have Excellent HighTemperature Strength Properties But AreHighly Susceptible to Ground BoundaryEmbrittlement

Tensile properties of Nimonic PE-16 irradiated at 500 C to ~6 dpa). Heat treatment: 4 h at 1040C + 1 hr at 900 C + 8 hr at 750 C.(Flags signify 2 h @ 1040 C + 16 hr at 700 C.)

Fracture is transgranular

Fracture is Highly Intergranular


a b

c d

200nm

Inconel 718 solution annealed at 1065C/30min and aged 750C/10h+ 650C/20h: (a) pristine,and 3500 keV Fe-ion irradiated at 200 C to (b) 0.1 dpa, (c) 1 dpa, and (d) 10 dpa.

Precipitates in Inconel 718 dissolve duringirradiation

a b

c d


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


Commercial, Standard CF8C

(TEM, As Cast)

New CAT/ORNL CF8C-PLUS

Creep Tested 850°C/23,000 h

Microstructural Evolution In Irradiated Stainless Steels Providedthe Key for Developing Improved High Temperature Alloys

Result of microstructural modification:• Formation of stable nanoscale MC carbide dispersions to pin

dislocations• Resistance to creep cavitation and embrittling grain boundary

phases (ie. sigma, Laves)• Resistance to dislocation recovery/ recrystallization

• Reactant Effects (ie. Ti, V+Nb enhance MCformation)

• Catalytic Effects (ie. Si enhances Fe2Mo or M6C)• Inhibitor Effects (ie. C, P or B retard the formation

of Fe2Mo or FeCr sigma phase during aging, G-phaseduring irradiation)

• Interference Effects (ie. N forms TiN instead of TiC;N does not form NbN instead of NbC; therefore Cand N can be added with Nb, but not Ti)

Creep Tested 850°C/500 h

2003 R&D 100 Award

FreedomCAR and Vehicle Technologies and DistributedEnergy and Electricity Reliability Programs of DOE-EERE P.J. Maziasz


Technology Transfer of CF8C-Plus Cast StainlessSteel

• MetalTek International, Stainless Foundry &Engineering, and Wollaston Alloys have triallicenses in 2005

• MetalTek has cast over 31,000 lb of CF8C-Plussteel through 2005 for Solar Turbines (end-cover,casings), Siemens-Westinghouse (large sectiontests for turbine casings), ORNL, and a globalpetrochemical company (tubes/piping).

• Stainless Foundry has cast CF8C-Plus exhaustcomponents for Waukesha Engine Dresser NGengines

6,700 lb CF8C-Plus end-cover cast by MetalTek forSolar Turbines Mercury 50gas turbine

80 lb CF8C-Plus exhaustcomponent cast byStainless Foundry forWaukesha NGreciprocating engine


Development of New Alumina-Forming,Creep Resistant Austenitic Stainless Steel

800oC in air, 72hY. Yamamoto, M.P. Brady et al., Science 316 (2007) 433


Creep Properties of New 3 Cr Steels have Advantages OverExisting 2 1/4Cr and 9-12Cr Steels

• Creep resistance improved over Japanese2.25Cr (T23,T24) steels•May not need to be tempered for high thermalcreep strength; no postweld heat treatment?•Properties better than HT9 and as good orbetter than modified 9Cr-1Mo steel

•Long-term creep behavior (>5000 h) stillneeds to be determined•Two 50 ton heats procured by industry arebeing tested to obtain ASME code approval•Improved creep resistance is due to fine V-rich MCneedles formed during 1100˚C normalizing

MCneedles

50 nm

2 1/4 Cr-1Mo New 2 1/4 Cr-2WV New 3 Cr-3WV

R.L. Klueh et al.


New Thermomechanical treatment (TMT) ProcessApplied to Commercial 9-12%Cr Steels Yields ImprovedMicrostructure

• Commercial 9Cr-1Mo and 12 Cr steelswere processed

• TMT (hot rolling) on 25.4–mm plates– Several TMT conditions were investigated

• Precipitates formed on dislocationsintroduced by hot rolling

• Precipitate dispersion is much finer thanobserved in conventionally processed 9-12Cr steel

Modified 9Cr-1Mo—New TMTR.L. Klueh et al., J. Nucl. Mater.367-370 (2007) 48


Large Increase in Rupture Life of Modified 9Cr-1Mo at 650°C

•Thermo-mechanical treatment (TMT) of modified 9Cr-1Mo produced steel with over an order-of-magnitudeincrease in rupture life

R.L. Klueh et al., J. Nucl. Mater.367-370 (2007) 48


Oxide dispersion strengthened Steels• There are two main options for ODS steels, based on pioneering work

by Ukai and coworkers– 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


New 12YWT Nanocomposited Ferritic Steel has SuperiorStrength compared to conventional ODS steels

O Y Ti10 nm

• Atom Probe reveals nanoscale clusters to be source ofsuperior 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 to thermally stablenanoscale (Ti,Y,Cr,O) particles during processing

• Nanoclusters not present in ODS Fe-13Cr + 0.25Y2O3alloy

0

200

400

600

800

1000

1200

400 600 800 1000 1200

ODS (SUMITOMO)

9Cr-2WVTa STEEL

12YWT

0DS (Al2O3)

ODS(TiO2)

ODS(MgO)

ODS (ZrO2)

MA956

Temperature K

12YWT

MA956

ODS (SUMITOMO)

9Cr-2WVTa

ODS(ZrO2)

ODS(TiO2)

ODS(MgO)ODS(Al

2O

3)

Stre

ss/

MP

a

The tensile strength of the 12YWT ODS ferritic alloy compared with the UTS of similar steels.

Yield strength

• Thermal creep time to failure is increased by several ordersof magnitude at 800˚C compared to ferritic/martensiticsteels

– 2% deformation after ~2 years at 800oC, 140MPa• Potential for increasing the upper operating temperature of

iron based alloys by ~200°C• Acceptable fracture toughness near room temperature

D.T. Hoelzer et al.


Theory Has Shown That Vacancies Play a PivotalRole in the Formation and Stability of Nanoclusters

The presence of Ti and vacancyhas a drastic effect in loweringthe binding energy of oxygen inFe (by C. L. Fu and M. Krcmar).


Conclusions• Existing structural materials face Gen IV reactor design

challenges due to limited operating temperature windows– May produce technically viable design, but not with desired optimal

economic attractiveness• Substantial improvement in the performance of structural

materials can be achieved in a timely manner with a science-based approach

• Design of nanoscale features in structural materials confersimproved mechanical strength and radiation resistance– Such nanoscale alloy tailoring is vital for development of radiation-

resistant structural materials for advanced fission reactors• Continued utilization of modern materials science techniques

would be valuable to uncover new phenomena associated withlocalized corrosion, ultra-high toughness ceramic composites,ultra-high strength alloys, etc.


Modern computational thermodynamics revealspathway to improve precipitation hardened stainlesssteels

• Within alloy specifications, large differences can be expected with standard heattreatment

• Calculations can lead to composition and heat treatment optimization

“average” Cr, Mo, Ni

J.M. Vitek

Poor Good

Both 15%Cr-7%Ni alloys are within allowable chemical composition

“low” Cr, Mo and “high” Ni

consequences of irradiation on behavior of structural...

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