Component materials in a nuclear energy system are exposed to the bombardment of energetic particles. Material's performance is usually degraded by irradiation, which is commonly supposed to result from irradiation-induced microstructural changes, such as the formation of point defects, dislocation loops, voids, solute precipitation and segregation, etc. It is, therefore, critical that a clear understanding of these radiation damage processes be achieved, not only for prediction of a given material's response to irradiation but also for development of new, advanced materials used for a sustainable nuclear energy system. A physical description of the complicated, non-linear radiation damage process involves a wide variety of time- and length-scales, from ballistic binary collisions to collective atomic motion in the thermal spike stage, followed by the thermal activation process. To understand this, therefore, an important attempt is to evaluate the process using various experimental and computational techniques, since the multiscale process cannot be evaluated by only a single method. Based on the results obtained by the individual techniques, a model description of the process is then constructed for each scale. Another attempt to be made is to bridge different scales between the models, where great care should be taken when bridging models, because the degrees of freedom of a system are generally different depending on scales. In the present study, Monte-Carlo (MC) simulations have been done to investigate the nucleation process of defect clusters in materials during irradiation. A part of the degrees of freedom of a system was here given to an effect of the stochastic fluctuation of defect fluxes in materials. Otherwise, the nucleation process cannot be simulated. The energetics of defects in materials were firstly obtained by the classical molecular dynamics (MD) and ab-initio calculations, which was later used at the MC calculations. Successfully, our simulations described the nucleation processes, with a better precision beyond the conventional steady-state nucleation model. It was here found that temperature for a void to nucleate has both the upper and lower limits, and that the peak temperature for void nucleation increases with increasing damage rate (dpa/s).

We define an effective temperature that describes the structure of a disordered material far from equilibrium, with a Boltzmann-like distribution. We find, in a simple molecular dynamics model, that under radiation, a cold material transforms from an initial state, which has a certain initial effective-temperature that to a steady-state structure with a different, but still well-defined effective temperature. We find that the value of the steady-state effective temperature depends on the temperature of the environment and on the intensity of radiation to which the material is exposed. We provide a theory that uses the effective temperature as a basic ingredient, and describes the rate of amorphization and the final effective-temperature (and thus the structure) to which the material decays at long times.

Displacement cascade damage and helium production by transmutation reactions will result from the intense neutron irradiation in the structural materials of future fusion and generation IV reactors [1,3]. It is a major concern since the combined effect of helium and cascades may induce strong embrittlement and swelling. In order to predict in-service properties of such materials in the various radiative environments, the microstructural evolution under irradiation of model materials (Fe and Fe-Cr alloys) is studied using the JANNuS (Joint Accelerators for Nano-science and Nuclear Simulation) platform [4]. Single- and dual-beam ion irradiations (Fe and Fe/He) have been performed at 500°C on pure Fe and Fe-(5, 10, 14 wt %) Cr model alloys over a large range of doses. To achieve that purpose, in-situ and ex-situ experiments were combined to access low doses (up to 1 dpa) and high doses (up to 100 dpa) respectively. The damage evolution (dislocation loops, cavities/bubbles, �) during and after irradiation has been studied by transmission electron microscopy (bright field and weak beam conditions). The presentation will be mainly focused on He effects on the damage formation [5]. [1] J.L. Boutard, A. Alamo, R. Lindau and M. Rieth, C. R. Physique 9 (3-4) (2008) p.287. [2] N. Wanderka, E. Camus and H. Wollenberger, Mat. Res. Soc. Symp. Proceedings 439 (1997) p.451. [3] R. Schaeublin, D. Gelles and M. Victoria, J. Nucl. Mater. 307-311 (2002) p.197. [4] JANNuS facility website: http://jannus.in2p3.fr/. [5] D. Brimbal, B. Décamps, A. Barbu, E. Meslin, J. Henry, J. Nucl. Mater. 418 (2011) p.313.

Hydrogen and helium will be present in the materials used in the next generation of nuclear reactor designs, however either element may contribute to undesirable macroscopic phenomena including embrittlement and swelling. Additionally, there is experimental evidence to suggest that there exist synergistic effects between hydrogen and helium. Having a fundamental understanding of how these elements affect the microstructure will aid in their management. We investigate, through atomistic simulation, the properties and effects of these elements on body-centered-cubic iron. Using molecular dynamics and Monte Carlo methods, we simulate a variety of sizes of clusters containing hydrogen, helium, and vacancies. Simulations of small hydrogen-vacancy clusters reveal that hydrogen does have a stabilizing effect on vacancies; also, novel structures for hydrogen atoms about a monovacancy will be discussed. Helium is subsequently introduced into the clusters to determine its effect on the binding of hydrogen. The energetics and kinetics of these small bubbles are described and compared to experimental results.

A beam window made of annealed Inconel 718 alloy was removed from a beamline at the Isotope Production Facility at the Los Alamos Neutron Scattering Center. The window was in use for 5 years while being irradiated with a 100 MeV, 200 microamp proton beam under water cooling. The window was analyzed to determine the effect of proton irradiation on the mechanical properties of Inconel 718. The beam window was moved to a hot cell at the CMR facility and removed from the surrounding beam tube. Samples were machined from the window and thinned to 0.010� thickness. Shear and mechanical properties were measured via shear punch test at various locations on the beam window. A description of the materials handling, machining and sample preparation techniques will be presented along with results of the mechanical testing with respect to dose and irradiation temperature in comparison with previous results.

Very High Temperature Reactor (VHTR), which is one of the most promising concepts for the future of nuclear energy applications particularly for the process heating, evolves from technologies of High-Temperature Gas-Cooled Reactor. In Japan, High-Temperature Engineering Test Reactor (HTTR) has been operated since 1998. The VHTR fuel is being developed on the experience of the HTTR fuel, which achieved an extremely low failure fraction even after a sustained operation with the reactor outlet helium temperature of 950 °C. More stringent conditions, namely, higher burnup and higher fuel temperatures of VHTR have prompted R&D of two types of fuel concepts: (a) conventional but more robust Triso design, and (b) ZrC-Triso. As for the conventional Triso, modeling efforts on the palladium attack of SiC and the Ag-110m diffusion through SiC have been renewed. There, attention is given to peculiarities of the CVD SiC grain boundaries. The ZrC-coated particle fuel has shown promising results particularly for its durability at temperatures beyond the 1600 °C criteria for Triso particles. Further optimization efforts on the CVD ZrC coating have been directed to the precise analysis of C/Zr atomic ratio. Since the C/Zr atom ratio is measured with accuracy of the order of 0.01 at best, the application of positron annihilation spectroscopy is being studied for detecting minute free-carbon inclusions and/or pores, which influence the microstructural stability. Intensive characterization of the ZrC/pyrolytic carbon boundary with TEM has been also made in order to optimize the mechanical performance of ZrC-Triso particles.

The behavior of silicon carbide (SiC) under particle irradiation is of great interest due to its present and potential applications in nuclear reactors and other high radiation environments. Using Rutherford backscattering spectrometry it has been observed that pre-damaged SiC shows a significant decrease in disorder after suitable swift heavy ion irradiation. In this work, we use molecular dynamics (MD) simulations to study the annealing of defects in SiC by thermal spikes from swift heavy ions. The local heating due to irradiation with 0.87 GeV Pb ions (S_e = 33 keV/nm) is modeled using the thermal spike model and is subsequently used as input in the MD simulations. The simulations are performed with the PARCAS MD code and the Gao-Weber SiC potential. We assess the amount of recovery for varying initial defect concentrations and for a buried amorphous layer in the simulation cell, and we determine the maximum track radius inside which defects are annihilated.

Silicon carbide (SiC) is a key material for numerous applications in the nuclear energy field where it is submitted to irradiation by particles in a broad energy range. The aim of this work is to examine the effect of very-high energy (0.87-GeV Pb â?" Se~33 keV/nm) ion irradiation on the microstructure of both 6H and 3C-SiC single crystals pre-damaged by low-energy (100-KeV Fe) ion irradiation. The influence of the initial damage state (partial amorphization vs. full amorphization) and of the swift ion fluence is tested. For this purpose, three characterization techniques are implemented, namely Rutherford backscattering spectrometry and channeling (RBS/C), X-ray diffraction (XRD) and transmission electron microscopy (TEM). Results show that both SiC polytypes exhibit a very similar behavior upon irradiation in the nuclear energy loss regime, i.e. irradiation rapidly (at ~0.4 dpa) leads to amorphization. On the contrary, swift heavy ion irradiation does not induce disorder, but allows a recovery of the damage generated at low energy. This finding is deduced from the decrease of both the disorder level and the lattice strain measured by RBS/C and XRD, respectively. A dramatic effect of the initial damage state is clearly observed by RBS/C: (i) recrystallization takes place at the buried amorphous/crystalline interface in the case of crystals that were completely amorphized, while (ii) this damage recovery occurs over the entire damaged thickness for partially amorphous crystals (i.e crystals where amorphization just initiated at the damage peak). In this latter case, the analysis of the recrystallization rate allowed estimating an effective recrystallization track-diameter that is found to be ~1.6 nm. Using this value, the recrystallization cross-section has been evaluated and the recrystallization kinetics has been derived.

Ionic and covalent materials play important and diverse roles in nuclear applications, with examples which include nuclear fuels, ceramics for encapsulating nuclear waste and diagnostic materials for fusion experiments. The structural integrity and function of these materials are generally affected by defects therefore their performance gradually deteriorates in a radiation environment. Ionic and covalent materials tend to be much more sensitive to radiation damage than metallic materials. Radiation imparts energy to the atomic nuclei, to the electrons or to both the nuclei and the electrons. Energy imparted to the atomic nuclei results in displacement damage (Frenkel defects and clusters) and lattice heating (phonons). The effects of energy imparted to the electrons are varied and complex. In metals the electronic energy diffuses through the material and redeposits in the lattice via electron phonon coupling. The presence of a band gap results in a range of effects in ionic and covalent materials. Molecular dynamics (MD) in general, and cascade simulations in particular, have made an enormous contribution to the understanding of damage processes. However such simulations use classical interatomic potentials and therefore they neglect the effects of excited electrons. We have recently developed a methodology for coupling a two temperature model (2TM) for the electronic system to MD which includes the effect of energy storage and redistribution by electrons1. The 2TM model is, however, inappropriate for band-gap materials, due to the variable number of free carriers. In this presentation we will describe an extension of the coupled 2TM-MD methodology that is appropriate for band gap materials2. We will present results for the damage produced by swift heavy ions in silicon and make a comparison with equivalent simulations using the standard 2TM-MD model, to emphasize the significance of the variable carrier density. We will also present results for radiation damage in SiC, which is a potential candidate material for next generation fission and fusion reactors. 1. D.M.Duffy and A.M. Rutherford, â?oIncluding the effects of electronic stopping and electron-ion interactions in radiation damage simulationsâ? J. Phys: Cond. Matt. 19, 016207 (2007) 2. S.L. Daraszewicz and D.M. Duffy â?oExtending the thermal spike model for insulating materialsâ? Nucl. Inst. Meth. B269, 1646 (2011)

Nanostructured materials provide the opportunity for tailoring physical, electronic, and optical properties for a variety of technological applications, including advanced nuclear energy systems. As the world increases its reliance on nuclear energy, there is an ever-increasing demand for radiation-tolerate materials that can withstand the extreme radiation environments in nuclear reactors, accelerator-based nuclear systems, and nuclear waste forms. Understanding radiation effects in nanomaterials is an urgent challenge, since it may hold the key to unlock the design of tailored materials for advanced nuclear energy systems. Silicon carbide (SiC) has outstanding electrical, thermal, and mechanical properties, which make electronic devices based on SiC superior to those based on Si for high-power, high-frequency and high-temperature applications. The excellent physical and chemical properties also make SiC a prominent candidate for use as a structural material in fusion and fission reactors, cladding material for gas-cooled and light water fission reactors, and an inert matrix for the transmutation of plutonium and other transuranics. Some classes of nanocrystalline materials have demonstrated enhanced resistance to radiation-induced amorphization due to absorption and annihilation of mobile point defects at the interfaces or grain boundaries. Understanding the role of nanoscale grain sizes on phase transformations in nanostructured SiC under ion irradiation has, therefore, significant implication in advanced nuclear energy systems. In the current work, the irradiation response of nanostructured SiC has been investigated using high-quality nanocrystalline 3C-SiC with sharp interfaces and grain boundaries. The objective of this study is to understand the response of nanocrystalline SiC films to ion irradiation, in particular with regard to defect production and damage accumulation relative to phase stability. The irradiation studies were performed at room temperature using MeV Si and 7 MeV ions. Complimentary characterization techniques were utilized, including Rutherford backscattering spectroscopy, glancing-incident angle X-ray diffraction, transmission electron microscope, and selected-area electron diffraction. The result indicates that heavy Au ions are more effective in producing damage or amorphous clusters than the Si atoms. The results also demonstrates that nanocrystalline SiC films with high quality grain boundaries are more radiation tolerant, and an order of magnitude higher dose may be required to amorphize these nanocrystalline films, as compared to single crystal SiC.

The interaction of interfaces with radiation-induced point defects is analyzed in terms of the atomic structure of the interface, defect energetics and kinetics, using atomistic modeling, reaction-diffusion model and ion irradiation experiments. Maximizing the area per unit volume of interfaces is shown to simultaneously reduce both the concentration and the flux of radiation-induced point defects. The radiation response of materials where greatest reductions in both (defect concentration and defect flux) are achievable, however, shows extreme sensitivity to the sink strength, of the interfaces. Design of materials for radiation resistance thus requires both a high interface area per unit volume and control of interface sink strength. The model predictions are compared with ion irradiation experiments on model interphase boundaries such as fcc-bcc (Cu-Nb) and metal-oxide (Fe-Y2O3). This research is supported by DOE, Office of Science, Energy Frontier Research Center.

Despite extensive study, a fundamental understanding of how point defects contribute to radiation hardening, swelling, and radiation induced segregation (RIS) in ferritic steels is still lacking. It is essential to develop a comprehensive understanding of the irradiation behavior of pure ferritic iron in order to ensure the reliable performance of more complex ferritic and ferritic/martensitic alloys being developed for future fusion and advanced fission reactors. The behavior of the complex microstructure of alloys such as HT-9 or ODS steel is essential to the safe operation of the reactor over its 30-60 year lifetime. To understand the contribution of the high sink density found in these alloys to the behavior of the material under irradiation, it is useful to study a dense network of grain boundary sinks in an otherwise clean microstructure. A systematic study of the aggregation of point defect clusters as a function of grain size in Fe nanocrystalline ferritic iron provides a means to study the effect of high sink density on the annihilation of point defects. This work presents the in-situ irradiation of free standing nanocrystalline Fe films to approximately 5dpa at 300°C using 1MeV doubly charged Kr ions at Argonne National Laboratory. Dislocation loops in nanocrystalline iron displayed strong size dependence with grain size, and an approximate minimum grain size for the prevention intragranular defect cluster formation was determined.

Despite the significant progress in the development of the ODS reduced activation ferritic steels for the future fission and fusion reactors, their ductility and fracture toughness remain still unsatisfactory. In this work, the possibilities of improvement these characteristics are reported, i.e.: via (1) a new thermo-mechanical treatment (hydrostatic extrusion followed by annealing) and (2) alloying with vanadium in the concentration ranging between 0.3â?"3% (in weight percent). Samples with the nominal composition of Fe-14Cr-2W-0.3Ti-0.3Y2O3 have been prepared using elemental (Fe, Cr, W, Ti) and Y2O3 powders. The powders were mixed and subjected to mechanical alloying in a planetary ball mill in hydrogen atmosphere, followed by degassing and HIPping at 1150oC, under pressure of 200 MPa for 3 h. Next, the samples were preheated at 900oC and extruded under hydrostatic pressure with the reduction ratio of ~4 followed by annealing at 1050oC for 1 hour. It has been demonstrated that such a processing is a promising method for increasing the density of sintered samples and refining their microstructure thus leading to a significant improvement of the mechanical properties of the as-HIPped ODS ferritic steels. In fact, the tensile strength at room temperature increased from 950 up to 1350 MPa. Also, Charpy impact upper shelf energy, significantly increased from 3.1 up to 6.2 J. However, ductile-to-brittle transition temperature (DBTT) remained relatively high (about 90°C). In order to further improve fracture characteristics of the ODS ferritic steel, it has been modified by vanadium addition. The 0.3% V-ODS steel exhibits the lowest DBTT of about 9.0°C. Higher V content (1â?"3%) does not result in a reduction of DBTT which remains at the level of about 100°C. However, 1-3% V-ODS samples show almost two times higher a lower shelf energy values in comparison with the 0.3% V-ODS and no vanadium ODS steels. These results are discussed in terms of microstructural features, in particular various oxide particles formed during processing.

A systematic study of a wide range of interface types is underway to determine how variation in interface properties such as misfit-dislocation density, excess volume, and misorientation affect the radiation induced defect absorption and recombination. In this study, the epitaxial thin films of metallic Cr, Mo, V and their alloys deposited on MgO(001) substrates were used as model systems. In this paper, we present our experimental results of application of heavy ion beam radiation to study the stability of a well ordered interface between a metal (Cr) and Cr-V alloy films and an oxide (MgO) substrate. A 100 nm thick Cr-V alloy films were epitaxially grown on an MgO substrate using the molecular beam epitaxy method. By controlling the composition of Cr-V alloys, the lattice mismatch with MgO can be adjusted so that the misfit dislocation density varies over a wide range. Ion irradiation experiments have been performed on these films at 300 K using 1 MeV Au+ ions over doses ranging from 0.4 to 300 dpa. The experimental conditions were selected in order to produce maximum damage levels near the film/substrate interface. The accumulation of damage in both film (Cr and V sublattices) and substrate (Mg sublattice) as well as the depth profiles of implanted gold ions have been investigated using Rutherford backscattering spectrometry in channeling and random geometries. Two stages of radiation-induced damage accumulation were observed in this system. In the first stage, damage increases very rapidly, and then, after a certain dosage, the rate of change of the damage appears to slow down until the high dosages are reached. However, the degree of disorder was far below the random level expected. Channeling results reveal that most of the implanted gold atoms substitute for Cr and /or V atoms in the film. These results show that the Cr-V-MgO interface appears to withstand high dose of irradiation (up to 300 dpa).

The performance of structural materials in nuclear applications (fission and fusion reactors) is strongly influenced by the presence of helium, which is produced in these materials by e.g. (n,α) reactions or direct injection into the near surface region of the first wall of a fusion reactor. As a consequence, helium interactions with these materials raise an important issue as their durability of the mechanical and confinement properties. The first step in the understanding of these interactions, which often lead to the deterioration of the structural properties, is the appreciation of the diffusion behaviour of helium through the material. However, owing to the very low solubility of this element and to its ability to be trapped by vacancy type defects and to form stable bubbles, its diffusion process, even in a simple single-element matrix, is a complex problem. This is all the more important as the effect of gradual radiation damage levels as well as materials aspects (microstructures, defects concentration, chemical heterogeneity�) must also be considered to approach the reality of the material application. Only few experimental data are available concerning the effect of microstructure and more precisely of grain boundaries density on helium behaviour. In that context, the aim of this study is to bring further elements of understanding in the diffusion mechanism of helium in well-characterised materials foreseen in the future nuclear plants such as iron-based systems. Polycrystalline Fe 99.95 % samples, with grain sizes of 2, 40 and 100 µm, have been implanted with helium at 8 and 60 keV energies. A complementary set of techniques was used to characterize helium/material interactions for the three microstructures previously-characterized by electronic microscopies. Using Thermal Desorption Spectroscopy (TDS), it was possible to get information about the nature and the structures of the He-trapping sites. Isochronal as well as isothermal annealing allowed discriminating between different diffusion mechanisms by which the helium migrated to the surface. Activation energies for every trapping site (mono-vacancies, clusters) have also been determined from conventional reaction model. In complement to this technique, 4He depth profiles implanted at 60 keV were deduced from Elastic Recoil Detection Analyses after implantation as well as after annealing treatments. Diffusion coefficients of helium in Fe-based materials were calculated using Fick's law derived models. Preliminary observations between the three microstructures highlighted that grain boundaries should play the role of path diffusion for He migration.

Molecular dynamics (MD) simulation is an important tool for gaining insights into radiation damage effects in a way which is often not possible in experiments. Systems of interest include materials used in nuclear and future fusion reactors as well as materials to be used for safe encapsulation of nuclear waste. High-energy radiation damage in these systems has not been studied so far, yet it is important to simulate. New energy and time scales will give a more realistic view of the phenomena that take place during the irradiation and possibly lead to observations of new effects. DL_POLY MD code, in combination with the massive parallel computing facilities of HECToR, UKâ?Ts National Supercomputer, set the stage for simulating systems with up to 1 billion atoms, and therefore open the possibility to model radiation damage due to high energy recoils in the range between 100 keV up to 1 MeV. We study the effects of high-energy radiation damage in several interesting nuclear materials, including iron and zirconia. We simulate recoil energies in excess of 200 keV in systems with over 100 million of atoms using a recently developed many-body potential. We investigate the dependence of the damage creation, evolution and recovery on time as well as the nature of the defects in the system.

Nuclear fuels experience unique microstructural evolution as a result of thermal and irradiation conditions, and fission that they experience during service. These result in a host of unusual processes including the generation, transport and release of fission gases; swelling due to fission product generation; formation of high-burnup rim structures; and component segregation. The ability to simulate these microstructural changes at the meso-scale would enhance the ability to understand, predict and control the engineering performance and service life of many different fuel forms in a variety of reactors. We present a model that combine two traditional microstructural evolution models, phase-field and Potts, to address all the evolution processes listed above. This hybrid Potts-phase field method is an efficient and effective method with many advantages. It can simulate microstructural evolution on a sufficiently large scale to provide engineering properties directly or generate constitutive models to inform continuum engineering scale models. It can couple multiple physical processes such as coarsening, diffusion, nucleation, recrystallization, phase transformations, and others that neither model can simulate alone, but combining them allows the hybrid to address all the processes simultaneously in a coupled simulation. Sandia National Laboratories is a multi-program laboratory managed and operated by Sandia Corporation, a wholly owned subsidiary of Lockheed Martin Corporation, for the U.S. Department of Energyâ?Ts National Nuclear Security Administration under contract DE-AC04-94AL85000.

The chemistry of nuclear reactor fuel is complex. A continuous loss of uranium and plutonium along with the formation of a broad range of new species due to the fission of UO₂ fuel adds to the overall complexity. The fuel ultimately contains multiple f-electron elements (actinides such as uranium, plutonium, americium, neptunium, and curium), as well as many lighter elements. These products affect a range of fuel properties, including mechanical stability, thermal conductivity, and microstructure evolution. In this study we have used atomistic simulation methods to examine the properties of various mixed actinide oxides. In particular, we have used first-principles calculations and empirical potentials to investigate the structure and bulk modulus of binary (U, Pu)O₂, (U, Np)O₂, (Th, Pu)O₂, and (Pu, Np)O₂. A comparison of variation in lattice parameter and elastic properties will be discussed with the results compared to available experimental data.

The current paradigm for nuclear fuels is the enclosure of Uranium Oxide ceramic within a thin Zircaloy clad. Under normal operating conditions this fuel system has been systematically improved resulting in an impressive performance record of less than 1 failure per million as defined by the release of fission product gas from the cladding to the reactor coolant. However, under accident situations significant issues can occur due to the combination of the relatively weak and reactive Zircaloy cladding and the inventory of fission products and fragmented fuel released when this single containment is breached. The purpose of this paper is to review progress in an applied materials science development program aimed at engineering a microencapsulated fuel for light water reactor (LWR) applications. Specifically, fuels are under development that mitigate fission product release through multiple materials barriers. The process involves the design of an tri-structural isotropic (TRISO) fuel specifically purposed for light water application where fission products are retained within a low-density pyrolitic carbon buffer layer and the ensuing pressure is withstood by a SiC micro-pressure vessel. These fuel spheres are on the order of 1 mm in diameter which are then compacted through a nano-powder transient eutectic process forming a ~40 volume percent TRISO fuel, ~60% dense SiC matrix which serves as a secondary barrier to fission product release. Optimized process conditions include balancing nano-powders of beta-SiC, alumina, yttria, and silica along with gadolinia (a neutron poison) with correctly adjusted specific surface and surface chemistry. Special attention was also required to avoid clustering of TRISO fuels and powders during processing. This composite fuel would then be a substitute for the UO2 fuel pellet within the current Zircaloy cladding though should inhibit fission product release in the event of cladding breach and additionally would not react with reactor coolant water. Through a transition away from evolutionary improvements in the current LWR fuel system, the attributes of this advanced fuel fundamentally alter the philosophy of fission product retention in LWRs and the associated regulatory requirements.

Irradiation alters the local stoichiometry of oxides significantly. The resulting stoichiometric changes play a critical role in the dynamics of defects and microstructure evolution in oxides under irradiation. Stoichiometry in oxides is also sensitive to the surrounding oxygen environment. In general, the levels of point defects and electronic charge carriers in an oxide are sensitive to the oxygen partial pressure in contact with the oxide at hand. We investigate the electrochemistry of defects in UO2 under irradiation, where both the atomic displacements by energetic collision cascades and the exchange of oxygen with the ambient drive stoichiometric changes in the material. The problem is cast in the form of balance laws of lattice and electronic defects under defect generation and diffusion, with boundary conditions dictated by the oxygen partial pressure at the free surface. Inherent to this problem is the electrostatic field resulting from the segregation of charged lattice and electronic defects in the material. Using this model, the scenario of dynamic stoichiometric changes in a UO2 film under ion irradiation will be illustrated in detail. This research was supported as a part of the Energy Frontier Research Center on Materials Science of Nuclear Fuel funded by the U.S. Department of Energy, Office of Basic Energy Sciences under subcontract # 00091538 from INL to Florida State University.

Metallic nuclear fuel (U-Zr) exhibits swelling and formation of inert gases during burn-up cycle. We use molecular dynamics simulations to understand the energetics of the vacancy and vacancy cluster formation and arrangement using a recently developed modified embedded atom method interatomic potential for Uranium. First, we vary the number of vacancies in pure bcc Uranium from 1 to 10, and calculate the formation energy for these vacancy clusters. Second, we calculate the correlation between swelling and number of vacancies. Also, we investigate correlation between transition from crystalline to amorphous state with vacancy concentration. This work provides fundamental insight with regard to the swelling of bcc-U based nuclear fuel.

CRUD is an acronym for Chalk River Unidentified Deposit and refers to corrosion products that deposit on internal LWR reactor components, specifically on the upper parts of fuel rods where sub-cooled nucleate boiling occurs. The major CRUD components are Ni, Fe and Cr oxides, which originate from corrosion of, e.g., steam generators. The porous CRUD further attracts B containing precipitates, which, due to their neutron absorbing properties, shift the power distribution along the rod axis and lead to so-called Axial Offset Anomaly (AOA). In order to better understand CRUD formation and growth we have performed density functional theory (DFT) calculations on a number of the Ni, Fe and Cr oxides as well as B containing phases. We will summarize results for the thermodynamic properties of NiFe2O4, ZnFe2O4, NiCr2O4, Ni1-xZnxFe2O4 and NiCr2xFe2-2xO4 spinel phases, including point defects governing the formation of non-stoichiometric compounds. Relevant unary and binary phases have also been investigated in order to facilitate prediction of phase equilibria and the development of thermodynamic models of CRUD. Based on the calculated thermodynamic properties of ternary spinels, binary oxides and unary phases, the stability of different known B containing phases is predicted as function of the chemical environment. The DFT derived thermodynamic data is validated against existing thermodynamic databases. Examples of integration of DFT results with CALPHAD models will also be given. This work was supported by the Consortium for Advanced Simulation of Light Water Reactors (CASL) program of the US DOE Office of Nuclear Energy.

In order to assess safety aspects characterizing the nuclear fuel during its operational life, both in-pile and after discharge, it is essential to be able to measure relevant fuel properties such as thermal transport, fission products distribution and behaviour, mechanical properties, and structure alterations. When performing laboratory investigations on spent fuel the effects associated with the accumulation of alpha-decay damage and helium occurring at low temperature during storage must be assessed in addition to the alterations caused by fission damage occurring during neutron irradiation in the reactor. Property evolution as a function of irradiation history and conditions must be studied at the microstructural and macroscopic level. This is even more important for fuels under extreme conditions; only by achieving full understanding of the macroscopic processes and of the underlying mechanisms it will be possible to design the appropriate predicting tools necessary for the safety of nuclear fuels in current&advanced fuel cycle concepts. An overview on the characterization methods used or under development in ITU, including key results from studies focused on the behaviour of nuclear fuels under off normal/extreme conditions, including simulated or actual severe accidents, is presented.

Understanding the thermal conductivity of UO2 is an important goal for both experimental and computational research on nuclear fuel materials. First-principles methods offer the potential to understand the intrinsic properties of fuels without assuming adjustable parameters and to predict performance of new materials. However, theoretical efforts to study the thermal conductivity of UO2 with phonon calculations have yet to establish a comprehensive theory consistent with various aspects of experimental observations. We recently proposed a new first-principles approach to UO2 combining an improved LDA+U method and analytical models to overcome the technical issues associated with calculating actinide compounds, including aspherical self-interaction errors and multiple local minima. The method is applied to calculate phonon properties of UO2. We found that the couplings between U and its 8 ligand O atoms are dependent the magnetic and orbital state of uranium f-electrons and give rise to artificial unstable phonon modes unless proper thermal average of U-O forces over magnetic and orbital degrees of freedom is considered. In particular, the phonon frequencies at the X point depend very sensitively on the orbital since these modes can be directly decomposed into local Jahn-Teller modes. The implication of the UO2 phonon modes on thermal conductivity is discussed.

Uncertainty in disposition plans for US spent nuclear fuel has reopened questions about the long term stability of fuel materials in various environments. Ideally, stability concerns raised for Yucca mountain could be avoided by siting a national repository in a reducing environment to minimize oxidative degradation of the UO2. Given a reducing environment we revisit a model for which the fuel could radiolytically generate localized oxidizing conditions. As a base model we consider the reaction kinetics of Christiansen, Sunder and Shoesmith[JALCOM 213/214 (1994) 93-99] applied to the heterogeneous system of UO2 dissolution at a solid-aqueous boundary. As expected, radiolysis products, increasing with dose rate, have a strong effect on the predicted oxidative dissolution rate. We find that these predicted rates are sensitive to the specifics of the radiolysis chemistry represented. For instance, even the inclusion of a new slow reaction: H + H2O -> H2 + OH can change the UO2 dissolution rate by almost a factor of two. Of course uncertainty in model parameters and reaction mechanisms results in uncertain predictions. We consider a limited analysis to quantify the sensitivity of dissolution rate to model parameters. Results identify where model uncertainty can be reduced to have greatest benefit to model predictability. Additionally, results should help identify experiments (existing or proposed) that can best reduce the driving model uncertainties.

Although uranium dioxide (UO₂) is the most common variety of nuclear fuel, its poor thermal conductivity causes both steep temperature gradients and high center-line temperatures fuel pellets during a reactorâ?Ts operation. Thermal stress caused by the large temperature gradient results in either cracking in a low temperature region or plastic deformation in the high temperature region. In a loss of coolant accident (LOCA), the Zircaloy cladding temperature is increased rapidly due to the high centerline temperature of the fuel pellet leading to significant reactions with water. ZrO₂ is produced on the cladding surface, decreasing heat conduction and causing cladding rupture. Hydrogen gases, produced by the reaction, increase the internal pressure of the reactor and may cause an explosion. The idea of incorporating high thermal conductivity material into a UO₂ pellet has been suggested, and silicon carbide (SiC) is a possible material to form routes through which heat would disperse in the fuel matrix. In our preliminary studies, however, the results revealed that poor sintering behavior hindered the densification of SiC-UO₂ pellets which would lead to low thermal conductivity. The first intent of this study is to establish sintering strategies, which will overcome this obstruction to fabricate high density SiC-UO₂ pellets. The second intent is to investigate the effect parameters such as SiC particle morphology and amounts on thermal conductivity. In this work, 5 and 10 vol% beta SiC powder and whiskers were mixed and sintered together with UO₂ powders. Hyper-stoichiometry and spark plasma sintering (SPS) were employed to fabricate high density SiC-UO₂ pellets. 5 duplicated pellets were produced at each condition and characterized. The Archimedean Immersion Method and Laser Flash Method were used to investigate density and thermal conductivity. SEM was used to investigate the grain size and SiC dispersion of fabricated pellets. Adhesion of SiC and UO₂ were also investigated thoroughly at the high magnification using high resolution FE-SEM. Hardness and toughness were evaluated through Vickers micro indentation and ultrasonic measurement. Initial results using hyper-stoichimetry and SPS showed that the density was dramatically enhanced. However, because pressure was required for good adhesion of the two materials, we could only obtain higher thermal conductivity from the pellets made by SPS. The increase of thermal conductivity was up to 35% with 10vol% SiC. The evaluated UO₂ grains in SiC-UO₂ pellets were almost twice smaller than those in pure UO₂ pellets due to the pinning effect of grain boundaries. Higher hardness and toughness values of the pellets were reflected by the small grain size. The result of this research advances our understanding of SiC-UO₂ composites and provides valuable insight for sintering strategies to fabricate high thermal conductivity nuclear fuel.

Interfaces play an important role in materials design and understanding radiation effects of these interfaces is essential to develop new materials for many energy and environment applications. Recently, there have been considerable efforts in conducting systematic studies to determine how variation in interface properties such as misfit-dislocation density, excess volume, and misorientation affect radiation induced defect absorption and recombination in a wide range of interface types. For example, it has been shown that some nanostructured materials including Cu/Nb multilayer thin films are radiation resistant during high dose light ion irradiations [1, 2]. However, the studies performed in our group on Al/Ti multilayer films show that the resistance of these interfaces to heavy ion irradiation is significantly lower in comparison to the He ion irradiated Cu/Nb multilayers. We have also investigated the radiation effects on metal/oxide interfaces of epitaxial thin films of metallic Cr, Mo and their alloys deposited on MgO(001) substrates in our group. We have shown that these interfaces help in moving the defects generated in the films towards the film surface. Although the experimental conditions were selected in order to produce maximum damage levels near the metal film/oxide interface, the results demonstrate that the degree of disorder was far below the levels expected for high dose irradiations (up to 300 dpa). Results from these material systems will be discussed along with the other relevant findings in the literature for these types of material systems. [1] X. Zhang, N. Li, O. Anderoglu, H. Wang, J.G. Swadener, T. Hochbauer, A.Misra, R.G. Hoagland, Nucl. Instrum. Methods Phys. Res., Sect. B 261, 1129 (2007). [2] M.J. Demkowicz, P. Bellon, and B.D. Wirth, MRS Bulletin 35 (2010).

The irradiation behavior of numerous candidate ceramics to immobilize actinides have been studied over the past few decades using both short-lived actinides, such as ²³⁸Pu and ²⁴⁴Cm, and energetic ion beams, with ions ranging from He to Pb and Au. In the case of Gd₂Ti₂O₇ pyrochlore, the ion beam data for a range of ion masses correctly predict the amorphization dose in ²⁴⁴Cm-doped Gd₂Ti₂O₇ at ambient temperature, and the critical temperature for amorphization is independent of ion mass, which is a possible criterion for correctly predicting the temperature dependence of actual actinide waste forms at much lower dose rates. However, in many other materials, such as other pyrochlore, zirconolite, perovskite, apatite and zircon structures, either the dose for amorphization at ambient temperature or the critical temperature for amorphization, or both, exhibit a significant dependence on the irradiating ion mass. In some of these cases, the dose for amorphization in materials containing short-lived actinides is a factor of 2 or 3 less than that predicted by ion-beam irradiation. The observed shifts in critical temperatures for amorphization of up to several hundred degrees (K) with ion mass creates a more crucial dilemma in deciding which temperature dependent data can be used, if any, to predict the temperature dependence of actinide-bearing waste forms. In some cases, the shift in dose and temperature dependence with ion mass can be related to the ratio of electronic to nuclear energy deposition. In other cases, the mechanism for amorphization may be more complex than a simple ballistic collision process. These existing data will be critically reviewed, and strategies for applying ion-beam irradiation methods to predict behavior in actinide waste forms will be discussed.

Ever since the pioneering speculations in the 1960's about the multiple effects that a collision cascade produced by an energetic particle would induce in a solid target, there has been a huge interest in understanding the complexity of this highly non-equilibrium many-body electron-ion process in detail. In order to describe radiation damage in condensed matter, a tremendous amount of research effortâ?"using both experimental techniques and computer simulationsâ?"has been fueled. Thereby, understanding the electronic stopping power, as the primary source of deposited energy, is of extreme significance for designing materials that withstand high levels of radiation damage caused by high-velocity ions (projectiles) produced by primary sources or by collision cascades. However, the vast majority of computational research in material science, including the field of radiation damage, has been done within the adiabatic Born-Oppenheimer approximation. By assuming that the electrons adjust instantaneously to moving ions (e.g. by remaining in the ground state), this approximation amounts to that the quantum dynamics of the electronic system is completely ignored. To overcome this drastic approximation, we present a newly-developed first principles approach to address the electron dynamics. Our approach is based on the real-time propagation of the time-dependent Kohn-Sham equations and shows an excellent scalability. Thus, it is suitable for large-scale simulations involving several hundreds of electrons. After discussing our method in some detail, we show an important application by investigating the stopping of fast H atoms in bulk materials, such as Al and Cu.

An ab-intio quantum mechanics theory was applied for developing a new interatomic potential for molybdenum that was verified by comparison with experimental data. The new potential was used to study the formation and time evolution of radiation defects, such as self-interstitial atoms (SIAs), vacancies, and small clusters of SIA and vacancies, as a first stage in the relaxation of a damaged lattice structure after generation of a radiation cascade by energetic Xe+ ion bombardment. Molecular dynamics models were developed to calculate â?" for the first time â?" the open surface sink strength and the diffusion coefficients of small dislocation loops (DL) containing from 2 to 37 SIAs. Interactions of small DL with SIAs and vacancies were simulated for the 1st time and the results show that rotation of SIA between the equivalent <111> directions is an important mechanism that significantly contributes to the calculated kinetic coefficients of diSIA formation and sia - DL association.

For certain irradiation conditions, all crystalline materials display strong dynamic annealing effects: radiation-generated point defects experience migration and interactions during irradiation. In such cases, bombardment with larger dose rates often results in enhanced disorder. An inherent problem arises in the emulation of neutron and radioactive-decay-induced damage using ion irradiation. A large discrepancy typically exists in effective dose rates between reactor operation or spent nuclear fuel storage conditions and ion irradiation experiments. A grand challenge lies in understanding time scales over which processes of defect evolution persist after the thermalization of collision cascades and the lengths over which defects diffuse for different materials and irradiation conditions. The answers are not well known even for the simplest and best studied materials and conditions such as Si at room temperature. Here, we propose a method to measure diffusion lengths and relaxation times of mobile defects that dominate the formation of stable post-irradiation disorder. A defect lifetime of about 10 ms is measured for Si at room temperature, independent of the average density of ballistic collision cascades. Our results also suggest that the flux effect in Si is caused not by nonlinearity in the damage production. Instead, the flux effect could be understood in terms of defect relaxation times and the particular type of stable defects that govern the order of the defect accumulation kinetics. These findings have important implications for development of predictive models of radiation damage buildup in solids. This work was performed under the auspices of the U.S. DOE by LLNL under Contract DE-AC52-07NA27344.

The long-term storage of nuclear wastes is a world-wide major environmental issue. In France, nuclear exhausted fuels are vitrified after a reprocessing step in nuclear plant facility in La Hague. Fission products and actinide solutions arising from reprocessing of spent uranium oxide (UOx) fuel are usually vitrified in specific sodium borosilicate glass matrices. This work is focused on the interaction of platinum group metals and particularly on ruthenium with glass matrices. Ruthenium oxide known to have a very low solubility in glasses crystallizes in the form of acicular RuO₂ particles in high-level waste containment glass matrices. These particles, responsible for significant modifications in the physicochemical behavior of the glass in the liquid state, are the subject of investigation. The chemical reactions and the mechanism of RuO₂ crystallization with specific acicular or polyhedral morphologies are described in a simplified radioactive waste containment glass. In situ high-temperature environmental scanning electron microscopy (ESEM) was used to follow changes in the morphology and composition of the ruthenium species formed by reactions between a simplified calcined precursor (RuO₂-NaNO₃) and a sodium borosilicate glass (SiO₂-B₂O₃-Na₂O) up to 1300°C. The microstructure of the intermediate compounds formed during these reactions was characterized by SEM, XRD and HRTEM, while their local structure was investigated by Ru K-edge EXAFS. This combined approach provides new information on the formation of phases containing ruthenium during the synthesis of the simplified glass system.

Although it is widely understood that water interacts extensively with vitreous silicates, atomistic simulations of the response of these materials to ballistic radiation, such as neutron or ion radiation, have excluded moisture. In this study, molecular dynamics simulations were used to simulate the collision cascades and defect formation that would result from such irradiation of silica in the presence of moisture. Using an interatomic potential that allows for the dissociation of water, it was found that the reaction between molecular water or pre-dissociated water (as OH- and H+) and the ruptured Si-O-Si bonds that result from the collision cascade inhibits a significant amount of the structural recovery that was previously observed in atomistic simulations of irradiation in perfectly dry silica. The presence of moisture not only resulted in a greater accumulation of non-bridging oxygen defects, but reduced the local density of the silica and altered the distribution of ring sizes. The results imply that an initial presence of moisture in the silica during irradiation could increase the propensity for further ingress of moisture via the low density pathways and increased defect concentration.

Swift heavy ions lose their energy predominantly through inelastic interactions with electrons, so-called electronic energy loss. The high density of energy transfer to the electrons along the ion path leads to local states of intense electronic excitation, a confined plasma-like state. Through various mechanisms, the excited electrons transfer their energy to atoms, for which collective processes drive the local structure far from equilibrium, resulting in complex structural modifications including defect formation, order-to-disorder transformations and amorphization. Energetic ions are an important tool to explore fundamental aspects of ion-matter interactions under a wide range of experimental conditions. This is relevant to nuclear materials to better understand their radiation performance with the aim to develop compounds with superior radiation resistance. This contribution focuses on the response of cubic cerium oxide (CeO₂) and orthorhombic calcium-zirconate perovskite (CaZrO₃) exposed to swift heavy ions. The irradiation experiments were performed at the UNILAC accelerator of the GSI Helmholtz Center for Heavy Ion Research in Darmstadt, Germany, using 1.7-GeV and 940-MeV Au ions, respectively. For CaZrO₃, the structure of a powder material was monitored as function of increasing fluence up to 1.5Ã-10¹³ ions/cm² by means of intermittent laboratory X-ray diffraction (XRD) measurements available in situ at the UNILAC-beamline M2. For CeO₂, a fluence series of up to 5Ã-10¹³ ions/cm² was prepared for synchrotron XRD investigation by irradiating several powder pellets enclosed in thin metal holders. According to XRD, as well as transmission electron microscopy (TEM), CaZrO₃ is completely amorphized at a fluence of 1.5Ã-10¹³ ions/cm². The integrated peak intensities as a function of fluence yielded a cross section for the crystalline-to-amorphous transformation with a corresponding track diameter of about 7 nm. This diameter was independently confirmed by high-resolution TEM. In clear contrast to CaZrO₃, no evidence for amorphization was found for CeO₂ exposed to the highest fluence. However, Rietveld refinement of synchrotron XRD patterns revealed a defect-induced unit-cell expansion. The accumulation of defects may also contribute to the pronounced broadening of diffraction maxima concurrent with a radiation-induced grain-size reduction. The very high structural integrity of cerium oxide under swift heavy ion irradiation dramatically changes when external pressure is included as additional parameter. Irradiation of cubic CeO₂ within a diamond-anvil cell at 20 GPa leads to the formation of an orthorhombic high-pressure phase. Interestingly, this transformation can occur without irradiation, but only at pressures well above 30 GPa.

Radionuclide decay in waste forms produces energetic beta particles, alpha particles, and gamma rays, as well as recoiling daughter products that are chemically different and whose change in ionic radius and valence state can significantly affect waste form structure. In the case of beta decay of the fission products 90Sr and 137Cs, substantial self-heating, chemical changes, and charge imbalance that can affect the structural stability of waste forms, leading to phase transformations and/or phase separation. In the work reported here, atomic-level models are developed that simulate these changes and are compared to experimental data for validation. The focus is on the decay of 90Sr2+ to 90Y3+ to 90Zr4+, in a model material that can also be studied experimentally. In this study, strontium titanate (SrTiO3 or STO) is used as a model material to simulate a waste form in which Sr decays to Zr. The model system was investigated with ab initio calculations to determine structures, total energies, and electronic states. The calculations were carried out with density functional theory (DFT) in the Generalized Gradient Approximation (GGA), applied to a periodic supercell representation of STO. Significant effort was applied to selecting calculation sets that would enable comparison to the experimental results. Formation energies and structures were determined for the full spectrum of isolated intrinsic and O and Zr-related defects possible in the STO+Zr system and the Zr-O defect complex. The theoretical results are compared to results from ion implantation experiments on STO. Sequential ion implantation was performed at 550 K to introduce 16O and 90Zr into the crystal structure. The elevated temperature was applied to avoid full amorphization of STO. To minimize the charge imbalance in the implanted STO, equimolar and overlapping implantation of 16O+ and 90Zr+ was performed up to a concentration of ~1.5 at.%. Thermal annealing at temperatures up to 1423 K for 10 hours in flowing Ar environments was also conducted. A number of characterizations followed, including Micro-XRD, HRXRD, RBS/C, PIXE/C, TOF-SIMS, and HRTEM. Experiments with injection of O and Zr atoms into STO qualitatively confirm the theoretical predictions of crystal structural changes. Discussion of the results, and a general assessment of the selected model waste form in terms of illustration of the development of a validated model will be provided.

Recycling of spent or used nuclear fuel may offer some significant advantages over direct disposal into a geologic repository in terms of materials recovery and volume reduction. Nevertheless, the environmental issues associated with the disposal of long-lived radioisotopes, such as technetium (t1/2 = 2.13 Ã- 105 a), are unlikely to disappear and may require the production of suitably durable nuclear waste forms. We conducted microanalysis with transmission electron microscopy (TEM) on a series of metallic alloy waste forms composed of Fe, Cr, Ni, Mo, and Zr with Pd, Tc, and Re added as waste elements. The hexagonal C14 phase was predominant for the (Fe,Cr)2Mo composition and the Laves phase C36 intermetallic was observed rather than the cubic C15 phase at the higher heat treatments for the (Fe,Ni)2Zr composition. This is consistent with the L + C14 to C15 transformation that occurs in these materials at ~500°C. The Pd2Zr phase was identified as a hexagonal close-packed modulated structure with the C36 structure. The remaining material was ferrite (bcc iron). Technetium and rhenium resided principally in the intermetallic phases in the respective waste form alloys.

High-porosity ceramic nanofiber membranes are developed and studied for separation and sequestration of gasified fission and poison radioelements related to the Modified Open Fuel Cycle. The one step gettering and sequestration of volatile fission and poison nuclear waste provides a simplified and universal waste form, while allowing for majority of the fuel to be recycled back into the fuel cycle. Fiber production methods including electrospinning and forcespinning have been applied to inorganic material processes using sol-gel and chelate chemistry. Fabrication strategies and parameters for development of high surface area membranes are explored, and surface functionalization for sensitizing and remediation of stimulant streams for radionucleotides of interest is demonstrated. Stability testing of the loaded fiber membranes in encapsulation glass waste forms also is demonstrated. Sandia National Laboratories is a multi-program laboratory managed and operated by Sandia Corporation, a wholly owned subsidiary of Lockheed Martin Corporation, for the U.S. Department of Energyâ?Ts National Nuclear Security Administration under contract DE-AC04-94AL85000.

A summary of experimental observations on the irradiation damage tolerance of so-called nanostructured ferritic alloys (NFA, aka ODS steels) is presented for a large body of recent irradiations results, including in situ He injection experiments. NFA contain an ultrahigh density of nm-scale oxides, along with high dislocation densities and fine grain sizes. These features provide a suite of outstanding properties and potential for remarkable irradiation tolerance. Thus three key questions are addressed related to: the effectiveness of NFA in: 1) managing high concentrations of He; 2) enhancing tolerance to displacement damage in the absence of high He concentrations; and, 3) forming bubble dominated microstructures to mitigate various manifestations of radiation damage. The experimental observations are analyzed and extrapolated to high dpa and a range of He/dpa using both detailed rate theory models and associated simplified scaling laws. The implications of these results to advanced fission and fusion energy are described.

Embrittlement of reactor pressure vessel steels may limit the operating lifetime of light water nuclear reactors. It is well accepted that in copper bearing steels, nanoscale copper precipitates, enriched with manganese and nickel, are the primary irradiation hardening feature. This led to the belief that steels with low or no copper would have a low sensitivity to irradiation-induced embrittlement. However, it was long ago predicted, and more recently experimentally shown, that nanoscale nickel-manganese precipitates can form in these steels in some composition and temperature regimes after a very large incubation fluence in RPV steels. These so-called late blooming phases could in some cases cause even more embrittlement in low or no Cu alloys than those containing copper rich precipitates. We show trends in atom probe data for variations in nickel-manganese for a number of different irradiated ferritic model alloys, as well as different irradiation conditions.

Environmental degradation of materials exposed to the hostile high-temperature water environments found in nuclear reactors is a major concern for life extension of existing reactors and design of new reactor systems. In particular, mechanisms controlling many localized corrosion and stress corrosion cracking processes remain poorly understood. High-resolution evaluations of corrosion and oxidation micro-to-nanostructures are critical to gain insights into underpinning degradation mechanisms. In this work, atom-probe tomography (APT) is used in conjunction with transmission electron microscopy (TEM) to investigate localized oxidation processes in both commercial Ni-base alloys (alloys 600 and 690) and model Ni-Cr binary alloys. In high-Cr alloys (e.g. model Ni-30Cr or alloy 690) exposure to high-temperature hydrogenated water results in intragranular oxygen penetration along dislocations while the grain boundaries that intersect the exposed surface exhibit dealloying and grain boundary migration without oxidation. In contrast, lower-Cr alloys (e.g. Ni-5Cr or alloy 600) exhibit preferential attack of grain boundaries without significant oxidation of matrix dislocations. Atomic-resolution observations have been performed at, and ahead of, the oxidation front revealing solid-state oxygen ingress and Cr-O rich cluster formation as a precursor to Cr-rich oxide formation.

Higher chromium alloy 690 has now replaced lower chromium alloy 600 for many critical components in pressurized water reactors (PWRs) due to its increased resistance to stress corrosion cracking (SCC). It is commonly accepted that this improved behavior is a result of a protective chromia layer formed on the alloy surfaces and crack walls. Unique, high-resolution microstructural observations have been made at SCC crack walls and polished surfaces of alloy 690 materials exposed to PWR primary water environments. A protective chromium oxide film does not form, instead penetrative oxidation has been discovered in the form of continuous oxide filaments (7-10 nm in diameter). This localized degradation appears to follow dislocation substructures and reaches depth of several hundred nanometers as the exposure is increased. The filamentous penetrative oxidation was characterized on cross-sections specimens using low voltage scanning electron microscopy (SEM), transmission electron microscopy (TEM) and atom probe tomography (APT). These filaments consisted of discrete, plate-shaped Cr2O3 particles surrounded by a distribution of nanocrystalline, rock-salt (Ni-Cr-Fe) oxide, both oriented to the matrix. Oxygen- and chromium-rich clusters are detected in the metal ahead of the filamentary oxidation by APT along with nickel-rich metal. This paper will highlight the use of high resolution SEM and TEM in combination with APT to better elucidate the microstructure/microchemistry of the oxide filaments and the mechanisms of penetrative oxidation.

Structural materials in nuclear reactors are required to withstand harsh environments due to imposed radiation damage, sometimes in the form of high rate helium (He) production. Limited solubility of He in metals, leads to bubble and void formation, often at interfaces â?" like grain and phase boundaries - which in turn can cause deterioration in structural robustness by embrittlment, void swelling and blistering. We report fabrication and mechanical characterization of 100nm-diameter bi-crystalline Cu(fcc)-Fe(bcc) nano-pillars, containing a single Cu-Fe interface, fabricated by templated electroplating, in comparison to He-irradiated Cu-Fe nano-pillars. In situ uniaxial nano-tension and compression experiments were performed in a custom-built nano-mechanical instrument, SEMentor, comprised of a scanning electron microscope (SEM) and a nanoindenter. Site-specific microstructural characterization by TEM was performed on individual pillars before and after deformation. Results suggest that Cu-Fe interfaces have Kurdjumov-Sacks orientation and have significant influence on mechanical properties. Interfacial strength, embrittlement, and deformability as a function of irradiation are discussed in context of nano-crystalline plasticity and defect microstructure evolution.