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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.

Irradiation induces a high concentration of defects in the structural materials of nuclear reactors, which are typically of body-centered cubic Iron (BCC Fe) and its alloys. The primary effect of irradiation is hardening which is caused by the blocking of dislocations with defects and defect clusters like point defects, self-interstitial loops, and voids. The dislocation-defect interactions are atomistic in nature due to the very small length and time scales involved, i.e., of the order of nanometers and picoseconds. To predict the effect of dislocation-defect interactions on the macroscopic mechanical and plastic behavior of the material, it is critically important to develop robust coupling schemes by which accurate atomic level physics of the rate-limiting kinetic processes can be informed into a coarse-grained model such as crystal plasticity. In this thesis we will develop an atomistically informed constitutive model. Relevant atomistic processes are identified from molecular dynamics simulations. The respective unit process studies are conducted using atomistic reaction pathway sampling methods like Nudged Elastic Band method. Stress-dependent activation energies and activation volumes are computed for various rate-liming unit processes like thermally activated dislocation motion via kinkpair nucleation, dislocation pinning due to self interstitial atom, etc. Constitutive laws are developed based on transition state theory, that informs the atomistically determined activation parameters into a coarse-grained crystal plasticity model. The macroscopic deformation behavior predicted by the crystal plasticity model is validated with experimental results and the characteristic features explained in the light of atomistic knowledge of the constituting kinetics. We also investigate on unique irradiation induced defects such as stacking fault tetrahedra, that are formed under non-irradiated condition. This thesis also includes our work on materials with internal interfaces that can resist irradiation induced damage. Overall, the research presented in this thesis involves the implementation and development of novel computational paradigm that encompasses computational approaches of various length and time scales towards robust predictions of the mechanical behavior of irradiated materials.

Silicon carbide and nanostructured ferritic alloy (SiC-NFA) composites have the potential to maintain the outstanding high temperature corrosion and irradiation resistance and enhance the mechanical integrity for nuclear cladding. However, the formation of detrimental silicide phases due to reaction between SiC and NFA remains a major challenge. By introducing a carbon interfacial barrier on NFA (C@NFA), SiC-C@NFA composites are investigated to reduce the reaction between SiC and NFA. In a similar way, the effect of chromium carbide (Cr3C2) interfacial barrier on SiC (Cr3C2@SiC) is also presented for Cr3C2@SiC-NFA composites. Both the coatings were successful in suppressing silicide formation. However, despite the presence of coatings, SiC was fully consumed during spark plasma sintering process. TEM and EBSD investigations revealed that spark plasma sintered SiC-C@NFA and Cr3C2@SiC-NFA formed varying amounts of different carbides such as (Fe,Cr)7C3, (Ti,W)C and graphite phases in their microstructure. Detailed microstructural examinations after long term thermal treatment at 1000oC on the microstructure of Cr3C2@SiC-NFA showed precipitation of new (Fe,Cr)7C3, (Ti,W)C carbides and also the growth of existing and new carbides. The results were successfully explained using ThermoCalc precipitation and coarsening simulations respectively. The oxidation resistance of 5, 15 and 25 vol% SiC@NFA and Cr3C2@SiC-NFA composites at 500-1000oC temperature under air+45%water vapor containing atmosphere is investigated. Oxidation temperature effects on surface morphologies, scale characteristics, and cross-sectional microstructures were investigated and analyzed using XRD and SEM. SiC-C@NFA showed reduced weight gain but also showed considerable internal oxidation. Cr3C2@SiC-NFA composites showed a reduction in weight gain with the increasing volume fraction of Cr3C2@SiC (5, 15 and 25) without any indication of internal oxidation in the microstructure. 25 vol% SiC-C@NFA and 25 vol% Cr3C2@SiC-NFA showed over 90% and 97% increase in oxidation resistance (in terms of weight gain) as compared to NFA. The results were explained using the fundamental understanding of the oxidation process and ThermoCalc/DICTRA simulations. Finally, the irradiation performance of SiC-C@NFA and Cr3C2@SiC-NFA composites was assessed in comparison with NFA using state-of-the-art TEM equipped with in-situ ion irradiation capability. Kr++ ions with 1 MeV energy was used for irradiation experiments. The effect of ion irradiation was recorded after particular dose levels (0-10 dpa) at 300oC and 450oC temperatures. NFA sample showed heavy dislocation damage at both 300oC and 450oC increasing gradually with dose levels (0-10 dpa). Cr3C2@SiC-NFA showed similar behavior as NFA at 300oC. However, at 450oC, Cr3C2@SiC-NFA showed remarkably low dislocation loop density and loop size as compared to NFA. At 300oC, microstructures of NFA and Cr3C2@SiC-NFA show predominantly 1/2<111> type dislocation loops. At 450oC, NFA showed predominantly <100> type loops, however, Cr3C2@SiC-NFA composite was still predominant in ½<111> loops. The possible reasons for this interesting behavior were discussed based on the large surface sink effects and enhanced interstitial-vacancy recombination at higher temperatures. The molecular dynamics simulations did not show considerable difference in formation energies of ½<111> and <100> loops for NFA and Cr3C2@SiC-NFA composites. The additional Si element in the SiC-NFA sample could have been an important factor in determining the dominant loop types. SiC-C@NFA composites showed heavy dislocation damage during irradiation at 300oC. At 450oC, SiC-C@NFA showed high dislocation damage in thicker regions. Thinner regions near the edge of TEM samples were largely free from dislocation loops. The precipitation and growth of new (Ti,W)C carbides were observed at 450oC with increasing irradiation dose. (Fe,Cr)7C3 precipitates were largely free from any dislocation damage. Some Kr bubbles were observed inside (Fe,Cr)7C3 precipitates and at the interface between α-ferrite matrix and carbides ((Fe,Cr)7C3, (Ti,W)C). The results were discussed using the fundamental understanding of irradiation and ThermoCalc simulations. / Doctor of Philosophy / With the United Nations describing climate change as 'the most systematic threat to humankind', there is a serious need to control the world's carbon emissions. The ever increasing global energy needs can be fulfilled by the development of clean energy technologies. Nuclear power is an attractive option as it can produce low cost electricity on a large scale with greenhouse gas emissions per kilowatt-hour equivalent to wind, hydropower and solar. The problem with nuclear power is its vulnerability to potentially disastrous accidents. Traditionally, fuel claddings, rods which encase nuclear fuel (e.g. UO2), are made using zirconium based alloys. Under 'loss of coolant accident (LOCA) scenarios' zirconium reacts with high temperature steam to produce large amounts of hydrogen which can explode. The risks associated with accidents can be greatly reduced by the development of new accident tolerant materials. Nanostructured ferritic alloys (NFA) and silicon carbide (SiC) are long considered are leading candidates for replacing zirconium alloys for fuel cladding applications. In this dissertation, a novel composite of SiC and NFA was fabricated using spark plasma sintering (SPS) technology. Chromium carbide (Cr3C2) and carbon (C) coatings were employed on SiC and NFA powder particles respectively to act as reaction barrier between SiC and NFA. Microstructural evolution after spark plasma sintering was studied using advanced characterization tools such as scanning electron microscopy (SEM), electron backscattered diffraction (EBSD), transmission electron microscopy (TEM) and energy dispersive spectroscopy (EDS) techniques. The results revealed that the Cr3C2 and C coatings successfully suppressed the formation of detrimental reaction products such as iron silicide. However, some reaction products such as (Fe,Cr)7C3 and (Ti,W)C carbides and graphite retained in the microstructure. This novel composite material was subjected to high temperature oxidation under a water vapor environment to study its performance under the simulated reactor environment. The degradation of the material due to high temperature irradiation was studied using state-of-the-art TEM equipped with in-situ ion irradiation capabilities. The results revealed excellent oxidation and irradiation resistance in SiC-NFA composites as compared to NFA. The results were discussed based on fundamental theories and thermodynamic simulations using ThermoCalc software. The findings of this dissertation imply a great potential for SiC-NFA based composites for future reactor material designs.

The purpose of this thesis is to improve predictions of irradiation hardening in metals with a focus on coarse-graining via meso-scale simulations. Increasing hardness and decreasing in ductility in nuclear reactor pressure vessel steel is the limiting factor of nuclear reactor life, and accurately predicting reactor life is of the utmost importance for the safe operation of nuclear facilities. This is an inherently multi-scale problem with primary damage occurring at the atomic scale and its effects propagating across ten orders of magnitude in length and time scale to changes in macroscopic material properties, which must be reflected in its methods of prediction. To achieve this goal, this thesis develops two novel approaches to simulate the motion of dislocations in irradiated alpha-iron. First, a dislocation dynamics simulation coarse-graining insight from atomistic dislocation-defect simulations is used to guide the selection of proposed constitutive models. Several studies investigating the effect of size distribution show that the mean defect size can be used with the selected models to predict material hardening without a complex treatment for the defect size distribution. The hardening effect of the commonly observed defect types are found independently and a superposition principle is proposed for materials with both defect types. Second, a link to transition state theory and thermally activated reactions is established using a new method augmenting a discrete dislocation dynamics simulations with the nudged elastic band method to characterise the minimum energy pathways of dislocation reactions. This development enables calculations of activation energy for dislocation events using a continuum method as well as the numerical calculations of dislocation attempt frequency. The thesis concludes with an extension to the analysis of coarse-graining unit events to large scale dislocation-obstacle bypass phenomena.

<p>An investigation of the influence of three alloying elements Chromium, Phosphorus, and Nitrogen with the solute types of oversized substitutional, undersized substitutional, and interstitial on the irradiation induced microstructural evolution and hardening</p>

The irradiation of crystalline materials in environments such as nuclear reactors leads to the accumulation of micro and nano-scale defects with a negative impact on material properties such as strength, corrosion resistance, and dimensional stability. Point defects in the crystal lattice, the vacancy and self-interstitial, form the basis of this damage and are capable of migrating through the lattice to become part of defect clusters and sinks, or to annihilate themselves. Recently, attention has been given to HEAs for fusion and fission components, as some materials of this class have shown resilience to irradiation-induced damage. The ability to predict defect diffusion and accelerate simulations of defect behaviour in HEAs using ML techniques is consequently a subject that has gathered significant interest. The goal of this work was to produce an unsupervised neural network capable of learning the interatomic dynamics within a specific HEA system from MD data in order to create a KMC type predictor of defect diffusion paths for common point defects in crystal systems such as the vacancy and self-interstitial. Self-interstitial defect states were identified and purified from MD datasets using graph-isomorphism, and a proof-of-concept model for the HEA environment was used with several interaction setups to demonstrate the feasibility of training a GCN to predict vacancy defect transition rates in the HEA crystalline environment.

The effect of proton (H+) irradiation on uranium mononitride (UN) and UN compositefuel with 10 at.% ZrN (UN10at%ZrN) was examined. Protons of 2 MeV with fluences of1E17, 1E18, 1E19 and 1E20 ions/cm2 were accelerated towards the fabricated samples in orderto investigate the evolution of the micro-structure. Stopping and Range of Ions in Matter(SRIM) calculations were performed to determine the displacements per atom associatedwith the depth of the highest damage, for each fluence.X-Ray diffraction (XRD) was used in both samples to identify the chemical composition ofeach pellet, which revealed the low presence of oxygen. Based on scanning electron microscopy(SEM), deterioration of the samples surface was observed, as the proton fluence increased.The applied stress due to the irradiation, led to the cracking of the pellets at the highestfluences. Blisters and craters appear to surround the cracked region, which might originatefrom the significant levels of hydrogen implantation within the samples.From Electron backscatter diffraction (EBSD) analysis, the grain size of the UN10at%ZrNcomposite was found to be smaller than in UN, due to the nano-particle nature of the ZrNpowder. The latter technique was also used to observe the elevated irradiated regions, whichwere further investigated by atomic force microscopy (AFM). Nano-indentation detectedirradiation hardening for both samples in the irradiated regions. Focused ion beam (FIB)milling was applied to remove lamellas from the cracked regions in both UN and compositesamples in order to be analyzed by transmission electron microscopy (TEM). The latter mightreveals the formation of dislocation loops in the irradiated areas. / Effekten av protonbestrålning på urannitrid (UN) och UN-kompositbränsle med 10 at.% ZrN (UN10at%ZrN) undersöktes. Protoner på 2 MeV med total dos på 1E17, 1E18, 1E19 och 1E20 joner/cm2 accelererades mot de tillverkade proverna för att undersöka utvecklingen av mikrostrukturen under bestrålning. SRIM-beräkningar (Stopping and Range of Ions in Matter) utfördes för att bestämma profilen på skadan och jonimplanteringen i förhållande till djupet, för varje dosnivå.  Röntgendiffraktion (XRD) användes på båda proverna för att identifiera den kemiska sammansättningen av varje kuts, vilket visade att syrehalten var låg. Med hjälp av svepelektronmikroskopi (SEM) observerades en försämring av provernas yta när protonflödet ökade. Den resulterande mekaniska spänningen överskred provets brottstyrka på djupet, eftersom nitriderna inte är så duktila, vilket ledde till sprickbildning i proverna som utsattes för de högsta doserna. Blåsor och kratrar omger det spruckna området, vilket beror på betydande väteimplantering i provet.  Genom electron backscatter diffraction analys (EBSD) konstaterades att kornstorleken hos UN10at%ZrN-kompositen var mindre än hos UN, på grund av ZrN-pulvrets nanopartikelnatur. Den sistnämnda tekniken användes för att observera de högt bestrålade områdena, som undersöktes ytterligare med Atomic force microscopy (AFM). Genom nano-indientation upptäcktes bestrålningshärdning för båda proverna i de bestrålade områdena. Fräsning med en fokuserad jonstråle (FIB) användes för att avlägsna lameller från de spruckna områdena i både UN- och kompositprovet för att kunna analyseras med transmission electron microscopy (TEM). Det senare visade att det bildades dislokationer i de bestrålade områdena.