Publications

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Irradiation resistance of metallic nanostructured multilayers is determined by the interactions between defects and phase boundaries. However, the dose-dependent interfacial morphology evolution can greatly change the nature of the defect-boundary interaction mechanisms over time. In the present study, we used atomistic models combined with a novel technique based on the accumulation of Frenkel pairs to simulate irradiation processes. We examined dose effects on defect evolutions near zirconium-niobium multilayer phase boundaries. Our simulations enabled us to categorize defect evolution mechanisms in bulk phases into progressing stages of dislocation accumulation, saturation, and coalescence. In the metallic multilayers, we observed a phase boundary absorption mechanism early on during irradiation, while at higher damage levels, the increased irradiation intermixing triggered a phase transformation in the Zr-Nb mixture. This physical phenomenon resulted in the emission of a large quantity of small immobile dislocation loops from the phase boundaries.

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The ability of nanoporous metals to avoid accumulation of damage under ion beam irradiation has been the focus of several studies in recent years. The width of the interconnected ligaments forming the network structure typically is on the order of tens of nanometers. In such confined volumes with high amounts of surface area, the accumulation of damage (defects such as stacking-fault tetrahedra and dislocation loops) can be mitigated via migration and annihilation of these defects at the free surfaces. In this work, in situ characterization of radiation damage in nanoporous gold (np-Au) was performed in the transmission electron microscope. Several samples with varying average ligament size were subjected to gold ion beams having three different energies (10 MeV, 1.7 MeV and 46 keV). The inherent radiation tolerance of np-Au was directly observed in real time, for all ion beam conditions, and the degree of ion-induced damage accumulation in np-Au ligaments is discussed here.

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This study was initiated to quantify and characterize the uncertainty associated with the degradation mechanisms impacting normal dry storage operations for used nuclear fuel (UNF) and normal conditions of transport in support of the Spent Fuel and Waste Science & Technology Campaign (SFWST) and its effectiveness to rank the data needs and parameters of interest. This report describes the technical basis and guidance resulting from the development of software to perform uncertainty quantification (UQ) by developing and describing a holistic model that integrates the various processes controlling Atmospheric Stress Corrosion Cracking (ASCC) in the specific context of Interim Spent Fuel Storage Installations (ISFSIs). These processes include the daily and annual cycles of temperature and humidity associated with the environment, the deposition of chloride-containing aerosol particles, pit formation, pit-to-crack transition, and crack propagation.

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For long-term storage, spent nuclear fuel (SNF) is placed in dry storage systems, commonly consisting of welded stainless steel canisters enclosed in ventilated overpacks. Choride-induced stress corrosion cracking (CISCC) of these canisters may occur due to the deliquescence of sea-salt aerosols as the canisters cool. Current experimental and modeling efforts to evaluate canister CISCC assume that the deliquescent brines, once formed, persist on the metal surface, without changing chemical or physical properties. Here we present data that show that magnesium chloride rich-brines, which form first as the canisters cool and sea-salts deliquesce, are not stable at elevated temperatures, degassing HCl and converting to solid carbonates and hydroxychloride phases, thus limiting conditions for corrosion. Moreover, once pitting corrosion begins on the metal surface, oxygen reduction in the cathode region surrounding the pits produces hydroxide ions, increasing the pH under some experimental conditions, leads to precipitation of magnesium hydroxychloride hydrates. Because magnesium carbonates and hydroxychloride hydrates are less deliquescent than magnesium chloride, precipitation of these compounds causes a reduction in the brine volume on the metal surface, potentially limiting the extent of corrosion. If taken to completion, such reactions may lead to brine dry-out, and cessation of corrosion.

Safety basis analysts throughout the U.S. Department of Energy (DOE) complex rely heavily on the information provided in the DOE Handbook, DOE-HDBK-3010, Airborne Release Fractions/Rates and Respirable Fractions for Nonreactor Nuclear Facilities, to determine radionuclide source terms from postulated accident scenarios. In calculating source terms, analysts tend to use the DOE Handbook's bounding values on airborne release fractions (ARFs) and respirable fractions (RFs) for various categories of insults (representing potential accident release categories). This is typically due to both time constraints and the avoidance of regulatory critique. Unfortunately, these bounding ARFs/RFs represent extremely conservative values. Moreover, they were derived from very limited small-scale bench/laboratory experiments and/or from engineered judgment. Thus, the basis for the data may not be representative of the actual unique accident conditions and configurations being evaluated. The goal of this research is to develop a more accurate and defensible method to determine bounding values for the DOE Handbook using state-of-art multi-physics-based computer codes. This enables us to better understand the fundamental physics and phenomena associated with the types of accidents in the handbook. In this fourth year, we improved existing computational capabilities to better model fragmentation situations to capture small fragments during an impact accident. In addition, we have provided additional new information for various sections of Chapters 4 and 5 of the Handbook on free fall powders and impacts of solids, and have provided the damage ratio simulations for containers (7A drum and standard waste box) for various drops and impact scenarios. Thus, this work provides a low-cost method to establish physics-justified safety bounds by considering specific geometries and conditions that may not have been previously measured and/or are too costly to perform during an experiment.

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We performed a systematic study of the threshold displacement energy (E_d) in metallic uranium as a function of both the recoil direction and temperature using Molecular Dynamics simulations. We developed a novel orientation sampling scheme that utilizes crystallographic symmetrical geodesic grids to select directions from the orientation fundamental zone to study the directional dependency. Additionally, we studied the temperature dependency by considering both the α-uranium phase, corresponding to the ground state for temperatures ranging from 0 K to 600 K, and the γ-uranium phase, corresponding to high-temperature state for temperatures above 900 K. In this study, we compared several definitions of the threshold energy: a direction-specific threshold displacement energy (E_d (θ,Φ)), an angle-averaged threshold energy ($E_d^{ave}$), a production probability threshold displacement energy ($E_d^{pp}$), and a defect count threshold displacement energy ($E_d^{dc}$). The direction-specific threshold displacement energies showed large angular anisotropy and variations in E_d results in accordance with crystallographic considerations. Specifically, preferred defect channeling directions were observed in the [120], [1$\bar{2}$0], [1$\bar{1}$1] directions for the α-uranium, and [001], [111] directions for the γ-uranium. The production probability threshold displacement energy ($E_d^{pp}$) is calculated as approximately 99.2659 eV at 10 K (α-U), 103.4980 eV at 300 K (α-U), 76.0915 eV at 600 K (α-U), and 42.9929 eV at 900 K (γ-U). With exception of those calculated at 10 K, threshold displacement energies decrease with increasing temperature. Analyses of the stable defect structures showed that the most commonly observed interstitial configuration in α-uranium consists of a ( 0 1 0 ) dumbbell-like interstitial; while in γ-uranium no preferential defect configuration could be identified due to thermally-induced lattice instabilities at the elevated temperatures.

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Semi-coherent cube-on-cube miscible U–Zr interfaces were studied using molecular dynamics simulations. The misfit accommodation of such semi-coherent phase boundaries was characterized by a two-dimensional dislocation network model utilizing a combination of theoretical predictions and analysis of the atomic system. The dislocation networks were discussed for various stacking orientation of the adjoining phases in terms of the composition of the dislocation sets, the partitioning between edge and screw components and the associated residual elastic fields. These analyses showed that the patterning of the network of dislocations forming these phase boundaries results from the competition between a structurally-driven process (i.e., function of the lattice misfit) and a chemically-driven process (i.e., due to the miscibility between U and Zr).

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This project focused on providing a fundamental mechanistic understanding of the complex degradation mechanisms associated with Pellet/Clad Debonding (PCD) through the use of a unique suite of novel synthesis of surrogate spent nuclear fuel, in-situ nanoscale experiments on surrogate interfaces, multi-modeling, and characterization of decommissioned commercial spent fuel. The understanding of a broad class of metal/ceramic interfaces degradation studied within this project provided the technical basis related to the safety of high burn-up fuel, a problem of interest to the DOE.

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The evolution and characterization of single-isolated-ion-strikes are investigated by combining atomistic simulations with selected-area electron diffraction (SAED) patterns generated from these simulations. Five molecular dynamics simulations are performed for a single 20 keV primary knock-on atom in bulk crystalline Si. The resulting cascade damage is characterized in two complementary ways. First, the individual cascade events are conventionally quantified through the evolution of the number of defects and the atomic (volumetric) strain associated with these defect structures. These results show that (i) the radiation damage produced is consistent with the Norgett, Robinson, and Torrens model of damage production and (ii) there is a net positive volumetric strain associated with the cascade structures. Second, virtual SAED patterns are generated for the resulting cascade-damaged structures along several zone axes. The analysis of the corresponding diffraction patterns shows the SAED spots approximately doubling in size, on average, due to broadening induced by the defect structures. Furthermore, the SAED spots are observed to exhibit an average radial outward shift between 0.33% and 0.87% depending on the zone axis. This characterization approach, as utilized here, is a preliminary investigation in developing methodologies and opportunities to link experimental observations with atomistic simulations to elucidate microstructural damage states.

The two-dimensional elastic Green’s function is calculated for a general anisotropic elastic bimaterial containing a line dislocation and a concentrated force while accounting for the interfacial structure by means of a generalized interfacial elasticity paradigm. The introduction of the interface elasticity model gives rise to boundary conditions that are effectively equivalent to those of a weakly bounded interface. The equations of elastic equilibrium are solved by complex variable techniques and the method of analytical continuation. The solution is decomposed into the sum of the Green’s function corresponding to the perfectly bonded interface and a perturbation term corresponding to the complex coupling nature between the interface structure and a line dislocation/concentrated force. Such construct can be implemented into the boundary integral equations and the boundary element method for analysis of nano-layered structures and epitaxial systems where the interface structure plays an important role.

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The complexity of radiation effects in a material’s microstructure makes developing predictive models a difficult task. In principle, a complete list of all possible reactions between defect species being considered can be used to elucidate damage evolution mechanisms and its associated impact on microstructure evolution. However, a central limitation is that many models use a limited and incomplete catalog of defect energetics and associated reactions. Even for a given model, estimating its input parameters remains a challenge, especially for complex material systems. Here, we present a computational analysis to identify the extent to which defect accumulation, energetics, and irradiation conditions can be determined via forward and reverse regression models constructed and trained from large data sets produced by cluster dynamics simulations. A global sensitivity analysis, via Sobol’ indices, concisely characterizes parameter sensitivity and demonstrates how this can be connected to variability in defect evolution. Based on this analysis and depending on the definition of what constitutes the input and output spaces, forward and reverse regression models are constructed and allow for the direct calculation of defect accumulation, defect energetics, and irradiation conditions. Here, this computational analysis, exercised on a simplified cluster dynamics model, demonstrates the ability to design predictive surrogate and reduced-order models, and provides guidelines for improving model predictions within the context of forward and reverse engineering of mathematical models for radiation effects in a materials’ microstructure.

Diffusion of point defects during irradiation is simulated via a two-way coupling between mechanical stress and defect diffusion in iron. This diffusion is based on a modified chemical potential that includes not only the local concentration of radiation-induced defects, but also the influence of the residual stress field from both the microstructure (i.e. dislocations or grain boundaries) and the eigenstrain caused by the defects themselves. Defect flux and concentration rates are derived from this chemical potential using Fick's first and second laws. Mean field rate theory is incorporated to model the annihilation of Frenkel pairs, and increased annihilation near grain boundaries is included based on the elastic energy of each grain boundary. Mechanical equilibrium is coupled with diffusion by computing eigenstrain from point defects and adding this to the total strain. Intrinsic stresses associated with the dislocations and grain boundaries are calculated using dislocation and disclination mechanics. Through this two-way-coupled model, regions of low concentration are seen near grain boundaries, and sink efficiency is calculated for different types of microstructure. The results show that the two-way mechanical coupling has a strong influence on sink efficiency for dislocation loops. The results also suggest that misorientation is a poor metric for determining sink efficiency, with sink density and elastic energy being much more informative.

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All grain boundaries are not equal in their predisposition for fracture due to the complex coupling between lattice geometry, interfacial structure, and mechanical properties. The ability to understand these relationships is crucial to engineer materials resilient to grain boundary fracture. Here, a methodology is presented to isolate the role of grain boundary structure on interfacial fracture properties, such as the tensile strength and work of separation, using atomistic simulations. Instead of constructing sets of grain boundary models within the misorientation/structure space by simply varying the misorientation angle around a fixed misorientation axis, the proposed method creates sets of grain boundary models by means of isocurves associated with important fracture-related properties of the adjoining lattices. Such properties may include anisotropic elastic moduli, the Schmid factor for primary slip, and the propensity for simultaneous slip on multiple slip systems. This approach eliminates the effect of lattice properties from the comparative analysis of interfacial fracture properties and thus enables the identification of structure-property relationships for grain boundaries. As an example, this methodology is implemented to study crack propagation along Ni grain boundaries. Segregated H is used as a means to emphasize differences in the selected grain boundary structures while keeping lattice properties fixed.

Molecular dynamics construction of the Arrhenius plot accounts for all possible diffusion paths in defective materials.

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