Mechanical Property Testing of Hydrogenated Zirconium Irradiated with Electrons

Homogeneous distribution of fine second-phase particles (SPPs) fabricated by cycles of deformation and annealing in zirconium alloys is a critical consideration for the corrosion resistance of fuel claddings. Different deformation degrees of zirconium alloys would result in distinctive microstructures, leading to a distinct growth of SPPs during subsequent annealing. Unfortunately, the homogenization and growth behavior of SPPs in deformed zirconium alloys have not been well studied. In this work, a β-quenched Zr–Sn–Nb–Fe–Cu–Si–O alloy was rolled and annealed at 580 °C or 680 °C. The morphologies, distributions, and sizes of SPPs resulting from the different processing procedures were investigated. A linear distribution of SPPs is found in the β-quenched sample. Afterward, SPPs grow and are randomly distributed during heat treatment as the deformation degree or annealing time (or temperature) increases. The homogenization and growth of SPPs are attributed to the Ostwald ripening mechanism that is governed by lattice diffusion and short-circuit diffusion. The sample with a higher deformation degree is speculated to have a larger number of defects that provide more shortcuts for the mass transfer of SPPs, thereby facilitating a homogeneous distribution of fine SPPs during annealing.

Section: Introductionmentioning

confidence: 99%

Homogenization and Growth Behavior of Second-Phase Particles in a Deformed Zr–Sn–Nb–Fe–Cu–Si–O Alloy

Chen

Sang

Zhang

et al. 2018

“…On the other hand, experimental observation and test offer the most reliable data for the analysis of irradiation-hardening and embrittlement. The informed knowledge ranges from the formation of irradiation-induced defects at nano scale [50][51][52], to the generation of defect-free channels within the grain interiors and across adjacent grains at micro scale [53,54], to the stress-strain relationship, hardness-depth curve and fracture toughness-temperature relationship at macro scale [55][56][57][58][59]. It can also provide the information for the type, distribution and density of irradiation defects at the static state [60,61], and the dynamic interaction between dislocations and defects during the irradiation and loading process [62,63].…”

Section: Introductionmentioning

confidence: 99%

Fundamental Mechanisms for Irradiation-Hardening and Embrittlement: A Review

Xiao

2019

It has long been recognized that exposure to irradiation environments could dramatically degrade the mechanical properties of nuclear structural materials, i.e., irradiation-hardening and embrittlement. With the development of numerical simulation capability and advanced experimental equipment, the mysterious veil covering the fundamental mechanisms of irradiation-hardening and embrittlement has been gradually unveiled in recent years. This review intends to offer an overview of the fundamental mechanisms in this field at moderate irradiation conditions. After a general introduction of the phenomena of irradiation-hardening and embrittlement, the formation of irradiation-induced defects is discussed, covering the influence of both irradiation conditions and material properties. Then, the dislocation-defect interaction is addressed, which summarizes the interaction process and strength for various defect types and testing conditions. Moreover, the evolution mechanisms of defects and dislocations are focused on, involving the annihilation of irradiation defects, formation of defect-free channels, and generation of microvoids and cracks. Finally, this review closes with the current comprehension of irradiation-hardening and embrittlement, and aims to help design next-generation irradiation-resistant materials.

“…Zirconium (Zr) alloys are widely used as fuel claddings in nuclear reactors due to their adequate mechanical properties, low neutron absorption rate, and excellent corrosion resistance in service environments [1][2][3][4]. Among these properties, corrosion resistance is one of the most important considerations, as corrosion limits the service life of Zr alloys and is related to the safety of nuclear reactors [5][6][7][8].…”

Section: Introductionmentioning

confidence: 99%

Calculation of Oxygen Diffusion Coefficients in Oxide Films Formed on Low-Temperature Annealed Zr Alloys and Their Related Corrosion Behavior

Zhang

Chen

Zhao³

et al. 2019

The growth of oxide film, which results from the inward oxygen diffusion from a corrosive environment, is a critical consideration for the corrosion resistance of zirconium alloys. This work calculates the oxygen diffusion coefficients in the oxide films formed on zirconium alloys annealed at 400~500 °C and investigates the related corrosion behavior. The annealed samples have a close size for the second-phase particles but a distinctive hardness, indicating the difference in substrate conditions. The weight gain of all samples highly follows parabolic laws. The weight gain of the sample annealed at 400 °C has the fastest increase rate at the very beginning of the corrosion test, but its oxide film has the slowest growth rate as the corrosion proceeds. By contrast, the sample annealed at 500 °C shows the lowest weight gain but the highest corrosion rate constant. Such a corrosion behavior is attributed to the amount of defects existing in the oxide film formed on the annealed samples; fewer defects would provide a lower fraction of short-circuit diffusion in total diffusion, resulting in a lower diffusion coefficient of oxygen in the oxide film, thereby producing better corrosion resistance. This is consistent with the calculated diffusion coefficients of oxygen in the oxide films: 3.252 × 10−11 cm2/s, 3.464 × 10−11 cm2/s and 3.740 × 10−11 cm2/s for the samples annealed at 400 °C, 450 °C, and 500 °C, respectively.