Atomistic Simulation for the Development of Advanced Materials Brian D. Wirth*, with significant contributions from M.J. Alinger**, A. Arsenlis 1 , H.-J. Lee, P.R. Monasterio 2 G.R. Odette 3 , B. Sadigh 1 , J.-H. Shim 4 , and K. Wong Presented at GCEP - MIT Workshop on Nuclear Fission, Cambridge, MA, 29 Nov 2007 * [email protected]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 Center 1 2 4 3
18
Embed
Atomistic Simulation for the Development of Advanced …
This document is posted to help you gain knowledge. Please leave a comment to let me know what you think about it! Share it to your friends and learn new things together.
Transcript
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
* [email protected] 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
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).
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
• 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