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There has been a significant push in robotics research toward robot autonomy. However, full autonomy is currently impractical for all but the most clearly defined tasks in the most structured environments. However, as tasks become less defined and environments become cluttered and less controlled, there is still a benefit to implementing semi-autonomous behaviors where aspects of the tasks are completed autonomously thus reducing the burden on the human operator. A key component of a robot control system that supports this functionality is a robust world model to act as a repository of environmental information. The research community has provided many world-modeling solutions to support autonomous vehicle navigation. As such, they focus primarily on preventing collisions with the environment. Modeling schemes designed for collision prevention are of limited use to robotic manipulators that must have contact interaction with the environment as a matter of course. This thesis presents a world-modeling scheme that abstracts the model of the environment into a graph structure. This abstraction separates the concepts of entities in the environment from their relationships to the environment. The result is an intuitive world model that supports not only collision detection, but also motion planning and grasping. The graph-based world model presented can be searched by semantic type and tag values, allowing any number of agents to simultaneously use and update the model without causing failures elsewhere in the system. These capabilities are demonstrated on two different automated hot-cell glovebox systems, and one mobile manipulation system for use in remote contamination testing. / text

Zirconium and its alloys are of technical importance, finding application as a structural material in the nuclear industry. Engineering components fabricated from zirconium slowly pick-up hydrogen as a result of in-reactor corrosion, degrading the components mechanical properties as a brittle hydride phase forms. This dissertation applies high energy X-ray diffraction to directly measure the mechanical properties of zirconium hydrides in the bulk and at stress concentrations in zirconium alloys. The current study is presented as a manuscript format dissertation comprised of three manuscript chapters. Chapter 3 reports the in-situ loading of hydrided Zircaloy-2 and discusses hydride/Zircaloy-2 matrix interactions as a function of applied load. Chapter 4 reports the mechanical behavior at a fatigue pre-crack in un-hydrided Zircaloy-2, comparing the results to finite element and polycrystalline plasticity models of the crack tip. Chapter 5 reports the effect of hydrides on the notch tip strain field. The three manuscript chapters are followed by a general discussion in Chapter 6 and conclusions in Chapter 7. / Thesis (Ph.D, Mechanical and Materials Engineering) -- Queen's University, 2009-09-27 20:32:01.455

Zirconium alloy pressure tubes are an important component in CANDU nuclear reactors. During operation these tubes can pick up hydrogen as a result of a corrosion reaction, which can eventually lead to the precipitation of a secondary, brittle zirconium hydride phase. Hydrides tend to first form at flaws (stress concentrations), and when they fracture can initiate a time-controlled crack growth mechanism known as delayed hydride cracking (DHC). Since DHC is a known failure mechanism for pressure tubes, and an ongoing concern in the nuclear industry, more fundamental knowledge is required about the behaviour of hydrides precipitated at flaws. Several approaches were employed in this thesis to better characterize the effects and behaviour of hydrides at such stress concentrations. High energy X-ray diffraction, as well as in-situ SEM testing coupled with digital image correlation, were used to map the strains around stress concentrations where hydrides were present. These studies highlighted important differences in the behaviour of the hydride phase and the surrounding zirconium. To gain greater insight into hydride morphology, neutron tomography was used in an attempt to measure the through-thickness hydride distribution at flaws. A finite element model was also developed and verified against the X-ray strain mapping results. This model provided greater insight into details that could not be obtained directly from the experimental approaches, as well as providing a framework for future modeling to predict the effects of hydride precipitation under different conditions. Taken as a whole, these studies provide important information for improving service guidelines and avoiding conditions that favour embrittlement due to hydride precipitation. / Thesis (Master, Mechanical and Materials Engineering) -- Queen's University, 2011-01-17 20:39:25.621

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The requirements for the next generation of nuclear reactors call for more radiation tolerant materials. One such material, nanostructured ferritic alloys (NFA) are a candidate material for use in cladding. The radiation tolerance of NFAs comes from the high number density of small oxide nanofeatures composed of Y, Ti, and O. These oxide nanofeatures or nano-oxides act as alternative nucleation sites for bubbles of transmutation He, thus preventing the accumulation of He atoms at the grain boundaries which would embrittle the metal. To further study the material, a mean-field rate theory model (MF-RTM) was created to simulate the radiation-induced segregation (RIS) of the alloy components Y, Ti, and O to the grain boundaries. Later, a kinetic Monte Carlo model (KMC) was made that replicated the results from the rate theory for the radiation induced segregation. Then the KMC model was modified to study the nano-oxide behavior in a range of different behaviors; the nano-oxide precipitation kinetics during heat treatment, resistance to dissolution under irradiation regimes similar to reactor conditions, and ability to trap He bubbles on the nano-oxide surfaces rather than the grain boundary. This KMC model is more complex than others as it includes 5 different atomic species (Fe, Y, Ti, O, and He) which migrate through three different mechanisms. Findings from the precipitation heat treatments were able to replicate the size, number density, and composition of nano-oxides from experiments and determined vacancy trapping at oxide interfaces was a significant for the NFA's coarsening resistance as opposed to interference from dislocations. In the irradiation simulations, the resistance of the nano-oxides to dissolution was confirmed and found the excess vacancy population plays an important role in healing the nano-oxides. He bubbles formed in the KMC simulations were found to preferentially form at the oxide interfaces, particularly the <111> interface, rather than the grain boundary and the characteristics of the He bubbles match expectations from literature. In the development of the KMC model, new insights into steady-state detection concepts were also found. A new type of steady-state detection (SSD) algorithm is described. Additionally, a method of forecasting the number of data points needed to make an accurate determination of steady-state, a 'predicting the pre-requisite to steady state detection' (ppSSD), is explored. / Doctor of Philosophy / Nuclear reactors need more radiation tolerant materials in the future, such as nanostructured ferritic alloys (NFA), used for nuclear fuel rod cladding, whose large amount of nanometer sized oxide particles contribute substantially to the radiation resistance of the metal overall. A mean-field rate theory method(MF-RTM) and a Kinetic Monte Carlo (KMC) computer model were made to study radiation induced segregation in the material. A more complex 5 element (Fe, Y, Ti, O, and He) KMC code was later developed to study the influence of the oxides at high temperatures and dose rates to gain insight into the causes the oxides remarkable thermal stability and resistance to irradiation. At all stages, the KMC model was able to replicate material behavior under high temperature heat treatment and irradiation. The model was used to simulate the formation of these oxides under different temperatures during their initial processing to gain more knowledge on how the oxide characteristics (size and number density) are influenced by temperature so we can tailor the processing method to achieve an ideal distribution of oxides in the material. Additionally, a mechanism for the oxides resistance to high temperature coarsening unrelated to the expected one caused by dislocations. The irradiation resistance of oxides to dissolution from irradiation was also investigated. While experimental measurements give a before and after picture of a material that underwent irradiation, the KMC can show the time evolution of the oxide size with increasing irradiation damage so the mechanisms behind the radiation resistance can be understood. The oxides remained stable at all temperatures and dose rates. Excess vacancies were found to play an important role in stabilizing the oxides against radiation damage. The KMC model also confirmed the ability of the oxides to trap transmutation He at the interfaces rather than the grain boundary and observed the process of He bubble nucleation. The He bubble form at the <111> oxide interface and they possess similar characteristics of He bubbles expected from literature. Additionally, a novel steady-state detection (SSD) algorithm was developed that can be used for long-term simulations and a method to determine how many data points the algorithm needs to accurately detect steady state is described here.

Made available in DSpace on 2014-10-09T12:34:36Z (GMT). No. of bitstreams: 0 / Made available in DSpace on 2014-10-09T14:01:01Z (GMT). No. of bitstreams: 0 / Dissertação (Mestrado) / IPEN/D / Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP