Atomistic Simulation for the

Development of Advanced Materials

Brian D. Wirth*, with significant contributions from

M.J. Alinger**, A. Arsenlis1, H.-J. Lee, P.R. Monasterio2

G.R. Odette3, B. Sadigh1, J.-H. Shim4, and K. Wong

Presented at

GCEP - MIT Workshop on Nuclear Fission,

Cambridge, MA,

29 Nov 2007

* bdwirth@nuc.berkeley.edu This work was partially supported by the the US Nuclear Regulatory Commission, the U.S.

Department of Energy, Office of Nuclear Energy, Science and Technology and Office of Fusion

Energy Sciences, and partially performed under the auspices of the U.S. Department of Energy

and Lawrence Livermore National Laboratory under contract No. W-7405-Eng-48.

** GE Global

Research Center1

2

4

3

Presentation overview

•!Motivation: Materials challenges associated with current & future

fission power plants and radiation damage processes (covered by

Zinkle)

•!Science-based multiscale approach to understanding radiation

effects in structural materials

• Cascade aging and Irradiation response of Fe-based alloys

- Formation of Cu Rich Precipitates & vacancy - Cu clusters

- Radiation induced segregation of Cr

• Impact of microstructure on mechanical properties & performance

- Dislocation - defect interactions

- Constitutive & mechanical property modeling

•!Promise of radiation resistant materials

•!Summary & Future directions

• Exposure to neutrons degrades the mechanical performance of structural materials and impacts theeconomics and safety of current & future fission power plants:

- Irradiation hardening and embrittlement/decreased uniform elongation (< 0.4 Tm)

- Irradiation (<0.45 Tm) and thermal (>~0.45 Tm) creep

- Volumetric swelling (0.3 - 0.6 Tm)

- High temperature He embrittlement (> 0.5 Tm); Specific to fusion & spallation accelerators

Irradiation effects on structural materials

Variables• Materials (Fe-based steels, Vanadium

and Ni-based alloys, Refractory metals& alloys, SiC) and composition

•!Initial microstructure (cold-worked, annealed)•!Irradiation temperature•!Chemical environment & thermal- mechanical loading

•!Neutron flux, fluence and energy spectrum - materials test reactor irradiations typically at accelerations of 102 - 104

Synergistic Interactions

0

2

4

6

8

10

0.0001 0.001 0.01 0.1 1 10

Effect of neutron irradiation on the uniform elongation

of bainitic and ferritic/martensitic steels

Fe-3Cr-3WVFe-9Cr-1MoVNbFe-9Cr-1MoVNb-2NiFe-9Cr-2WVFe-9Cr-2WVTa

Un

ifo

rm E

lon

gat

ion

(%

)

Dose (dpa)

Tirr~70 C; Ttest~25 C

Byun & Farrell (2004)

0

V-4Ti-4Cr

Bond, Sencer, Garner, Hamilton,

Allen and Porter, 2001

304 Stainless steel irradiated

in EBR-II, 380°C, ~22 dpa,

1% swelling

Materials behavior is inherently multiscale

Radiation damage produces atomic defects and transmutants at the shortest time andlength scales, which evolve over longer scales to produce changes in microstructure

and properties through hierarchical and inherently multiscale processes

Radiation damage produces atomic defects and transmutants at the shortest time andlength scales, which evolve over longer scales to produce changes in microstructure

and properties through hierarchical and inherently multiscale processes

Reactor Cavity

Cooling System

Reactor Pressure

Vessel

Control Rod Drive

Stand Pipes

Power Conversion

System Vessel

FloorsTypical

Generator

RefuelingFloor

Shutdown Cooling

System Piping

Cross Vessel

(Contains Hot &

Cold Duct)

35m(115ft)

32m(105ft)

46m(151ft)

Multiscale modeling approach

Approach: apply multiple complementary modeling, experimental and theoreticaltechniques at appropriate scales to determine underlying mechanisms

Approach: apply multiple complementary modeling, experimental and theoreticaltechniques at appropriate scales to determine underlying mechanisms

• Exposure to neutrons embrittles pressure vessel steels, manifested by transition

temperature increases (!T) and upper shelf energy decreases (!USE)

0

25

50

75

100

Ecv

n (

J)

T (°C)

Unirradiated

!T

Irradiated

!USE

-100 0 100 200

Brittle

Ductile

Objective: Develop a model predicting the evolution of both CRPs and MatrixFeatures to predict dependence on composition, dose rate & temperature

Objective: Develop a model predicting the evolution of both CRPs and MatrixFeatures to predict dependence on composition, dose rate & temperature

RPV embrittlement

Copper Rich Precipitates

1 nm

Cu Si Ni Mn

Cu Ni

Mn Si

Atom Probe*

* Mike Miller, ORNL

Matrix Features (vacancy -

solute clusters)

7 vac/10 Cu

3 vac/6 Cu

10 vac/4 Cu

1 nm

vacancyCu

" (ps) - 288°C

222

520

" (ps) - 60°C

178

355

288 °C

60 °C

Positron annihilation

Low dose: <0.1 dpa

over 40-60 years

Damage accumulation, 290°C, 10-11 dpa/sec

vacancy

"clustered# Cu

5 nm

Fe - 0.3% Cu

Temperature = 290 C

Dose Rate = 10-11

dpa/s

Monasterio, Wirth and Odette, J. Nuc. Mater. 361, 127 (2007).

Key features observed

vacancy

"clustered# Cu

4 vac/9 Cu

7 vac/10 Cu

1 nm

vacancyCu

Transient sub-nmvacancy - Cu

clusters

Growing nm Cuclusters/precipitates

10 vac/4 Cu

3 vac/6 Cu

5 vac/9 Cu

2 nm

2.3 mdpa,

290°C, 10-11 dpa/s

9.9 mdpa,

290°C, 10-11 dpa/s

Irradiation hardening and ductility loss

Shear strain

Shear

str

ess [

MP

a] Proton irradiated single crystalline Cu

Unirradiated

Increasing

irradiation dose

Ref. (a) M. Victoria et al. J. Nucl. Mat. 276, 114 (2000) (b) R. Schaublin et al, Journal of Nuclear

Materials, 276 p251-257 (2000) (c) Z. Yao et al, J. Nucl. Mat. 329, 1127 (2004)

500nm

b

• Radiation damage produces atomic defects, which drive microstructure and

macroscopic property changes.

a

•Dislocation-obstacle interaction mechanism

•Evolution of microstructure and localized deformation

Screw dislocation-SFT interaction in FCC Cu

x=[_11]

z=_[I_2]

y=_[II0]

#xy#xy

•Visualization by atoms with hcp and neither

fcc/hcp structure (Common Neighbor Analysis)

•SFT size: 2.3nm and

4.6nm (45 and 153

vacancies)

•T=100K

•Applied shear

stress=0,100,300 MPa

•Mishin EAM potential

31.4nm

44.3nm

22.5nm

Screw dislocation-SFT interaction

A

B

D

MD simulation at T=100K,

no applied stress. SFT

size=4.6nm (153 vacancies)

SFT size=2.5nm

Is complete absorption of an SFT by a screw dislocation possible?

Final helical dislocation

proposed by Kimura & Maddin

SFT

size=8.5nm

Screw dislocation-SFT interaction in FCC Cu

•Snapshots of SFT and screw dislocationinteraction process at "xy=300MPa

•Remaining structure

immediately after the interaction

Lee, Shim and Wirth, J. Materials Research. 22, 2758 (2007).

Ref) Y. Matsukawa et al., Journal of Nuclear Materials

329-333 (2004) 919.

Comparison to in-situ TEM results

Edge dislocation

Screw dislocation

Mixed dislocation

Deformation of

Au at room

temperature

Lee, Shim and Wirth, J. Materials Research. 22, 2758 (2007).

B. N. Singh et al,

J. Nucl. Mat. 224, 131 (1995).

Tension Experiments* Full FEM Tension Simulations

•Isotropic polycrystal plasticity incorporates coarse grained scaling laws governing

dislocation density evolution and interactions determined for single crystals

•Dislocation - (radiation damage) defect interactions included based on MD simulations

•Resulting models can be further modified to include the effects of dispersed particles,

solute atoms, and other known resistance mechanisms

Isotropic plasticity model for irradiated metals

Increasing defect

cluster density

Plastic instability in tension

geometry leads to flow

localization and failure

Arsenlis, Wirth and Rhee, Phil. Mag. 84, 3617 (2004).

Extreme environments in Advanced Nuclear Energy Systems

* S.J. Zinkle, ORNL

*

Very High Temperature Reactor Lead-cooled Fast ReactorSuper-Critical Water Reactor Sodium-cooled Fast ReactorGas-cooled Fast Reactor Molten Salt Reactor

(ABR)

Are radiation resistance materials possible?

*

* L.K. Mansur, E.H. Lee, ORNL

Swelling can be greatly reduced by dispersingfine(nm-) scale precipitates

He

n

n

i

v

n

NF

v

NCFB

L

FMSDislocation

GB

NFHe bubble

SIA recombines

trapped vacancy

SIAvacancy

voids

GB He

bubbles

Structural materials future: Advanced ferritic alloys

•!Advanced Nano-Composited Ferritic

(NCF) steels under development at

ORNL/UCSB exhibit a range

of excellent mechanical properties

- high creep strength

- corrosion resistance

- toughness

derived from a high number of

< 3nm Y-Ti-O nano-features (NF)

Cr Y Ti O

10 nm

APT-Miller

ORNL

Y-Ti-O (-Fe) NF

• Use high sink strength of nano-features to trap (getter) both He (infine bubbles) and vacancies (to enhance self-healing of damage byrecombination with SIA)

•!Current and advanced future nuclear technologies require advanced

materials to withstand incredibly harsh environments

•!Radiation damage involves inherently multiscale phenomena -

fundamental understanding of radiation damage requires multiscale

modeling, closely coupled with theory & experiments

•!Examples of modeling radiation effects in structural alloys at T <~450°C

- Cu precipitate & sub-nm Vac-Cu cluster formation in RPV steels

- New understanding of dislocation - SFT interactions that agree with

in-situ TEM, leading to new understanding and models of irradiation hardening and flow localization

• Future challenges -> development of modeling techniques (Monte Carlo,

phase field, rate theory) that incorporates sufficient physics with ability

to reach meaningful doses required to simulate microstructural evolution,

incorporate multi-elemental chemistry and microstructural complexity,

determine range of microstructural characteristics that can offer

radiation resistance in a thermally stable material with a good balance

of properties

•!Scientific, multiscale approach will impact materials development, but

modeling from atoms-up by itself is not likely in the near-term

Summary & Future Challenges