Electronic Thesis and Dissertation Repository

Thesis Format

Integrated Article

Degree

Doctor of Philosophy

Program

Mechanical and Materials Engineering

Supervisor

Abdolvand, Hamidreza

Abstract

Due to their low neutron absorption cross-section and good corrosion properties, zirconium and its alloys have been widely used as the structural material in the core of nuclear reactors. These alloys are exposed to an intensive neutron flux which may lead to dimensional instabilities and the degradation of the mechanical properties of the alloy over the service time of the reactor. The changes in deformation behavior and mechanical properties can be traced back to the formation, evolution, and interaction of the irradiation-induced microstructural defects, e.g., point defect clusters, dislocation loops, and complex dislocation line networks. However, the materials constitutive models are rarely correlated to the irradiated-induced defects at the grain scale. Further, the available modeling approaches for simulating the deformation of irradiated materials are mostly empirical and generally do not incorporate the effects of microstructure or defect densities.

To simulate the mechanical behavior of zirconium alloys exposed to neutron radiation, the present research focuses on updating a crystal plasticity finite element model by firstly including the effects of dislocation densities. The results of the model are compared against previously published data for a series of in-situ neutron diffraction and high angular resolution electron backscatter diffraction (HR-EBSD) experiments conducted on un-irradiated α-zirconium specimens. The effects of implementing different formulations for determining dislocation densities are also investigated. It is shown that the calculated dislocation densities, stress, and rotation fields, as well as internal elastic strains, agree with the measured ones. The effects of irradiation growth are subsequently integrated into the model and the numerical results are compared to the previously published data. It is shown that the model is capable of determining the effects of material texture, grain size, and prior cold work on the evolution of average growth strain. It is shown, for the first time, that the growth strain is non-uniformly distributed among different grains and localized at the grain boundaries or slip bands.

Summary for Lay Audience

When any engineering device is designed, there is a fundamental question that needs to be answered: Will the product fail under the operating condition? The answer to this question is extremely important, to ensure the safe operation of the engineering component. To accurately assess the life cycle of engineering components, there has been an increasing demand in advanced numerical models. In this research, we are updating and providing a physical-based numerical tool which can be used to simulate the deformation of nuclear reactor core components made of zirconium. The presented numerical model can help nuclear engineers understand the performance of nuclear reactors core components. The model links the effects of zirconium microstructure to its macroscopic mechanical behaviour. It simulates the localization of stress and strain fields that might affect the failure of engineering components.

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