Dislocation loops in irradiated zirconium

Gulden, T. D.; Bernstein, I. M.

doi:10.1080/14786436608244779

Cited by 47 publications

(14 citation statements)

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“…In fission reactors and nuclear waste disposal, zirconium and Zr-based alloys experience severe irradiation environments 3 . The exposure to extremely high heat 3 , high neutron and gamma doses 4 5 , and high energy product particles resulting from neutron-solid interactions 6 will induce microstructure and microchemistry changes, due to the atomic displacements and formation of point defects and defect clusters (e.g., the formation of < a > dislocation loops, < c > component dislocation loops and cavities (voids and bubbles) 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 ). These will lead to various structural failures including growth, hardening, creep, corrosion and amorphization 20 21 22 23 24 , which in turn, will impose severe limitations on possible operating parameters in terms of neutron and gamma doses and allowable stresses and temperatures.…”

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confidence: 99%

Irradiation deformation near different atomic grain boundaries in α-Zr: An investigation of thermodynamics and kinetics of point defects

Arjhangmehr

Feghhi

2016

Sci Rep

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Understanding radiation performance of nanocrystalline Zr-based alloys is essential to develop internal components and external cladding materials with self-healing capabilities for longer and safer life cycles in harsh reactor environments. However, the precise role of interfaces in modifying defect production and evolution in α-Zr is not yet determined. Using atomistic simulation methods, we investigate the influence of different atomic grain boundaries (GBs) in thermodynamic and kinetic properties of defects on short timescales. We observe that the sink efficiency and sink strength of interfaces vary significantly with the boundary structures, with a preference to absorb interstitials (vacancies) when the GBs are semi-parallel (semi-perpendicular) relative to the basal planes. Further, we identify three distinct primary cascade geometries, and find that the residual defect clustering in grain interiors depends on how the atomic GBs modify the spatial distribution of defects within the crystal structure. Finally, we explain and discuss the dynamic results in terms of energetic and kinetic behaviors of defects near the pristine and damaged boundaries. Eventually, these will provide a microscopic reference for further improving the radiation response of Zr by using fine grains or by introducing a high density of dispersoids in material metallurgy.

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confidence: 99%

Irradiation deformation near different atomic grain boundaries in α-Zr: An investigation of thermodynamics and kinetics of point defects

Arjhangmehr

Feghhi

2016

Sci Rep

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“…The degradation of mechanical properties and damage after irradiation have a strong relationship with the generation of irradiation defects. Dislocation loops of ⟨a⟩ and ⟨c⟩ types are well known as the main defect structures in Zr after high energy particle irradiation [1][2][3][4][5][6][7][8][9][10][11][12][13][14]. The ⟨a⟩ loops, no matter interstitial or vacancy type, were found forming on the primary and secondary prismatic plane with the same Burgers vector 1∕3 ⟨ 11 20 ⟩ [1,4].…”

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confidence: 99%

Transmission electron microscopy characterization of dislocation loops in irradiated zirconium

Liu

Han

2021

Tungsten

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Characterization of irradiation defects is of great importanceto mitigate irradiation damage, reduce irradiation growth and tune mechanical properties in Zr alloys. Here, we describe a practical method to characterize the dislocation loops in irradiated Zr using conventional transmission electron microscopy (TEM). Vacancy or interstitial nature of dislocation loops is determined using the inside and outside contrast method. The habit plane of dislocation loops is determined by tilting the sample to multiple zone axes and judged based on the projected loop shape. The size of ⟨a⟩ loops is measured by tilting the sample to an edge-on position and the loop number is counted under a weak-beam dark-field TEM condition. ⟨c⟩ loops have a line contrast under viewing direction of a-axis and a circular shape under viewing direction of c-axis. In addition, a large number of triangle-shaped vacancy platelets (TVPs) were formed on the basal plane. With increasing the irradiation damage from 0.5 to 1.5 dpa, the number density of ⟨a⟩ loops keeps constant, while the number density of TVPs increased significantly, owing to the anisotropic diffusion and accumulation of point defects within basal plane. The methods introduced here are easy to follow and extend into other related investigations.

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“…Irradiation hardening is primarily the result of the high number density of small hai loops (4-30 nm diameter) that are formed on the prism planes (f1 0 1 0g) with a Burgers vector b hai ¼ 1 3 h1 1 2 0i [1-3,6-10, [13][14][15][16][24][25][26][27][28][29][30][31][32][33]. These hai loops are nucleated after a neutron fluence of 0.3-1.1 Â 10 24 n/m 2 (E > 1 MeV) and then rapidly saturate to a relatively constant size/number density at a fluence of about 1-5 Â 10 25 n/m 2 , depending on the irradiation temperature [1][2][3][4][14][15][16].…”

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confidence: 99%

The recovery of irradiation damage for Zircaloy-2 and Zircaloy-4 following low dose neutron irradiation at nominally 358 °C

Cockeram

Leonard

Byun

et al. 2015

Journal of Nuclear Materials

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a b s t r a c tThe recovery of irradiation damage in wrought Zircaloy-2 and Zircaloy-4 was determined following a series of post-irradiation anneals at temperatures ranging from 343°C to 510°C and for time periods ranging from 1-h to 500 h. The materials had been irradiated at nominally 358°C in the High Flux Isotope Reactor (HFIR) at neutron fluences of nominally 3 Â 10 25 n/m 2 (E > 1 MeV). Irradiation at nominally 358°C resulted in a coarser distribution of hai loops that result in a 25-45% lower irradiation hardening than reported in the literature for irradiations at 260-326°C. The irradiation hardening and recovery were determined using tensile testing at room-temperature. Post-irradiation annealing at 343-427°C was shown to result in an increase in irradiation hardening to values even higher than for the as-irradiated material in the first 1-10 h of annealing. This Radiation Anneal Hardening (RAH) was followed by a relatively slow recovery of the irradiation damage. Much faster recovery with no RAH was observed for post-irradiation annealing at temperatures of 454-510°C. Irradiation at 358°C was shown to result in different recovery kinetics than observed in the literature for irradiation at 260-326°C. While the general trend described above is true for the four materials tested (alpha-annealed and beta-treated Zircaloy-2 and Zircaloy-4), notable and yet unexplained differences in RAH and in recovery are observed between the materials that might be a result of differing solute effects. Examinations of microstructure using Transmission Electron Microscopy were used to investigate the RAH and recovery mechanisms. Agreement between the measured and calculated irradiation hardening using a generalized Orowan hardening model to account for the observed loop structure was not as close for the post irradiation annealed condition as for the as-irradiated condition, which can likely be attributed to unaccounted for changes in the configuration of the hai loops to dislocation lines, segregation of solutes to dislocation loops, and the potential for the formation of fine clusters of point defects or solutes during annealing.

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Dislocation loops in irradiated zirconium

Cited by 47 publications

References 5 publications

Irradiation deformation near different atomic grain boundaries in α-Zr: An investigation of thermodynamics and kinetics of point defects

Irradiation deformation near different atomic grain boundaries in α-Zr: An investigation of thermodynamics and kinetics of point defects

Transmission electron microscopy characterization of dislocation loops in irradiated zirconium

The recovery of irradiation damage for Zircaloy-2 and Zircaloy-4 following low dose neutron irradiation at nominally 358 °C

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