Homogeneous hydride formation path in α-Zr: Molecular dynamics simulations with the charge-optimized many-body potential

Zhang, Yongfeng; Bai, Xian-Ming; Yu, Jianguo; Tonks, Michael; Noordhoek, Mark J.; Phillpot, Simon R.

doi:10.1016/j.actamat.2016.03.079

Cited by 49 publications

(24 citation statements)

References 39 publications

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“…These metastabe phases are believed to be important for the nucleation process of hydrides. In line with a previously suggested nucleation mechanism [162], recent MD simulations demonstrated that the γ hydride could form by shearing the ζ hydride [163]. A negligible barrier was identified for the shear, in agreement with previous DFT calculations [150].…”

Section: Hydrogen Transport and Hydride Formationsupporting

confidence: 89%

Mechanistic materials modeling for nuclear fuel performance

Tonks

Andersson

Phillpot

et al. 2017

Annals of Nuclear Energy

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Section: Hydrogen Transport and Hydride Formationsupporting

confidence: 89%

Mechanistic materials modeling for nuclear fuel performance

Tonks

Andersson

Phillpot

et al. 2017

Annals of Nuclear Energy

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“…It is possible that although the matrix H concentration was kept below the solubility limit, the local H concentration near surface could have exceeded that limit when surface segregation occurred. Consequently, precipitation of hydrides such as coherent hydride clusters37 may have occurred. As the hydrides habit on basal planes with small thickness along <c>, they block H diffusion along <c>, reducing D c without affecting D a much, similar to the situation for H in hcp Mg36.…”

Section: Resultsmentioning

confidence: 99%

Anisotropic hydrogen diffusion in α-Zr and Zircaloy predicted by accelerated kinetic Monte Carlo simulations

Zhang

Jiang

Bai

2017

Sci Rep

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This report presents an accelerated kinetic Monte Carlo (KMC) method to compute the diffusivity of hydrogen in hcp metals and alloys, considering both thermally activated hopping and quantum tunneling. The acceleration is achieved by replacing regular KMC jumps in trapping energy basins formed by neighboring tetrahedral interstitial sites, with analytical solutions for basin exiting time and probability. Parameterized by density functional theory (DFT) calculations, the accelerated KMC method is shown to be capable of efficiently calculating hydrogen diffusivity in α-Zr and Zircaloy, without altering the kinetics of long-range diffusion. Above room temperature, hydrogen diffusion in α-Zr and Zircaloy is dominated by thermal hopping, with negligible contribution from quantum tunneling. The diffusivity predicted by this DFT + KMC approach agrees well with that from previous independent experiments and theories, without using any data fitting. The diffusivity along is found to be slightly higher than that along , with the anisotropy saturated at about 1.20 at high temperatures, resolving contradictory results in previous experiments. Demonstrated using hydrogen diffusion in α-Zr, the same method can be extended for on-lattice diffusion in hcp metals, or systems with similar trapping basins.The diffusion of hydrogen in hcp metals has attracted extensive research interests for both its scientific merit and technological importance. It is the rate-limiting step for solid hydrogen storage in Mg based alloys 1 and for mechanical property degradation in Ti 2 and Zr alloys 3 . For instance, Zr-based alloys such as Zircaloy2 and Zircaloy4 (denoted as Zircaloy) are widely used as the cladding of fuel rods in nuclear reactors for their excellent corrosion resistance and very low absorption cross-section of thermal neutrons. During their service time, these alloys operate in extremely harsh environments, combining high temperatures and corrosive coolants such as water. In light water reactors, Zircaloy claddings directly contact coolant water, with a local temperature at the interface of about 400 °C. Consequently, Zircaloy react with water, producing an oxide layer at the interface and hydrogen (H), of which a substantial fraction infiltrates into the cladding interior 3 . Due to the low solubility of H in α -Zr, which is the main component of claddings, hydrides precipitate in the cladding matrix 4-6 . Hydride formation can detrimentally affect the cladding integrity primarily in two ways. First, the formation of hydrides reduces the fracture toughness of the cladding matrix 7,8 , increasing the probability of fuel failure when fuel-cladding mechanical interaction (PCMI) occurs during fuel operation. Second, upon cyclic thermal and mechanical loadings, hydrides can dissolve and reorient. During used-fuel storage, fracture may propagate via the so-called delayed-hydride-cracking (DHC) mechanism, which is induced by the dissolving of matrix hydrides and their subsequent re-formation at crack tips 3,9 . To mitigate thes...

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“…[1] since the hydride orientation is determined by the nucleation process and is maintained constant during growth. In presence of external stresses, elastic interaction between the precipitation-induced strain and the applied stress may have an influence on the orientation of the hydrides [26][27][28][29]. In addition, Zhu and coworkers completed a study on the ductility of each zirconium hydrides.…”

Section: Non-irradiated Materialsmentioning

confidence: 99%

Irradiation Hardening and Microstructure Characterization of Zr -1% Nb During Low Dose Neutron Irradiation

Vázquez¹,

Zelaya²,

Fortis³

et al. 2021

J. Mater. Appl.

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Due to low neutron absorption cross section, high mechanical strength, high thermal conductivity and good corrosion resistance in water and steam, Zirconium alloys are widely used as fuel cladding material in nuclear reactors. During life-time of a reactor the microstructure of this alloy is affected due to, among other factors, radiation damage and hydrogen damage. In this work mechanical properties changes on neutron irradiated Zr-1wt.% Nb at low temperatures (< 100 °C) and low dose (3.5 ´ 1023 n m-2 (E > 1 MeV)) were correlated with hydrides and crystal defects evolution during irradiation. To achieve this propose, tensile tests of: 1) Non-hydrided and non-irradiated material, 2) Hydrided and non-irradiated material and 3) Hydrided and irradiated material were performed at 25 ºC and 300 ºC. Different phases, hydrides and second phase precipitates were characterized by transmission electron microscopy (TEM) techniques. For the hydrided and irradiated material, the ductility decreased sharply with respect to the hydrided and non-irradiated material, among other factors, due to the change in the microstructure produced mainly by neutron irradiation. Even if the presence of the hydride ζ (zeta) was observed, both in the irradiated and non-irradiated material, tensile tests showed that ζ-hydrides did not affect ductility, since hydrided samples are more ductile than non-hydrided samples.

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Homogeneous hydride formation path in α-Zr: Molecular dynamics simulations with the charge-optimized many-body potential

Abstract: 2016-11-03T14:11:40

Cited by 49 publications

References 39 publications

Mechanistic materials modeling for nuclear fuel performance

Mechanistic materials modeling for nuclear fuel performance

Anisotropic hydrogen diffusion in α-Zr and Zircaloy predicted by accelerated kinetic Monte Carlo simulations

Irradiation Hardening and Microstructure Characterization of Zr -1% Nb During Low Dose Neutron Irradiation

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