Computationally-efficient stochastic cluster dynamics method for modeling damage accumulation in irradiated materials

An improved version of a recently developed stochastic cluster dynamics (SCD) method {[}Marian, J. and Bulatov, V. V., {\it J. Nucl. Mater.} \textbf{415} (2014) 84-95{]} is introduced as an alternative to rate theory (RT) methods for solving coupled ordinary differential equation (ODE) systems for irradiation damage simulations. SCD circumvents by design the curse of dimensionality of the variable space that renders traditional ODE-based RT approaches inefficient when handling complex defect population comprised of multiple (more than two) defect species. Several improvements introduced here enable efficient and accurate simulations of irradiated materials up to realistic (high) damage doses characteristic of next-generation nuclear systems. The first improvement is a procedure for efficiently updating the defect reaction-network and event selection in the context of a dynamically expanding reaction-network. Next is a novel implementation of the $\tau$-leaping method that speeds up SCD simulations by advancing the state of the reaction network in large time increments when appropriate. Lastly, a volume rescaling procedure is introduced to control the computational complexity of the expanding reaction-network through occasional reductions of the defect population while maintaining accurate statistics. The enhanced SCD method is then applied to model defect cluster accumulation in iron thin films subjected to triple ion-beam ($\text{Fe}^{3+}$, $\text{He}^{+}$ and $$\text{H\ensuremath{{}^{+}}}$$) irradiations, for which standard RT or spatially-resolved kinetic Monte Carlo simulations are prohibitively expensive

Micro-mechanical testing of ceramic matrix composites; Extraction of critical interface properties and impact on composite optimization

Ceramic matrix composites have gained significant attention for improving strength and efficiency of high temperature components in the aerospace and nuclear industries. A critical step for this technology is the development of models that capture macroscopic failure behavior and probability based on constituent-level characteristics. This research presents the applicability of small-scale mechanical testing to probe relevant interface properties in the context of composite toughening mechanisms. An experimental case study applying micro-pillar compression for interfacial shear strength is presented. The test scope compares property data for two fiber types, HNLS and SA3, across varying pyrolytic carbon interface thickness (50-1000nm). The Mohr-Coulomb friction criterion was applied to extract the cohesive debond shear strength and internal friction coefficient, shown in figure 1(left). Experimental results showed a non-linear trend as a function of increasing normal load. These results were further characterized with FEM to develop a modified criterion that corrects for boundary conditions and helps separate property dependencies on roughness, thickness, and normal stress. These properties are discussed and used to inform analytical models that traditionally back-calculate a blanket parameter, τ, based on uniform crack spacing shown in figure 1 (right). Finally, novel techniques and preliminary results to characterize dynamic friction and wear in operando are discussed. Please click Additional Files below to see the full abstract

Localized mechanical properties of SiC-SiC fiber composites in extreme environments – a micromechanical study

Silicon carbide ceramics is a promising candidate material for the use in applications where a structural material able to withstand extreme environments, in particular high temperature, is required. These include applications in aero-engines (high-pressure turbine shrouds, combustor liners and turbine nozzles, among others); in recent years there is also a considerable interest in its use in nuclear applications, in particular as an accident-tolerant fuel (ATF) cladding material. The practical use of SiC is hindered by its inherent brittleness, and therefore it is usually suggested to be used in the form of a fiber composite. Mechanical properties of the composites are largely determined by the properties of their constituents, in particular interphases but also matrix and fibers; of particular practical interest is the evolution of such properties as a function of temperature and/or radiation damage. These can be measured using micromechanical testing tools, such as nanoindentation (measuring hardness and elastic modulus) and microcantilever fracture (measuring fracture strength and toughness). In this contribution we present the results of such measurements performed in the range of temperatures, including samples that were exposed to ion irradiation. Thus, the effects of temperature and radiation are investigated and rationalized. Material used in this study was manufactured using Tyranno-SA3 fibers in plain weave geometry, coated with pyrolytic carbon, and matrix grown by chemical vapor infiltration (CVI) method. Microstructure was investigated using transmission electron microscopy (TEM), with texture information obtained with transmission Kikuchi diffraction (TKD). It was found that both matrix and fibers are nanocrystalline, with the preferred grain growth direction in the matrix being \u3c111\u3e, and no texture present in the fibers. Extensive twinning was found in both matrix and fiber materials. Temperature effects were investigated using high-temperature nanoindentation performed in vacuum at the temperatures of up to 700ºC, and cantilever testing at the temperatures of up to 300ºC. Measured values of hardness had a clear trend of decease with the increase of temperature – from 45 to 20 GPa. In the same temperature range Young’s modulus decreased from 450 to 300 GPa. On the other hand, fracture toughness of fibers and matrix doesn’t change significantly, but that of the interphases dramatically drops from ~0.8 to ~0.15 MPa*m1/2. At the same time, the character of fracture changes in the interphase as well – unlike fiber and matrix at the elevated temperature, it features essentially ductile behavior. Radiation damage was introduced via Si ion irradiation, at the temperatures of 300 and 750°C, up to the damage level of 2.6 dpa. Ion irradiation didn’t lead to noticeable hardening neither of matrix, nor of fiber material. This is in contrast to the behavior of simultaneously irradiated single-crystal SiC, which showed a noticeable increase in hardness. Young’s modulus at the same time decreased slightly. Fracture toughness increased in all the constituents (interphase, matrix and fiber) following irradiation, with the trend towards progressive increase with the increase of irradiation dose. More significant changes of properties in the composite compared to single crystal material was explained by the nanocrystalline nature of the composite constituents, providing high density of sinks for radiation-induced defects. On the other hand, increase of toughness is attributed to the radiation-induced near-surface stresses. These findings are discussed in relation to their impact on the further development of SiC-SiC composites for extreme environments applications, together with the perspectives of further development of implemented methodology for such studies

Elevated temperature nanoindentation and in-situ SEM mechanical testing of uranium fuels

Due to the Fukushima nuclear accident there has been a large effort by several countries to develop accident tolerant fuel forms for commercial light water reactors. A challenge with the current UO2 fuel is its low thermal conductivity which leads to higher center line temperatures in the fuel. New nuclear fuel forms are looking to increase the thermal conductivity and other thermophysical proprieties while also maintaining adequate mechanical properties and uranium loading. The elastic modulus, fracture toughness, and creep properties of the fuel are important for modeling the pellet clad mechanical interactions during operation of a nuclear reactor. During the operation of a nuclear reactor the cladding material creeps down and fuel pellet swells which leads to physical contact between the two. The pellet clad mechanical interactions can lead to potential cladding failures and release of radioactive material. The advanced fuel forms that are under consideration for replacing UO2 in commercial light water reactors is UN, U3Si2, composite UO2 and UO2 with additives. The composite UO2 is looking to increase the thermal conductivity with different additions and the UO2 with additives are intended to increase the grain size of the UO2. The increase in grain size can reduce the release of fission gas products into the plenum of the cladding rod improving the operational lifetime of the fuel. While there is a large amount of work on the thermal properties of these accident tolerant fuel forms the literature is quite sparse on the mechanical properties necessary for modeling such interactions as the pellet clad mechanical interactions. Please click Additional Files below to see the full abstract

Micromechanical testing of ion-irradiated ferritic/martensitic steels

Ferritic/martensitic steels are the leading candidate material for structural components in the design of Gen IV reactors. They are known to exhibit high mechanical strength, ductility and toughness in the unirradiated condition and show good irradiation resistance with respect to void swelling, irradiation creep, irradiation-induced phase instabilities and high temperature helium embrittlement. Determining the mechanical properties of these structural materials after exposure to irradiation damage is essential for the safe design of the reactors. Testing neutron irradiated, bulk specimens is expensive and requires the use of a hot-cell, however, self-ion irradiation can been used as a proxy to emulate the irradiation damage caused in these materials. A disadvantage to using ion-implantation is that only a small volume of irradiated material can be achieved, hence micromechanical testing methods are required. In this work, a sample of T91 steel was irradiated using 70MeV Fe ions. Use of a high-energy accelerator provides a damage profile that extends to a depth of 6μm beneath the sample surface; a damage level of 20dpa is reached at approximately 5μm into the surface, before the Bragg peak. Although this is still a small volume of material, it provides ample material to perform micromechanical techniques including micro-cantilever bend testing and nano-indentation. Such experiments were performed both on the surface, and on cross-sections of the irradiated material. The poster reports data on results from nanoindentation experiments, both perpendicular to the irradiated surface and parallel in cross-section, as well as the yield stress measured from micro-cantilever testing. All experiments are performed in the irradiated and unirradiated regions of the sample

Room temperature and high temperature micromechanical testing of SiC- SiC fiber composites for nuclear fuel cladding applications

Silicon carbide ceramics are a candidate material for the use in nuclear power generation and are suggested to be used in novel accident tolerant fuel (ATF) cladding designs due to its favorable properties, in particular reduced (compared to Zircaloy) oxidation under accident conditions, good neutronic performance, high temperature strength and stability under irradiation. Due to its inherent brittleness, it is suggested to be used in the form of SiC-fiber reinforced SiC-matrix composite. In order to reliably model behavior of highly non-uniform and anisotropic composite materials the knowledge of the individual properties of fiber and matrix, and, crucially, the fiber-matrix interfaces, is required. In addition, nuclear fuel cladding materials are exposed to elevated temperatures during their operation, and therefore the understanding of the temperature dependences of the relevant properties is essential. Micromechanical testing techniques, such as nanoindentation and microcantilever beam fracture, allow determination of such localized properties, and can be implemented in the wide range of temperatures. Please click Additional Files below to see the full abstract

A comprehensive study on the deformation behavior of ultra-fine grained and ultra-fine porous Au at elevated temperatures

Modern design and engineering of highly efficient devices and machines demand innovative materials to satisfy requirements such as high strength at low density. The purpose of this study was to compare mechanical properties and deformation behavior of ultra-fine grained Au and its ultra-fine porous counterpart, both fabricated from the same base material. Microstructural investigations of the foam surrendered a ligament size of approximately 100 nm consisting of ~60 nm grains in average. The ultra-fine grained Au features a mean grain size of 250 nm. Nanoindentation is a convenient technique to obtain materials properties at ambient but also at non-ambient conditions and elevated temperatures. In this work, a broad indentation test series was performed in order to determine hardness, Young’s modulus, strain-rate sensitivity, and activation volume between room and elevated temperatures up to 300 °C for both materials. Due to the small characteristic dimensions, high hardness values were noted for both materials, which rapidly drop at elevated temperatures. In addition, an enhanced strain-rate sensitivity accompanied by low activation volumes was determined, increasing with elevated temperatures for both states. This can clearly be associated with interactions between dislocations and interphases. Moreover, for ultra-fine porous Au, a considerable increase of hardness was observed after annealing, which potentially can be attributed to starvation of mobile dislocations not occurring in the ultra-fine grained state. Cross-sections of indentations in ultra-fine porous Au combined with quantitative analysis of the resulting porosity maps allow visualizing the occurring deformation of the foam properly, showing distinct differences for tests at varying conditions. While the as-fabricated material exhibits distributed plasticity underneath the indent, this changes to strongly localized failure events in the annealed condition. At increased temperature, the deformation morphology reverts to more distributed deformation favored by the additional thermal activation

MATERIAL ISSUES FOR CURRENT AND ADVANCED NUCLEAR REACTOR DESIGNS

In all engineering applications, design and materials together determine the functionality and reliability of a device. This is particularly important in nuclear systems where the materials are pushed to their limits and phenomena not present anywhere else occur. In nuclear systems a combination of high temperature and pressure, stress, corrosive environment and high radiation environment combined causes significant materials challenges. Majority of commercial LWRs today are licensed for 40 years of operation, but many of them undergo lifetime extension to 60 or possibly 80 years. Materials degradation has always been a significant issue. However, due to the lifetime plant extension, finding materials that could sustain prolonged exposure to these extreme conditions has become a significant problem. In addition to the materials challenges in current LWRs, advanced reactors usually deal with even more difficult issues due to their operational requirements. Unusual heat transport media, such as liquid metals, liquid salts or other types of coolants, lead to a whole new set of material challenges. While corrosion has been the main issue, much higher operating temperatures create additional difficulties. In this paper, we present an overview of materials issues for current and advanced nuclear reactor designs.In all engineering applications, design and materials together determine the functionality and reliability of a device. This is particularly important in nuclear systems where the materials are pushed to their limits and phenomena not present anywhere else occur. In nuclear systems a combination of high temperature and pressure, stress, corrosive environment and high radiation environment combined causes significant materials challenges. Majority of commercial LWRs today are licensed for 40 years of operation, but many of them undergo lifetime extension to 60 or possibly 80 years. Materials degradation has always been a significant issue. However, due to the lifetime plant extension, finding materials that could sustain prolonged exposure to these extreme conditions has become a significant problem. In addition to the materials challenges in current LWRs, advanced reactors usually deal with even more difficult issues due to their operational requirements. Unusual heat transport media, such as liquid metals, liquid salts or other types of coolants, lead to a whole new set of material challenges. While corrosion has been the main issue, much higher operating temperatures create additional difficulties. In this paper, we present an overview of materials issues for current and advanced nuclear reactor designs