Atomistic Simulation for the Development of Advanced Materials

Transcription

1 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 ** GE Global Research Center 4 * 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.

2 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

3 Irradiation effects on structural materials Exposure to neutrons degrades the mechanical performance of structural materials and impacts the economics 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 ( Tm) - High temperature He embrittlement (> 0.5 Tm); Specific to fusion & spallation accelerators Effect of neutron irradiation on the uniform elongation of bainitic and ferritic/martensitic steels 10 8 Uniform Elongation (%) V-4Ti-4Cr Fe-3Cr-3WV Fe-9Cr-1MoVNb Fe-9Cr-1MoVNb-2Ni Fe-9Cr-2WV Fe-9Cr-2WVTa Variables Tirr~70 C; Ttest~25 C Byun & Farrell (2004) Dose (dpa) Bond, Sencer, Garner, Hamilton, Allen and Porter, Stainless steel irradiated in EBR-II, 380 C, ~22 dpa, 1% swelling 1 10 Materials (Fe-based steels, Vanadium and Ni-based alloys, Refractory metals & alloys, SiC) and composition!initial microstructure (cold-worked, annealed)!irradiation temperature!chemical environment & thermalmechanical loading!neutron flux, fluence and energy spectrum - materials test reactor irradiations typically at accelerations of Synergistic Interactions

4 Materials behavior is inherently multiscale 35m(115ft) Reactor Cavity Cooling System Refueling Floor Control Rod Drive Stand Pipes Generator Reactor Pressure Vessel Cross Vessel (Contains Hot & Cold Duct) 46m(151ft) Power Conversion System Vessel Shutdown Cooling System Piping Floors Typical 32m(105ft) Radiation Radiationdamage damageproduces producesatomic atomicdefects defectsand andtransmutants transmutantsat atthe theshortest shortesttime timeand and length lengthscales, scales,which whichevolve evolveover overlonger longerscales scalesto toproduce producechanges changesin inmicrostructure microstructure and andproperties propertiesthrough throughhierarchical hierarchicaland andinherently inherentlymultiscale multiscaleprocesses processes

5 Multiscale modeling approach Approach: apply multiple complementary modeling, experimental and theoretical techniques at at appropriate scales to to determine underlying mechanisms

6 Exposure to neutrons embrittles pressure vessel steels, manifested by transition temperature increases (!T) and upper shelf energy decreases (!USE) 100 Ductile RPV embrittlement Copper Rich Precipitates Atom Probe* Cu Ni 75 Unirradiated!USE E cvn (J) 50 25!T Irradiated 0 Brittle T ( C) Low dose: <0.1 dpa over years 1 nm Cu Si Ni Mn Matrix Features (vacancy - solute clusters) Positron annihilation 60 C 288 C " (ps) - 60 C " (ps) C Mn Si * Mike Miller, ORNL 3 vac/6 Cu 7 vac/10 Cu 1 nm vacancy Cu 10 vac/4 Cu Objective: Develop a model predicting the evolution of of both CRPs and Matrix Features to to predict dependence on on composition, dose rate & temperature

7 Damage accumulation, 290 C, dpa/sec Fe - 0.3% Cu Temperature = 290 C Dose Rate = dpa/s 5 nm vacancy "clustered# Cu Monasterio, Wirth and Odette, J. Nuc. Mater. 361, 127 (2007).

8 Key features observed Transient sub-nm vacancy - Cu clusters 1 nm 2.3 mdpa, 290 C, dpa/s Growing nm Cu clusters/precipitates 2 nm vacancy "clustered# Cu 3 vac/6 Cu vacancy Cu 4 vac/9 Cu 9.9 mdpa, 290 C, dpa/s 5 vac/9 Cu 7 vac/10 Cu 10 vac/4 Cu

9 Irradiation hardening and ductility loss Radiation damage produces atomic defects, which drive microstructure and macroscopic property changes. Shear stress [MPa] a Proton irradiated single crystalline Cu Unirradiated Increasing irradiation dose b Shear strain Dislocation-obstacle interaction mechanism Evolution of microstructure and localized deformation 500nm Ref. (a) M. Victoria et al. J. Nucl. Mat. 276, 114 (2000) (b) R. Schaublin et al, Journal of Nuclear Materials, 276 p (2000) (c) Z. Yao et al, J. Nucl. Mat. 329, 1127 (2004)

10 Screw dislocation-sft interaction in FCC Cu 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 22.5nm # xy # xy Applied shear stress=0,100,300 MPa Mishin EAM potential 44.3nm 31.4nm z=_[i_2] y=_[ii0] x=[_11]

11 Screw dislocation-sft interaction Is complete absorption of an SFT by a screw dislocation possible? D A B MD simulation at T=100K, no applied stress. SFT size=4.6nm (153 vacancies) SFT size=2.5nm Final helical dislocation proposed by Kimura & Maddin SFT size=8.5nm

12 Screw dislocation-sft interaction in FCC Cu Snapshots of SFT and screw dislocation interaction process at " xy =300MPa Remaining structure immediately after the interaction Lee, Shim and Wirth, J. Materials Research. 22, 2758 (2007).

13 Comparison to in-situ TEM results Screw dislocation Deformation of Au at room temperature Edge dislocation Mixed dislocation Ref) Y. Matsukawa et al., Journal of Nuclear Materials (2004) 919. Lee, Shim and Wirth, J. Materials Research. 22, 2758 (2007).

14 Isotropic plasticity model for irradiated metals Tension Experiments* Full FEM Tension Simulations Increasing defect cluster density B. N. Singh et al, J. Nucl. Mat. 224, 131 (1995). Plastic instability in tension geometry leads to flow localization and failure 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 Arsenlis, Wirth and Rhee, Phil. Mag. 84, 3617 (2004).

15 Extreme environments in Advanced Nuclear Energy Systems * (ABR) * S.J. Zinkle, ORNL Very High Temperature Reactor Super-Critical Water Reactor Gas-cooled Fast Reactor Lead-cooled Fast Reactor Sodium-cooled Fast Reactor Molten Salt Reactor

16 Are radiation resistance materials possible? * * L.K. Mansur, E.H. Lee, ORNL Swelling can be greatly reduced by dispersing fine(nm-) scale precipitates

17 vacancy Structural materials future: Advanced ferritic alloys SIA APT-Miller ORNL Y-Ti-O (-Fe) NF voids Cr Y Ti O NCF!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) Use high sink strength of nano-features to trap (getter) both He (in fine bubbles) and vacancies (to enhance self-healing of damage by recombination with SIA) 10 nm NF B He n L n v i GB He bubbles NF He bubble SIA recombines n Dislocation v GB FMS trapped vacancy

18 Summary & Future Challenges!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