It is well established that interfaces and grain boundaries can act as efficient sinks for radiation-induced defects. Exactly how interfaces interact with defects and how this interaction depends on both the structure of the interface and the radiation conditions, however, are still uncertain. Here, we examine coherent heterointerfaces in oxide thin film bilayers to determine how radiation-induced defects interact with those interfaces and modify the radiation tolerance of the material. While these particular interfaces are often nearly fully coherent, with no thermodynamic trap states at the interface, the interface nevertheless greatly influences how the materials on both sides respond to the produced defects. Both enhancement and degradation of radiation tolerance is observed in experimental studies of model oxide heterocomposites. This behavior is rationalized using atomistic calculations and mesoscale simulations via which differences in chemical potential and bulk migration properties of defects in each phase are hypothesized to be the controlling factors. We identify different regimes of defect evolution in irradiated composites that may provide new opportunities for developing radiation tolerant nanocomposites. We discuss the implications for composites more generally, such as nanostructured ferritic alloys (NFAs), that have potential applications in nuclear energy systems.

Crystalline ceramics are one of a number of candidate materials for the immobilisation of radio-nuclides arising from the nuclear fuel cycle. In particular, ceramics have been suggested as the most promising option for containment of transuranics such as U, Pu and Am. Transuranic elements undergo decay by alpha particle emission and recoil of the parent nucleus. These recoil events causes disruption of the crystal lattice and after sufficient events many crystalline materials can be rendered amorphous. Little is known about the structure of the amorphised material and the subsequent effect on key wasteform properties. This research seeks to investigate radiation damaged in crystalline wasteform materials using a combination of spectroscopic techniques. Previous work by the authors has shown that the local environment of cations in titanate based ceramics changes significantly as a result of radiation induced damage, particularly the Ti, which was shown to change from six- to five-fold coordination. This contribution expands the study to investigate the behaviour in iron based materials specifically LaFeO3 and LaSrFeO4. Our approach involved heavy ion implantation of bulk ceramic samples, to simulate heavy atom recoil, combined with grazing angle X-ray absorption spectroscopy (GA-XAS) to characterise the resulting amorphised surface layer. Quantitative analysis was performed on the GA-XAS data to determine the change in valence and local co-ordination environment of cations in the amorphised surface layer.

The immobilization of the long-lived radiotoxic fission product I-129 into a durable waste form is important for effective nuclear waste management as iodine is highly mobile and has significant environmental and health concerns. The consolidation into a dense waste form without significant iodine loss is also a challenge due to the highly volatile nature of iodine. In this work, iodine bearing apatite Pb10(VO4)6I2 nanopowder was synthesized by high energy ball milling at room temperature with a high iodine loading (9 wt%). Dense iodine-apatite pellets were fabricated by spark plasma sintering (SPS) at 700 °C with a very short duration less than 3 minutes. The thermal stability of the iodine-bearing apatite powder and SPS densified pellets was studied by post-thermal annealing, and the iodine loss was investigated by thermal gravimetric analysis (TGA). The iodine apatite power is stable annealed at 200 and 300 °C with improved crystallinity and larger grain size. Phase decomposition occurred for apatite powder annealed at 400 °C, leading to significant iodine loss. In contrast, the SPS densified pellets are stable without phase decomposition or iodine loss at temperature up to 670 °C.
The radiation stability was investigated by energetic ion beam irradiations using 1 MeV Kr2+ irradiation under in-situ TEM observation. The as-milled Pb10(VO4)6I2 nanocrystals embedded in amorphous matrix can be easily amorphized at room temperature. The iodine-bearing Pb10(VO4)6I2 annealed at 300 °C exhibits enhanced radiation tolerance with a lower critical amorphization temperature (Tc) of 242 °C, as compared with lead/calcium vanadate fluorapatite (PbxCa1-x)10(VO4)6F2. SPS process further improves the radiation stability of Pb10(VO4)6I2 and the critical temperature for SPS densified pellets is reduced to 229 °C. The greater radiation tolerance of the iodine-bearing apatite is consistent with the enhanced crystallinity upon thermal annealing and SPS densification. These results indicate that SPS-fabricated Pb10(VO4)6I2 is a promising waste form for I-129 immobilization with greatly enhanced thermal and radiation stability and iodine confinement.

UO2 has been studied widely since it is the basis of conventional reactor fuels. It can be blended with other actinide oxide powders, in particular PuO2, to form what is commonly called mixed oxide fuel (MOX). Alternative fuel cycles are being studied based on other combinations, notably with ThO2, since the rate of higher actinide breeding is much lower. Furthermore, the higher actinides are problematic for waste forms as they often have long half lives, therefore, the incorporation of minor actinides such as AmO2, CmO2 and NpO2 with UO2 or ThO2 is desirable so that these species can undergo transmutation in a reactor or accelerator driven system. The difficulty in reproducing the many-body effects of these actinide oxides (such as the Cauchy violation) using an empirical pairwise description of ionic interactions was proven a stumbling block for previous atomistic simulation studies, particularly when investigating thermomechanical properties, such as bulk modulus, over a broad temperature range.
In this work we present a novel approach to simulating actinide oxides by including many-body effects using the Embedded Atom Method. This ensures a good description of a range of thermophysical properties (lattice parameter, bulk modulus, enthalpy and specific heat) between 300 K and 3000 K for AmO2, CeO2, CmO2, NpO2, ThO2, PuO2 and UO2. The oxygen-oxygen interactions are fixed across the actinide oxide series to enable the simulation of MOX fuels. The new potential is also used to predict Schottky and Frenkel pair energies.

The inherent grain-growth problem in nanocrystalline ceramic-oxides renders their highly attractive properties practically unusable, and controlling the nano-grain sizes continues to be an uphill task. We elucidate a framework to design dopant-pinned grain boundaries that prevent this grain growth. Using atomic simulations, we show that effective grain boundary pinning depends upon dopant-oxygen vacancy interactions, i.e., (a) dopant migration energy in the presence of oxygen vacancy, and (b) dopant-oxygen vacancy binding energy. Our prediction agrees with and explains previous experimental observations. This new concept is in complete contrast to the dopant-host atomic size mismatch concept prevalent in metallic systems, and elucidates that nanograin stabilizing concepts are not inter-transferable between metallic and ceramic-oxide systems.
This work was supported as part of the Materials Science of Actinides, an Energy Frontier Research Center funded by the U.S. Department of Energy, Office of Science, Office of Basic Energy Sciences. The computer simulations were performed at the National Energy Research Scientific Computing Center at Lawrence Berkeley National Laboratory.

The effect of transmutation of radionuclides, especially “short-lived” Sr-90 and Cs-137, to chemically distinct daughter products (Zr and Ba respectively) will impact nuclear waste form stability. Due to the technical challenges associated with this studying problem, the topic of transmutation has received limited attention during the past 30 years of waste form development. In order to develop a predictive capability to design radiation tolerant and chemically robust nuclear waste forms, we must first address a fundament materials science question: What is the impact of daughter product formation on the stability of solids comprised of radioactive isotopes? To answer this question, a multidisciplinary approach integrating first principles modeling with the synthesis and characterization of small, highly radioactive surrogate samples has been instigated. We present the details of this approach as well as recent results for a range of materials systems, including: ¹⁰⁹Cd_1-xAg_xS, ⁵⁵Fe_2-xMn_xO₃ and ¹⁷⁷Lu_2-xHf_xO₃.

Transmutation of constituents may offer a novel approach to synthesize compounds far from equilibrium conditions - a phenomenon we refer to as radioparagenesis. Especially in the cases isotopes that decay via beta - or electron capture, transmutation leads to significant changes in the valence and radius of the transmuting ion, often resulting in a daughter product that is incompatible with the original parent crystal structure. However, exploring these issues for “short-lived” fission products of interest is not feasible due to the ~30 year half-life, and previous experimental approaches to accelerate the process focused on former isotopes with shorter half-lives via neutron activation were inconclusive. In this work, we present a new accelerated chemical aging approach, which combines density functional theory calculations with experiments on isotopically pure samples to investigate radioparagenesis under well-defined. We will specifically present recent experiments using aberration corrected transmission electron microscopy (TEM), energy dispersive X-ray and electron energy loss spectroscopies to study the synthesis and characterization of 177Lu2O3. 177Lu decays via beta minus to 177Hf with a 6 day half-life. We present experimental results of the impact of Hf formation on the structural stability of bixbyite Lu2O3. These results are compared to complementary DFT calculations, which ultimately will allow for predictions of structural stability as function of compositional evolution.

The study of materials with unusual properties offers to provide new insights into structure-property relationships and promise in the design of novel composites with tailored properties. In this spirit, we have chosen to study β-eucryptite, a technologically relevant lithium aluminum silicate that exhibits negative thermal expansion (NTE), radiation tolerance and pressure-induced amorphization (PIA) under moderate applied pressures. These exotic physical properties make β-eucryptite suitable for various specific applications like heat exchangers, ring laser gyroscopes, and nuclear breeder reactors. Using density functional theory calculations [Narayanan et al., Phys. Rev. B 81, 104106 (2010)], we found that the linear compressibility of β-eucryptite along the c-axis is positive consistent with recent ultrasonic experiments, as opposed to a negative value reported by earlier direct measurements. More importantly, this finding indicated that the NTE behavior in β-eucryptite occurs due to tetrahedral tilting and cation disordering rather than elastic effects arising from negative compressibility. Recently, our reactive force field (ReaxFF) molecular dynamics [Narayanan et al., J. Appl. Phys. 113, 033504 (2013)] showed that at radiation doses below 0.21 displacements-per-atom or less, β-eucryptite retains its long-range crystalline order while exhibiting tetrahedral tilting, change in atomic coordination around Al/Si and disordering of Li atoms. Furthermore, upon thermal annealing, most of the under-coordinated Si-polyhedra formed during radiation regained their tetrahedral coordination via a mechanism involving tilting of Al-, and Si-centered polyhedra. Our metadynamics simulations based on ReaxFF revealed that β-eucryptite begins to amorphize under moderate pressure ~3 GPa close to empirically known transition pressure [Narayanan et al., submitted to Appl. Phys. Lett. (2013)]. We also identified the atomic scale mechanisms underlying PIA in β-eucryptite that consist of (a) progressive tetrahedral tilting that eventually results in change in O-coordination around several Al atoms (~41.7%) while keeping SiO₄ intact, and (b) spatial disordering of Li atoms forming Li-Li, Li-O and Li-O-Li linkages. We show that the atomic-scale processes in β-eucryptite induced by thermal, radiation, and pressure environments arise from the inherent flexibility of the three-dimensional network of corner-sharing AlO₄ and SiO₄ tetrahedra. These results will be discussed in the context of a possible trend between NTE, radiation tolerance and PIA under moderate pressure in flexible framework structures.

Gen IV nuclear reactor designs require materials that can withstand long term operation in extreme environments of elevated temperatures, corrosive media, and fast neutron fluences (E>1MeV) with up to 100 displacements per atom (dpa). Full understanding of irradiation response is paramount to long-term, reliable service. The Mn+1AXn phases have recently shown potential for use in such extreme environments because of their unique combination of high fracture toughness values and thermal conductivities, machinability, oxidation resistance, and ion irradiation damage tolerance. Herein we report, for the first time, on the effect of neutron irradiation of up to 0.5 dpa at 70°C and 700 °C on Ti3AlC2, Ti2AlC, Ti3SiC2, and Ti2AlN. Evidence for irradiation induced dislocation loops and their effect on electrical resistivity is also presented. X-ray diffraction refinement of the resultant microstructures is provided. Based on the totality of our results, it is reasonable to assume that the MAX phases, especially Ti3AlC2, are very promising materials for high temperature nuclear applications.

Iron phosphate glasses are being considered as an immobilisation matrix for high level nuclear waste (HLW), including minor actinides and plutonium residues, due to their high chemical durability and ability to incorporate diverse chemical compositions. Iron borophosphate glasses are of particular interest as the addition of boron, which has a high thermal neutron absorption cross-section, increases glass thermal stability and provides criticality control. Incorporated actinides undergo α-decay, resulting in the formation of α-particles (MeV He nuclei) and energetic (~100KeV) daughter recoil nuclei. Interactions between recoil nuclei and glass atoms results in atomic displacements which form collision cascades, potentially altering glass network polymerisation and cation valance states. Such modifications can affect glass durability and long-term performance as an immobilisation matrix. In this study, heavy ion implantation (e.g. 2MeV Kr or 2MeV Au irradiation) was used as an analogue for α-recoil damage. Iron borophosphate glasses, with nominal molar composition 60P2O5 - (40 - x) Fe2O3 - xB2O3 (x = 0, 10, 20) were irradiated at room temperature, producing a damaged region extending from the surface to a depth of approximately 1µm. To probe exclusively the damaged region, surface sensitive techniques were employed. The effects of simulated α-recoil damage were investigated by probing the speciation and valence of Fe, and by examining the glass structure. In this contribution, we report on structural and chemical modifications as a consequence of heavy ion irradiation, elucidated using Reflectance Fourier-Transform Infrared (FT-IR), Mossbauer, and X-ray absorption spectroscopies.

Predicting high temperature thermodynamic and kinetic properties of materials for nuclear applications from first principles remains a major challenge. Important properties of materials used in nuclear applications include their resistance to degradation and chemical corrosion. The corrosion of nuclear materials involves surface and interface reactions, electronic and ionic transport and the occurrence of a variety of phase transformations, all driven by extreme chemical and mechanical driving forces. First-principles statistical mechanical methods are now capable of predicting a wide variety of thermodynamic and kinetic properties as well as the couplings between chemistry and mechanics that determine the rate and mechanisms of degradation of structural materials. In this talk, we will describe how corrosion of Zr in aqueous environments can be modeled from first principles. The approach relies on the use of first-principles parameterized effective Hamiltonians that rigorously account for all relevant atomic and electronic excitations. A combination with Monte Carlo techniques allows the statistical mechanical prediction of finite temperature thermodynamic and kinetic properties relevant to corrosion processes in nuclear materials.

Given an optimized set of neutronic and mechanical properties, zirconium alloys play a very important role in the nuclear field, as fuel cladding and by default as a barrier against radioactive material release during used fuel storage. Zirconium hydride formed in normal operation and accident scenarios is a major concern, and in particular, hydrogen behavior during vacuum annealing of used nuclear fuel, in addition to other de-hydriding processes, is an area of significant interest.
We describe the hydrogen desorption behavior from zirconium hydride and zirconium metal in vacuum observed during coordinated experimental and modeling activities. A δ-zirconium hydride is produced in an oxygen-free tube furnace from Zircaloy-4. The resulting hydride phase and hydrogen concentration have been verified by x-ray diffraction, weight change and gas desorption. Subsequently, the kinetics of hydrogen during thermal processessing has been studied using Thermal Desorption Spectroscopy (TDS) to directly obtain the hydrogen desorption spectra of δ-zirconium hydride as a function of initial conditions under a pre-determined temperature profile. The TDS results have been analyzed and compared to a one-dimensional, two-phase moving boundary model coupled with a kinetic description of hydrogen desorption from a two-phase region of δ-zirconium hydride and α-zirconium. The model and experimental comparison demonstrates the ability to successfully reproduce the TDS experimental results, which validates the assumption of zeroth-order and second-order hydrogen desorption kinetics for δ-zirconium hydride and α-Zr, respectively.
This study provides fundamental insights into the behavior of hydrogen and zirconium hydride, in addition to demonstrating a modeling paradigm to predict the performance of the hydride fuel and the cladding failure under vacuum annealing of used nuclear fuel.

Fuel cladding is the first barrier for retention of fission products and nuclear fuel. Safety analyses of dry casks containing high-burnup light water reactor (LWR) fuel require measurement of cladding mechanical properties in order to better understand fuel behavior. Pre-storage drying-transfer operations and early stage storage expose cladding to higher temperatures and much higher pressure-induced tensile hoop stresses relative to normal operation in-reactor and pool storage. Under these conditions, radial hydrides could precipitate during slow cooling and provide an additional embrittlement mechanism as the cladding temperature decreases below the ductile-to-brittle transition temperature. As a means of simulating this behavior, hydrided Zircaloy-4 samples were fabricated at Oak Ridge National Laboratory (ORNL) by a gas charging method to levels that encompass the range of hydrogen concentrations observed in current used fuel. For low hydrogen content samples, the hydrided platelets appear elongated and needle-like, orientated in the circumferential direction. In addition, a hydride reorientation system was developed at ORNL to simulate the effects of drying-storage temperature histories. Mechanical testing was carried out by the ring compression test (RCT) method at various temperatures to evaluate the sample&’s ductility for both as-hydrided and hydride reorientation treated specimens. As-hydrided samples with higher hydrogen concentration resulted in lower strain before fracture and reduced maximum load. The trend between temperature and ductility was very clear: increasing temperatures resulted in increased ductility of the hydrided cladding. A single through-wall crack was observed for a hydrided sample having very high hydrogen concentration under ring compression testing, but fracture surfaces traversing in the circumferential direction were observed for samples having lower hydrogen concentrations (<300 wppm). Following hydriding, the as-hydrided samples were subject to radial hydride treatment using the hydride reorientation system under high pressure at high temperatures. A systematic radial hydride treatment was conducted at various pressures (hoop stress 60 - 150 MPa) and temperatures (300-400C) for the hydrided samples with H content around 200 ppm. Following the drying-storage simulation, microstructural examinations were conducted on hydride reoriented samples to determine the radial hydride transition pressure and temperature. The RCTs on the hydride reoriented samples were conducted and compared to the ductility data from as-hydrided samples.

Zirconium-based alloys are used as fuel cladding and in-core materials in light water nuclear reactors.¹ The corrosion resistance and mechanical stability depend on the chemical composition and microstructure of the alloy.² Nb substitution of Zr is suggested to impart a better corrosion resistance to Zr alloys.³ However, the chemical and structural reasons behind this behaviour are not clear. In addition, the quantitative and microscopic understanding of the initial oxidation kinetics and the oxidation state of Zirconium at this phase has been an outstanding challenge. Even though the initial oxidation of most metals could be described by the Cabrera-Mott theory⁴, whether the logarithmic law could apply to Zr oxidation is debated. This is because Zr dissolves oxygen into the metal phase while oxidizing the surface. In order to uncover the initial oxidation kinetics and the chemical, structural and electronic properties of the oxide, synchrotron x-ray photoelectron spectroscopy (S-XPS), in situ angle resolved x-ray photoelectron spectroscopy (AR-XPS), low energy electron diffraction (LEED), and in situ scanning tunneling microscopy and spectroscopy (STM/STS) have been used in this work. The experimental results were, in part, elucidated by density functional theory calculations of the Nb defects in tetragonal ZrO₂.
The initial oxidation kinetics for pure Zr and Zr-2.5%Nb were found to be similar, but a more stoichiometric oxide layer is found on Zr-2.5%Nb with respect to pure Zr, indicating the Nb have an effect on slowing down the oxygen transport through the oxide. The chemical content analysis showed Nb segragation near the oxide surface during oxidation. The results indicate that alloying Zr with Nb helps to form a more stoichiometric surface oxide which provides a better passivation barrier for further oxidation. The initial oxidation kinetics of single crystal Zr has been studied by SXPS at different temperatures (100K, 300K, 500K). The spectrum shows clear formation of Zr sub-valence states, answering a long-debated question of whether Zr 1+, 2+ and 3+ states are possible to form at the metal-oxide interface. A modified initial oxidation model has been proposed, by taking into account of the dissolving and diffusion of oxygen in the metallic phase. The initial oxidation on single crystalline Zr surface was probed by STM/STS. It was found that the adsorbed oxygen on the surface forms island-like clusters, then spreads laterally to cover the surface. This STM visualization serves to provide new structural information for the initial Zr oxide.
1 Cox, B. Journal of Nuclear Materials 336, 331-368, doi:10.1016/j.jnucmat.2004.09.029 (2005).
2 A. T. Motta, et al. Journal of ASTM International 5 (2008).
3 Yilmazbayhan, A. et al. Journal of Nuclear Materials 324, 6-22, doi:10.1016/j.jnucmat.2003.08.038 (2004).
4 Cabrera, N. & Mott, N. F. Reports on Progress in Physics 12, 163-184 (1948).

Zirconium has a low neutron capture cross section, and its alloys have high melting points and generally good corrosion resistance. Its alloys are therefore useful as a fuel cladding material for fuel in certain designs of nuclear reactors. The corrosion behavior of zirconium alloys follows a cyclic profile, where, after an initial period of reducing corrosion rate, it suddenly accelerates again with rates similar to the initial growth. The point at which this change occurs is called transition. This cyclic behavior can repeat more than once, and each cycle tends to correspond to about 2-3 µm of oxide growth. After some indeterminate number of transitions, the oxidation rate can take off, and continue at a very high rate - called breakaway. Ideally, low corrosion rates are required; however, the detailed mechanisms underlying transition and breakaway are not fully understood. There is therefore a requirement to understand these mechanisms to provide a more accurate prediction of oxide growth with time.
Positron annihilation spectroscopy is an efficient tool for studying vacancy-related defects in crystalline solids. Positrons can get trapped at negative and neutral vacancy defects, and at negatively charged non-open volume defects provided that the temperature is low enough. The trapping of positrons at these defects is observed as well-defined changes in the positron-electron annihilation radiation. The combination of positron lifetime and Doppler broadening techniques with theoretical calculations provides the means to deduce both the identities (sublattice, decoration by impurities) and the concentrations of the vacancies [1].
We present results obtained with depth-resolved positron annihilation spectroscopy on oxidized Zircaloy-4 samples. The positron data suggest that Zr sublattice vacancy defects are observed in the layers immediately below the surface, whilst closer to the metal-oxide interface O sublattice vacancies are more abundant, creating larger Zr-O vacancy complexes (clusters).
[1] F. Tuomisto and I. Makkonen, Defect identification in semiconductors with positron annihilation: experiment and theory, Reviews of Modern Physics, to be published.

EIS has often been used to investigate surface films on conductive substrates. In this application EIS is employed to characterize oxide films on Zirconium 2.5Nb CANDU Pressure Tubes (PT). Key to this application of the technique is the model used to fit the data, as well as the physical interpretation of fit parameters, In this regard a specific software application has been developed to fit, validate and provide a systematic interpretation of the barrier oxide film. The barrier oxide film is the oxide layer adjacent to the metal interface that is effective in slowing the ingress of water to the metal-oxide interface where oxidation and associated deuterium uptake occurs. Where as the oxide film on a PT can be in excess of 20 microns, the barrier film on Zr 2.5Nb is typically on the order of 1 micron. Experience has shown that the barrier oxide film consists of multiple independent dielectric response features, thus the application of Parallel (Independent) Electrical Dielectric Response Analysis was developed. Using the application, the impedance data is first fit using the appropriate number of responses and then the fit parameters are converted to physical attributes of each feature present in the spectra (path resistance, oxide thickness, fractal distribution) using a predetermined relative dielectric constant. Here, results are reported on CANDU PTs removed from service. Measurements were made remotely on active PT sections in Hot Cells. This work is presented to demonstrate the application of EIS and the PEDRA fitting technique rather than evaluating material for service. A website has been developed to introduce the PEDRA technique: www.eispedra.com.

The complex fuel-cladding chemical interaction (FCCI) between U, Pu-based fuels and Fe-based cladding at elevated temperatures was investigated using scanning/transmission electron microscopy (STEM/TEM), selected area diffraction (SAD), and X-ray energy dispersive spectroscopy (XEDS) techniques. Microstructure and phases formed prior to and during annealing of diffusion couples were examined using wavelength dispersive spectroscopy (WDS) in scanning electron microscope (SEM). Upon completion of initial examination, cross-sectional specimens were prepared from the identified interaction zone in focused ion beam (FIB) tool using a lift-out approach. This contribution will report results from ongoing work on interdiffusion between fuel constituents and cladding in various diffusion couples. The discussion on phase evolution will be based on existing equilibrium phase diagram and phase-segregation mechanisms.
This work is supported by the Fuel Cycle Research and Development (FCRD) program of US Department of Energy.

Hydrogen pickup in zirconium alloys poses a prominent challenge to the design of zirconium alloys for fuel cladding in water reactors. As early as 1960 a volcano-like relationship was identified between the fraction of hydrogen picked up and the 3d transition metals that are typically used to alloy zirconium. The peak of the volcano was found to be coincident with Nickel [1]. The origin of this clear trend is yet to be uncovered in order to design resistant alloys on physical grounds rather than pure empiricism based on large amounts of data.
To elucidate this behavior we adopt the view that protons generated from water splitting on the surface of the passive zirconium oxide layer experience two competing processes. The first is gaining electrons from the surface of the oxide and evolving as hydrogen gas molecules. This is ideally the desired outcome to avoid hydrogen pickup and embrittlement of the metal. The second is the incorporation of hydrogen to the subsurface of the oxide and subsequently reaching the underlying zirconium metal. The presence of transition metals dissolved in the zirconium oxide layer can alter the equilibria of charged defects in the bulk of the oxide and the reaction kinetics on the surface, to favor one or the other of the above two processes.
In this work we present a systematic study for the effect of the 3d transition metals on the defect equilibria in monoclinic zirconium oxide (M-ZrO2). By combining density functional theory calculations of the formation free energy of all point defects (native or due to the extrinsic dopants) with thermodynamic modeling, we constructed the Kröger-Vink diagrams for M-ZrO2 co-doped with a transition metal and hydrogen simultaneously. This analysis revealed the type and the charge state of the dominant defect in the space charge zone near the oxide-water interface. Preliminary results indicate that most of the alloying transition metals will produce negatively charged substitutional defects in the space charge zone, and these defects are charge neutralized by the protons adsorbed on the surface of the oxide.
To supplement the thermodynamic analysis with kinetic considerations, we are in the process of computing the activation barriers for the hydrogen recombination and evolution reactions on the surface of the oxide in the presence of the above identified dominant defects in the space charge zone (for each 3d transition metal). We believe that the type of analysis presented here can open the route to physics-based alloy design to eliminate the hydrogen pickup challenge in the nuclear industry.
[1] B. Cox, M. J. Davies, A. D. Dent, “The oxidation and corrosion of zirconium and its alloys. Part X. Hydrogen absorption during oxidation in steam and aqueous solutions.”, AERE-M621, HARWELL, 1960.

The hydrogen pick up fraction (HPUF) during the aqueous corrosion of zirconium alloy fuel cladding in nuclear reactors is often described as a major factor limiting the burn-up fraction that can be allowed in the uranium fuel. Although this phenomenon has been studied for many decades there is still no agreement on the key mechanisms of hydrogen transport through the oxide scale to the underlying metal or the microstructural features in the alloy that many control HPUF. In part this is because of the technical difficulties of measuring local hydrogen concentrations in complex microstructures. This presentation will describe recent work in Oxford using the current generation of high resolution analytical techniques to study the transport paths for hydrogen during corrosion, including the use of deuterium spiking and SIMS imaging to provide direct evidence for the final destination of hydrogen isotopes during selected stages of the corrosion cycle.

Perspective on the Opportunities for Advancing Fuel Performance Modeling
Joseph Y R Rashid
Most of the important challenges facing the nuclear power industry today can be traced to fuel performance issues, primarily driven by materials behavior problems. During the last fifty years of light water reactor (LWR) operations fuel rods have undergone numerous design changes, both in material composition and geometric makeup, with the aim of eliminating or reducing fuel cladding failures. The UO2 part of the fuel rod, with far fewer options in material selection, has seen relatively minor, but no less important, fewer changes. The most remarkable aspect of these design changes is that, while they involve major redesign of the fuel rods and assemblies, they were implemented with relatively minor disruption of reactor operation. The process by which these design changes were introduced involved three major steps: new materials development; design, construction and irradiation of lead test assemblies (LTAs); and performance modeling and analysis in support of fuel licensing. However, despite the great care used in the development of those improvements, fuel rod failures continued to occur with the passage of time and the accumulation of burnup.
The protagonists for this inability to pre-predict fuel failures are the inherent deficiencies in the material behavior models and fuel performance codes that are designed to simulate engineering-scale type phenomena, whereas LWR fuel is a multi-component system that is subjected to complex multi-physics phenomena that occur over time scales ranging from less than a microsecond to years, and act over distances ranging from inter-atomic spacing to meters. These conditions impose challenging and unique fuel performance modeling and simulation requirements in order to accurately determine the state of the fuel during its lifetime in the reactor. The opportunities for advancing fuel performance modeling capabilities beyond the current engineering-scale of 1D and 2D fuel performance codes are discussed, and the challenges of employing higher fidelity performance modeling techniques are presented.
Fuel behavior issues that currently face fuel performance modelers span the three phases of the fuel cycle, namely, normal steady-state operations, operational and accident transients, and back-end used-fuel storage and transportation. The nuclear, physico-chemical and thermo-mechanical processes that are active in each of these phases are interdependent and require a multi-phenomenological modeling approach to capture their collective effect on fuel behavior during the whole fuel cycle. The treatments of each phase of the fuel cycle in isolation of one another, which has characterized current practice, is no longer viable under the current regulatory climate, which presents new and greater opportunities for multi-physics/multi-regime fuel behavior modeling.

The Peregrine fuel performance code is under development by the Consortium for Advanced Simulation of LWRs (CASL) program to provide a 3-D fuel performance modeling capability for predicting the impact of plant operation and fuel rod design on performance, including Pellet-Cladding Interaction (PCI) failures in current PWRs. The multi-physics and multi-dimensional nature of nuclear fuel performance, and the PCI failure mechanism, makes it a challenging choice as a focus for advanced modeling and simulation. PCI is controlled by the complex interplay of thermal, mechanical, and chemical behavior of a fuel rod during plant operation; thus modeling PCI requires an integral fuel performance code to simulate the intricacies of fuel behavior. This paper presents results documenting the initial verification and validation of a 2-dimensional, axi-symmetric version of Peregrine through benchmarking comparisons to Falcon model predictions and Halden Instrumented Fuel Assembly (IFA) experiments of both thermal and mechanical behavior. Initial benchmark comparisons indicate that Peregrine predictions agree quite well with 2-D Falcon predictions and Halden experimental data on fuel centerline temperature but that further developments are necessary for some models, including fission gas release and gaseous swelling. The mechanical behavior benchmarking study has compared predictions of clad deformation to dilatational measurements in IFA-585.4 and cladding elongation data from IFA-562.1, and the results show quite promising agreement. Following an overview of the verification and benchmarking activities, the paper will discuss Peregrine predictions to evaluate the effects of PCI by means of comparing to experiments performed in the third RISOslash; Fission Gas Release project and the Super Ramp project.

Advanced burner reactors are designed to reduce the amount of long-lived radioactive isotopes in spent nuclear fuel. The input feedstock for advanced fuel forms derives from either recycled light water reactor fuel, or recycled fast burner reactor fuel. In order to achieve higher performance and increase operational safety, these advanced reactors require novel fuel concepts, made from new materials. One promising pathway to improve fuel performance is the creation of metal or ceramic scaffolds, into which fuel may be placed with greater precision than in existing fuel types. In this presentation, the design and manufacture of novel structures by “freeze casting” will be described. Freeze casting (or “ice templating”) is a directional solidification process ideal for the production of both metal and ceramic fuel scaffolds. This process inherently allows for the manufacture of a range of custom-tailored fuel pellet designs. The mechanical, thermal, and neutronic properties of both metal and ceramic scaffolds will be compared and contrasted, in order to shed light on the performance and lifetime behavior of these novel fuel designs. Additionally, the results of Monte Carlo N-Particle models of a sodium fast reactor and a pressurized water reactor to which freeze-cast scaffold fuel pins were added will be summarized. The inert scaffold was found to decreases k-effective, but offered advantages, because it divides the fuel pin into smaller segments. The potential of the scaffold lies in the ability to design fuel pore by pore. This level of exactness could be used to make reactors run more efficiently or safely by reducing maximum fuel temperatures. Individual pins could be constructed specifically for actinide transmutation or medical isotope production.

Under neutron irradiation in nuclear power plants, uranium dioxide (UO₂), the most used nuclear fuel, changes gradually its chemical composition because of the incorporation of new chemical elements which are created by fissions and named Fission Products (FP). As a consequence, the fluorine-type crystalline structure and its lattice parameters are also modified.
In order to better understand this crystallographic behavior, neodymium-doped UO₂ ceramics are prepared with the aim to simulate the solid matrix of irradiated fuels, since Nd is one of the most abundant FP. In a previous work, high temperature X-ray diffraction was realized on a sample containing about 9 at% in Nd and annealed under reducing conditions. The diffractrograms evidenced, for the first time, the existence of a miscibility gap in the system.
Since there is a lack of experimental information about it, we have employed a theoretical model in order to obtain a first complete description of this miscibility gap. In this paper, a thermodynamic modeling of the ternary system U-Nd-O is presented, based in the Calculation of Phase Diagrams (CALPHAD) method.
The results of this modeling confirm the presence of a region presenting two FCC phases (instead of a single solid solution, which is expected from literature). At room temperature, the gap appears from a Nd content as small as 0.02 at% and a ratio O/M slightly lower than two. At higher temperatures the FCC solid solution (U,Nd)O₂ covers a larger domain of Nd content and the miscibility gap area decreases; and finally, at 1000 K, the biphasic domain completely disappears.
In addition to this theoretical approach, new experiments have been realized on (U,Nd)O₂ samples. They consist of XRD characterizations of samples containing different Nd contents, some of them annealed under different atmospheric conditions in order to evaluate the influence of O/M ratio. They will be also briefly presented and discussed in this paper.

Much effort is still being put on the improvement of first-principles methods to study radiation effects in actinide-based nuclear materials, and especially uranium dioxide, since obtaining precise data at the atomic scale is foremost for the development of models at higher scales [1]. Even if progress has recently been made in the description of the strong 5f electron correlations and of their effect on the energetic properties of actinide compounds thanks to the DFT+U method [2], their treatment is still a challenge.
First, it is particularly important to get precise energies for charged defects as a function of the stoichiometry of the material. We will present a scheme for the calculation of the formation energies of charged defects in UO2 and the results obtained in the DFT+U approximation for interstitial and vacancy defects in UO2.
Second, the combination of DFT with the dynamical mean field theory (DFT+DMFT) [3], which has been recently implemented in the Abinit code [4,5], can further improve the modeling of strongly correlated materials such as UO2. This method allows one to describe the dynamical correlations by taking into account the possibility for localized electrons to change configuration among correlated orbitals. The DFT+DMFT also circumvents the issue of local energy minima and facilitates greatly the study of paramagnetic systems. We will present the magnetic, electronic and mechanical properties obtained for UO2 using the DFT+DMFT method in the Hubbard I approximation and compare them with the results obtained in standard DFT, DFT+U and experimental results.
[1] F-BRIDGE deliverable D226, www.f-bridge.eu
[2] B. Dorado, B. Amadon, M. Freyss, M. Bertolus, Phys. Rev. B 79, 235125 (2009).
[3] G. Kotliar, Rev. Mod. Phys. 78, 3(2006).
[4] B. Amadon, F. Lechermann, A. Georges, F. Jollet, T.O. Wehling, A.I. Lichtenstein, Phys. Rev. B 77, 205112(2008).
[5] B. Amadon, J. Phys.: Condens. Matter 24, 075604 (2012).

A defect disorder model of UO2 founded on density functional theory results of defect energetics has been extended to investigate local off stoichiometry near UO2 surfaces. While bulk UO2 crystals contain defects and electronic charge carrier densities that solely depend on the oxygen partial pressure and temperature, surfaces were found to significantly modify the defect equilibrium states. Analysis of local defect densities near flat surfaces in UO2 showed that significant defect segregation occurs and that, under fixed thermochemical environment, local stoichiometry can change from hyper to hypo as a function of distance from the surface. A generalization of the theory to void surfaces has led to the discovery that voids in UO2 must contain oxygen gas. This important finding implies that, as a major component of UO2, oxygen may have a bigger role to play in the irradiation response of the material than previously believed. It was also found that voids are surrounded by significant defect segregation regions at void sizes in the range few tens to few hundred nanometers. This discovery also implies that the average O:U ratio of irradiated UO2 crystals containing ensembles of voids will be different from the unirradiated material under the same thermochemical conditions. The discovered electrochemical aspect of voids in UO2 gave us new insight into how to construct microstructure evolution models. A number of chemical characterization experiments are now being designed to validate this prediction. This research was supported as a part of the Energy Frontier Research Center for Materials Science of Nuclear Fuel funded by the U.S. Department of Energy, Office of Basic Energy Sciences under award number FWP 1356, through subcontract number 00122223 at Purdue University.

In the prospect of future nuclear reactors, U-Pu mixed oxides incorporating high amounts of plutonium are considered. During its lifetime within the nuclear cycle, the fuel is subjected to drastic changes in temperature associated with various cooling-heating rates. The U-Pu-O ternary diagram is still not precisely delimited, especially in the UO₂-PuO₂-Pu₂O₃ domain. More precisely, for Pu content higher than 20%, the literature reports the occurrence of a phase separation, depending on the temperature and on the oxygen stoichiometry [1-9]. Since these fuels will probably be designed with an O/M ratio < 2.0 corresponding to the hypostoichiometric domain of the ternary system, it is then necessary to study these compounds under drastic temperature variations corresponding to eventual in-pile conditions. In this study, using room-temperature X-ray diffraction after reducing sintering, we have evidenced a phase separation occurring in U_0.55Pu_0.45O_2-x compounds cooled at different rates (from ~0.08 to ~300 K.s^-1). From the results, it was concluded that the phase separation can&’t be avoided within the considered stoichiometry domain and, surprisingly, the lattice parameters of the obtained phases were identical regardless of the cooling rate, only their proportions were different. Using a novel in situ fast X-ray diffraction device, we have revealed that the phase separation temperature of a U_0.55Pu_0.45O_2-x compound is the same regardless of the cooling and heating rates (at 2 K.s^-1) and is also identical to the values available in the literature [3,8], i.e., 770 ± 20 K for much lower cooling rates (e.g. 0.05 K.s^-1). Furthermore, the effect of the cooling rate on the mixed oxide&’s microstructure has been studied focusing on sample microstructure characterization by optical microscopy. It has revealed that the cooling rate strongly impacts the microstructure of the fuel pellet by inducing severe macroscopic cracks. In addition to their obvious fundamental interest concerning the U-Pu-O ternary system, we believe our results are of utmost importance in the prospect of using uranium-plutonium mixed oxides with high plutonium content as nuclear fuels for future reactors. Considering the associated safety issues, they dictate a cautious attitude when defining the elaboration conditions of such materials.
[1] L.E. Russell et al., J. Nucl. Mater. 5, 1962, p. 216-227
[2] N.H. Brett, L.E. Russell, Trans. Brit. Ceram. Soc, 62, 1962, p. 97-118
[3] T.L. Markin, R.S. Street, J. Inorg. Nucl. Chem. 29, 1967, p. 2265-2280
[4] T.L. Markin, E.J. McIver, Plutonium 1965, Chapman and Hall, London, 1967, p.845-857
[5] C. Sari et al., Thermodynamics of Nuclear Materials, 1968, p. 587-611
[6] C. Sari, U. Benedict, H. Blank, J. Nucl. Mater. 35, 1970, p. 267-277
[7] G. Dean et al., Plutonium and Other Actinides, Plutonium 1970, 1970, p. 753-761
[8] T. Truphémus et al., Proc. Chem. 7, 2012, p. 521-527
[9] T. Truphémus et al., J. Nucl. Mater. 432, 2013, p. 378-387

The spent nuclear fuel is characterised by the presence of a large amount of impurities due to the fission of uranium nuclei and the associated formation of fission products (Iodine, Krypton and Xenonhellip;). According to the ab-initio calculations [1,2], sites of preferential incorporation of iodine and krypton are complex defects with uranium and oxygen vacancy defects such as Schottky defects. Both gases are considered weakly soluble in the material [3], which will therefore accelerate their diffusion, including their precipitation in the form of bubbles. Most studies published in literature are unanimous on the role of vacancy defects in the behavior of each gas from its insertion to its release outside the material.
The study of these vacancy defects generated by krypton and iodine in polycrystalline uranium dioxide and their annealing stages is the main objective of this work. To characterize these defects we performed ion implantations in the near surface at different fluences (from 1013 to 3x1016 at.cm-2). The samples were characterized by using a slow positron accelerator coupled to a Doppler broadening spectrometer (DB-SPB). The samples were therefore annealed under a controlled atmosphere and characterized again to study the evolution of defects as function of temperature.
These experiments have shown that 4 MeV Krypton and 8 MeV iodine implantations lead to the creation of a predominant detected vacancy defect which corresponds to the displacement of U atoms and could be the Schottky defects VU-2VO, as first DFT results seem to show.
In the case of iodine irradiations the detected vacancy defects concentration appears homogeneous as a function of depth, and increases as a function of fluence. The nature of the induced vacancy defects does not change with the fluence. Two stages of defects annealing have been observed before 700°C, and a third one appeared at about 1100°C.
Krypton induced defects are different and evolves after annealing at temperature higher than 1000 °C indicating precipitation of vacancy defects. The role of Kr in the detection and evolution of vacancy defects will be discussed.
References
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[3] S. Kashibe, K. Une, K. Nogita, J. Nucl. Mater, 206 (1993) 22.

CrUO₄ has been produced, studied using diffraction techniques and modelled using empirical pair potentials. Cr₂O₃ was found to preferentially form CrUO₄ over going into solution in hyper-stoichiometric UO₂. Further, it was found that the formation of CrUO₄ can proceed by removing excess oxygen from the UO₂ lattice. Partition energies, the energy to remove fission products from hyper-stoichiometric UO_2+x and incorporate them into CrUO₄, have been calculated. Cation partition into CrUO₄ was only found to be preferable for smaller cations (e.g. Zr4+, Mo4+ and Fe3+) while all divalent cations are predicted to remain in the hyper-stoichiometric UO_2+x phase. X-ray diffraction confirmed the structure of CrUO₄ and predicts an (Al,Cr)UO₄ compound for the first time. The reduction of UO_2+x due to the formation of CrUO₄ will have important effects on the solution limits of other fission products as many species are less soluble in UO₂ in comparison to UO_2+x.

The mechanical behavior of nuclear fuel during irradiation depends on a great number of individual phenomena, only a few of which are adequately understood. We use atomic-level simulation methods to determine the interaction of metallic fission product, Ru4+ with the core of ao/2<110>{110} and ao/2<110>{001} edge dislocations in UO2. Comparisons are made with both continuum-elastic results and with the results of atomistic simulations on strained single crystals. Analysis shows that the trends in segregation energy can be understood in terms of bulk behavior and continuum elasticity. Segregation behavior is found to be a strong function of the elastic strain field around the dislocation core and is affected by the orientation of the dislocation and electrostatic interactions at the atomic defect site. This work provides insight into how atomic structure of edge dislocations influences segregation of species in nuclear fuels. This work is supported by the DOE-NE Nuclear Energy University Program and by the DOE-NE Advanced Modeling and Simulation (NEAMS) Program, and FUELS: Integrated Performance and Safety Code (IPSC) Project.

Ceramics have multiple roles within the nuclear technology, ranging form novel fuel types/additives, through to magnetic containment in fusion cores. One major drawback with ceramics is their response to radiation damage. In many cases the effects of damage are minor, whereas in others it can be catastrophic, with a loss of required properties. Understanding the effects of radiation damage is a complex process with multiple competing processes during recovery. These processes are linked to the composition, adopted structure, order/disorder, radiation fluence and temperature.
For example in some systems, recovery from damage is linear with compositional change, while in others the reverse is found. How ceramics behave under such non-equilibrium conditions is difficult to predict, but the degree of predictability is improving. Using model systems such as perovskites or pyrochlores, with changeable composition and structures, it is possible to derive new insights into recovery processes that can be applied to other systems.
Model systems will be presented, based on perovskites with compositional/structural change, and pyrochlores with novel compositions/order. Descriptions of recovery mechanism will also be presented, and models for recovery developed.

During in-reactor irradiation actinide fission produces large quantities of volatile fission products, which have a significant influence on the structural and mechanical properties of nuclear fuels and claddings. It is therefore essential to get further insight into the behaviour of these elements in materials to improve the understanding of the behaviour of fuel systems and improve their performance. The incorporation sites and activation energies determine the mobility in the material, as well as the influence of temperature and defects on this mobility. It is then of major importance to evaluate these parameters.
Silicon carbide (SiC) is an envisaged cladding for future nuclear reactor and as such its behaviour regarding fission products, especially volatile ones such as xenon, krypton, iodine or caesium, must be evaluated.
We will present the investigation of the atomic transport properties of volatile fission products in cubic silicon carbide (SiC) using density functional theory. In particular, incorporation energies of these fission products in SiC vacancies with various charge states have been determined and the activation energies to their migration have been calculated using a five-frequency model. The results are compared with experimental results.

Silicon carbide (SiC) is considered a promising candidate for structural and cladding applications in advanced nuclear reactors, and thus radiation damage processes in SiC have been extensively investigated. A recent experimental study showed that a nano-engineered SiC, containing nano-grains with a high-density of stacking faults (SFs), exhibits a better radiation resistance than single crystals. Since SFs are inevitably incorporated in a practical material, understanding the effects of SFs is important for predicting realistic performance and for development of a more radiation-resistant material. The present study aims to assess the influence of SFs on defect recovery processes.
To achieve this aim, quantum mechanical calculation based on density functional theory (DFT) was performed using the VASP code. Two exchange-correlation functionals were employed. One is the Perdew-Burke-Ernzerhof (PBE) functional used for relaxation of defect configurations and the other is the Heyd-Scuseria-Ernzerhof (HSE) type screened hybrid functional for accurate determination of energies and electronic structures. The screening parameter of the HSE functional was refined so that the band gaps of SiC polytypes are correctly reproduced. As a model 3C-SiC crystal containing a SF, a system of -ABCAB- stacking sequence was prepared. The contained SF is an intrinsic type. In SiC, interstitials are considered to be more mobile than vacancies for both silicon and carbon. Thus, the calculations were conducted mainly for interstitials. Defects in several hexagonal polytypes, namely 2H, 4H and 6H, were also investigated for comparison.
Initially, the stable configurations of interstitial defects were determined. In single crystal 3C-SiC, split interstitials are energetically favorable for both silicon and carbon. A silicon interstitial can also reside at a vacant site to form a carbon-coordinated interstitial. In addition to these fundamental defects in 3C-SiC, the introduction of SF invoked a few extra defect configurations with comparable stabilities. The stabilities of defects located at different stacking layers in the SF-containing system differed by at most 1 eV. These stabilities could be roughly arranged as a function of the distance from the SF. Because these defect configurations correspond to the initial and/or final states of defect migration pathways, the migration energy was also largely altered. Additional insights on the interaction between an intrinsic defect and SFs will be provided, and the influence of SFs on the radiation resistance will be discussed.

The interaction of ions with solids results in energy loss to both atomic nuclei and electrons. At intermediate ion energies, nuclear and electronic energy losses are of similar magnitude and can lead to synergistic or competitive processes that affect the evolution of irradiation damage. This energy regime includes energies of primary knock-on atoms created by fission and fusion neutrons, as well as the energies of ions used to investigate neutron damage in materials. A previous study on SiC indicated that irradiation damage production was dependent on the ratio of electronic to nuclear energy loss [1] for intermediate energy ions (0.8 to 2 MeV). More recently, an experimental and computational study on SiC [2] has clearly demonstrated that very high electronic energy loss (33 keV/nm) can induce defect recovery and recrystallization of pre-existing irradiation damage. To better understand and quantify the effect of electronic energy loss, we have performed experimental studies on ionization-induced recovery on pre-damaged disordered states in SiC at 300 K over a range of electronic energy loss from 1.9 to 8.0 keV/nm. The pre-damaged states were prepared by irradiation with 0.9 MeV Si ions to fractional disorder levels of 0.3 or 0.7. The effects of ionization induced recovery were studied using ions and energies (MeV to tens of MeV) with a high ratio of electronic to nuclear energy loss in order to minimize the effect of additional damage production. The experimental results clearly show that ionization-induced damage recovery occurs as a distinct separate effect at electronic energy loss values of 1.9 keV/nm and higher under these conditions. These results also provide confirmation of the role of electronic energy loss in the previous study [1], where electronic energy loss ranged from 1.4 to 5.0 keV/nm, but at much lower ratios of electronic to nuclear energy loss. The impact of these results on the interpretation of radiation damage processes in SiC and other materials will be discussed.

[1] W. J. Weber, Y. Zhang, and L. M. Wang, Nucl. Instr. and Meth. in Phys. Res, B 277 (2012) 1.
[2] A. Debelle, M. Backman, L. Thomé, W. J. Weber, M. Toulemonde, S. Mylonas, A. Boulle, O. H. Pakarinen, N. Juslin, F. Djurabekova, K. Nordlund, F. Garrido and D. Chaussende, Phys. Rev. B 86 (2012) 100102(R).
This work was supported by the U.S. Department of Energy, Office of Basic Energy Sciences, Materials Sciences and Engineering Division.

Silicon carbide (SiC) has been proposed as a material for fission-product barrier coating in novel nuclear fuel particles, as a cladding material for fission reactors, as well as for structural components in fusion reactors due to its high temperature stability, small neutron capture cross-section and chemical inertness. Therefore the material will be exposed to radiation of a wide energy scale both from neutrons and from heavy fission fragments, and understanding its full irradiation response is important for advanced nuclear energy systems.
In the low energy regime, the displacements from nuclear energy loss account for the majority of damage in a crystalline material; however, in the intermediate to high-energy irradiation regime, the electronic stopping dominates and has been shown to induce defect recovery and recrystallization in SiC [1], which is well described by an inelastic thermal spike (i-TS) formalism.
Molecular Dynamics simulations, which include the energy deposition from electronic stopping at different irradiation temperatures following the i-TS calculation input, and/or ballistic collisions, complement our recent ion-beam experiments and clearly show that the irradiation-induced defect recovery process in SiC is active to low values of electronic stopping, in a regime where electronic stopping is often considered negligible. The competitive processes of damage production and defect recovery are relevant for understanding radiation damage production for many materials in nuclear energy applications and for investigating radiation damage in potential nuclear materials using ion irradiation methods.
[1] A. Debelle, M. Backman, L. Thomé, W. J. Weber, M. Toulemonde, S. Mylonas, A. Boulle, O. H. Pakarinen, N. Juslin, F. Djurabekova, K. Nordlund, F. Garrido and D. Chaussende, Phys. Rev. B 86 (2012) 100102(R).
This work was supported by the U.S. Department of Energy, Office of Basic Energy Sciences, Materials Sciences and Engineering Division.

Metal transition carbides and nitrides are considered as excellent candidate materials for nuclear fuel applications in Generation IV fission reactors and for coating applications in fusion machines due to their high thermo-mechanical and radiation tolerance properties. The consequences of helium accumulation in this type of materials need to be clearly understood to obtain better predictions about their ageing processes.
To study mobility of helium in polycrystalline material under thermal annealing includes various challenges. a) to accurately determine the He depth profile and further to calculate various activation energies of He migration using mathematical models. b) the trapping of He atoms into the point defects created during He implantation resulting in the formation of He-V (vacancy clusters) or bubbles. c) to determine the role of grain boundaries which acts as effective short circuits for He movement and release. d) to know the role of He implantation concentration and presence of native vacancies into the material.
To solve above challenges our approach includes 7 steps:
- 3 MeV 3He+ ion implantation with fluence of 5x1016 at/cm2 (~ 2 at. % at Bragg peak) into polycrystalline samples of TiC, TiN and ZrC.
- Thermal annealing at various temperatures between 1000 °C and 1600 °C for 2 hours each.
- He depth profiling measurement of as-implanted and annealed samples using the 3He(d,p0)4He nuclear reaction and the use of mathematical models like AGEING and SIMNRA to calculate migration parameters.
- 3-D elemental distribution image of He atom from micro-NRA to study the role of grain boundaries.
- TEM to observe nucleation and growth mechanisms of He bubbles in as-implanted and annealed samples.
- Raman micro-spectrometry to study the defect created during He implantation and subsequent changes in defects after thermal annealing.
- Thermal desorption spectroscopy (TDS) to know different types of He-Vacancy cluster into the material.
Some of the results includes the low value of activation energy for He release in case of ZrC with no thermal diffusion. Whereas, a higher activation energy for He release along with thermal diffusion is observed for TiC and TiN. Above 1500 °C, exclusive blisters are observed on the surface of ZrC suggesting the presence of over pressurized bubbles, whereas no such surface changes are observed on TiN and TiC.
The heterogenous 3-D elemental distribution image of He atom obtained through µ-beam NRA confirms the role of grain boundaries. Raman spectrometry showed reduction of defects on annealing. To study He migration parameters as a function of He concentration, the He implantation fluence has been varied from 5x1015 to 5x1016 at/cm2. TEM observations are in progress to observe He bubbles growth on thermally annealed samples. TDS experiments are planned in near future to know different type of He-V clusters.

The typical use conditions of a rubber O-ring plays at least as important of a role in the effectiveness of the seal it forms as the physical properties of the O-ring itself. Under normal use conditions, O-rings are subject to wear, the formation of nicks and cuts, environmental contamination by hair or dirt, and mishandling by the users. We systematically examine how each of these factors impacts the leak-tightness of a nuclear material storage container, and the likelihood that any one of these factors will allow the inadvertent release of radioactive material.

The development of nanostructured ferritic alloys (NFA) is a novel approach to ferrous metallurgy for advanced reactor applications in which the traditional carbides in steels are replaced by oxygen-based nanoclusters and precipitates. In order for the oxygen and yttrium to be incorporated into the solid solution, the alloys are fabricated by mechanical alloying yttria powder with a master iron-chromium-tungsten-titanium alloy rather than the more customary casting route. The resulting flakes are consolidated into bulk form by high temperature extrusion. This processing route not only introduces the yttrium and oxygen into the solid solution but also introduces high numbers of vacancies. The fundamental principle behind this approach is based on the strong binding energies between O, Ti, Y atoms and vacancies, as estimated by first principles calculations, and results in limited solute diffusion due to the trapping of the vacancies. The limited diffusion also results in a highly stable microstructure after the initial formation stage, with little coarsening of the high number density of Ti-Y-O-enriched nanoclusters, precipitates or grains when exposed to extreme conditions. In addition, Cr and W segregation to and preferential precipitation on the grain boundaries of this ultrafine grained alloy effectively pin the grain boundaries. When these alloys are exposed to low temperature, high dose ion irradiations, the atoms in the nanoclusters are distributed into a random solid solution by ballistic collusions during the displacement cascades. However, when the irradiations are performed at high temperatures, the nanoclusters are evident. This difference in behavior arises from the lack of diffusion in the low temperature case which prevents the re-nucleation of the nanoclusters, whereas at the higher temperatures, the combination of the additional vacancies produced during the displacement cascade and high temperatures allows the excess solute in the cascade affected zone to form nanoclusters. This stable microstructure results in an NFA with excellent creep properties and, most importantly, a viable method to create radiation tolerance in a structural ferrous alloy to high doses of irradiation.
Research sponsored by the Materials Sciences and Engineering Division, Office of Basic Energy Science, U.S. Department of Energy. Research at the Oak Ridge National Laboratory ShaRE User Facility was sponsored by the Scientific User Facilities Division, Office of Basic Energy Sciences, U.S. Department of Energy.

To meet growing energy needs of the United States and the planet, more will be demanded of all energy technologies, including nuclear energy. Almost all of the advanced nuclear reactor concepts require operations under severe conditions of temperature, stress and radiation. To materials scientists, the primary challenge in realizing any of the advanced fission and future fusion energy systems is to design new high-performance structural materials for components -- such as the cladding and structural materials for fission reactors and first wall and blanket structural materials for fusion systems -- that can withstand such extreme operating conditions without compromising the structural integrity of the reactor over a long period of time. Nanostructured ferritic alloys (NFAs) are considered excellent candidate materials for such structural applications as they exhibit exceptionally high creep strength due to the presence of highly stable nanometer sized Y-Ti-O oxide precipitates within the primarily iron matrix. NFAs have also shown particular promise for their high radiation tolerance and it is believed that metal/oxide interface is responsible for such excellent radiation resistance in NFAs. Thus insight about the atomic structure of the metal/oxide interface is critical in understanding the origin of the enhanced properties of this material and ultimately designing new radiation resistant alloys.
Y₂O₃ has also been shown to form nanoprecipitates in iron and is a simpler surrogate for the Y-Ti-O precipitates. In this work, we present the structure/property relationship of the interface between the iron matrix and Y₂O₃ using density functional theory. We observe that, depending on the external partial pressure of oxygen, a critical number of defects -- iron vacancies and/or interstitial oxygen -- are essential in stabilizing the atomic structure of the metal/oxide interface. Importantly, the accommodation of these defects is very sen