ASTM Conference, May 17 2016, Hilton Head Island, SC Understanding Irradiation Growth through Atomistic Simulations: Defect Diffusion and Clustering in Alpha-Zirconium and the Influence of Alloying Elements M. Christensen, W. Wolf , C. Freeman, E. Wimmer, Materials Design Inc., Santa Fe, NM, USA R. B. Adamson, Zircology Plus, Freemont, CA, USA L. Hallstadius, Westinghouse Electric Sweden, Västerås, Sweden P. Cantonwine, Global Nuclear Fuels, Wilmington, NC, USA E. V. Mader, Electric Power Research Institute (EPRI), Palo Alto, CA, USA
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ASTM Conference, May 17 2016, Hilton Head Island, SC
Understanding Irradiation Growth through Atomistic
Simulations:
Defect Diffusion and Clustering in Alpha-Zirconium and
the Influence of Alloying Elements
M. Christensen, W. Wolf, C. Freeman, E. Wimmer,
Materials Design Inc., Santa Fe, NM, USA
R. B. Adamson, Zircology Plus, Freemont, CA, USA
L. Hallstadius, Westinghouse Electric Sweden, Västerås, Sweden
P. Cantonwine, Global Nuclear Fuels, Wilmington, NC, USA
E. V. Mader, Electric Power Research Institute (EPRI), Palo Alto, CA, USA
Overview
EPRI Channel Distortion Program
Simulation methods, Zr-Fe forcefield development
Point defect formation: vacancies, interstitials, H
Point defect diffusion: vacancy, SIA, H, O, alloying elements
Cluster formation in pure Zr
Defect cluster formation in Zr involving Fe atoms
Effect of Fe on Zr mobility and irradiation growth
Investigation on secondary phase particles (SPPs)
Summary on effects of Fe
17 May 2016 2ASTM Conference, Hilton Head Island, SC – Understanding Irradiation Growth through Atomistic Simulations,
Materials Design, Inc.
Overview on Computation Methods
17 May 2016 3ASTM Conference, Hilton Head Island, SC – Understanding Irradiation Growth through Atomistic Simulations,
Materials Design, Inc.
MedeA® computational environment
Structural models with periodic boundary conditions (~100 atoms for
ab-initio, ~10000 atoms for forcefields)
Embedded atom potentials applied for large time- and length-scale
formation, structure and energy of dislocation loops, diffusivities
MedeA’s Forcefield Optimizer for fitting forcefields to ab-initio data
Ab-initio calculations (VASP) for geometries, total energy and
formation energy for stable structures and transition states
Ab-initio calculations (VASP) for elastic moduli
Ab-initio phonon calculations for temperature effects within a
(quasi)harmonic approximation
Eyring’s transition state theory for diffusivities
Zr-Fe Forcefield: Development and
Application
17 May 2016 4ASTM Conference, Hilton Head Island, SC – Understanding Irradiation Growth through Atomistic Simulations,
Materials Design, Inc.
Previously applied forcefields and references publishing the details:
• Zr-H system: ASTM STP 1543, 2015, pp. 55-92
• Zr - Fe, Cr, Ni, Nb, Sn: J. Nucl. Mater., Vol. 445, 2014, pp. 241-250.
Objective of current forcefield development for Zr-alloying element systems:
improve the accuracy of the Zr-Fe interactions in simulations with higher Fe
concentration, expected to be important for clustering
Ab-initio trainings set for interactions:
• Zr-Zr: MD-trajectories of interstitials and vacancies in Zr, hcp, fcc, and bcc Zr
for a set of different volumes
• Fe-Fe: bcc Fe at different volumes
• Zr-Fe: MD-trajectories of Fe interstitials in Zr, and Zr2Fe and Zr3Fe for a range
of different volumes
Validation: bulk densities (within 1%), elastic moduli (within a few percent)
Simulations: supercell of Zr with 11520 atoms and Fe (0.05%, 0.15%, or 0.5%).
Effect of irradiation is simulated by simultaneously inserting both vacancies and
Zr interstitials (0.05%, 0.1%, or 0.5%) in random positions followed by molecular
dynamics. Repeat this procedure 5 times and finally optimize the geometry.
Point Defects
Vacancy formation: +193 kJ/mol; 1% vacancies causes shrinkage by
0.25% in the a-direction and 0.43% in the c-direction
Self-interstitial formation: +286 at basal octahedral, +288 at octahedral, +304 at split and +321 kJ/mol at crowdion sites
H slightly prefers tetrahedral sites by a few kJ/mol, increasing with temperature, at operating temperatures about 94% of H atoms occupy the tetrahedral sites
H absorption isotherms and solubility were calculated from DFT, H solubility increasing under tensile strain (preferred hydride precipitation at crack tips)
Increased lattice dimensions and bulk moduli with H contents quantified
H is attracted by vacancies (up to 9 H atoms) and SIAs, increasing solubility, increased H solubility in irradiated Zr, recombination may cause supersaturation and provide nucleation sites for hydride formation
Alloying elements (Cr, Ni, Nb, Sn) prefer substitution (O interstitials) and (except Nb) tend to segregate to surfaces, grain boundaries, vacancies
17 May 2016 5ASTM Conference, Hilton Head Island, SC – Understanding Irradiation Growth through Atomistic Simulations,
Materials Design, Inc.
Point Defect Diffusion
Self-interstitial diffusion: fast and anisotropic (a>c)