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
[1] T. Petit, M. Freyss, P. Garcia, P. Martin, et al, J. Nucl. Mater, 320 (2003) 133.
[2] R.W. Grimes, R.G.J. Ball, C.R.A. Catlow, J. Phy & Chem. Sol., 53 (1992) 475.
[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 sensitive to the misfit dislocation structure at the interface, with the vast majority of these defects being accommodated into the misfit dislocations. Finally, we will show that it is possible to predict the segregation behavior of alloying elements to any given metal-oxide interface based on simple thermodynamic concepts like Ellingham diagram and Hume-Rothery rules. Predicted segregation behavior is later verified by calculated segregation energy obtained from electronic structure calculations. The insight gained in this research provides the fundamental understanding needed to develop new NFAs tailored to meet challenges in fission and fusion applications, including the current fleet of light water reactors.

One of problems for development of the nuclear fusion demonstration reactor (DEMO) is the high heat load on heat-resistant equipments; e.g., the divertor and the blanket. The materials for such equipments are required a high thermal conductivity, while, the thermal conductivity of reduced-activation ferritic/martensitic steels, which are candidate materials for the first wall of a fusion reactor due to the excellent resistance to radiation damage, does not satisfy the requirements for the heat load expected in DEMO. In this study, the development of iron-based composite materials with a high thermal conductivity materials were challenged in order to add a role of heat sink to the complicated joint structure in the heat-resistant equipments.
Iron/carbon nanotube (CNT) and iron/cupper composite materials were developed by vacuum hot pressing (VHP) and spark plasma sintering (SPS). Note that CNT and cupper has a high thermal conductivity, respectively. VHP and SPS were conducted in some conditions. The range of pressing temperature were from 873 to 1473 K. Pressing pressure and holding time were 25 MPa and 30 min, respectively. The 10mm diameter disk were cut from the composite materials. The thermal conductivity was measured by the laser flash method and the microstructure in junction interface were investigated by scanning electron microscopy (SEM).
The thermal conductivity in the iron/CNT composite materials exhibited little improvement in the experimental condition. Instead, the formation of carbides was observed at the interface of joints. One reason of a poor thermal conductivity would be non-aligned dispersion of CNT. Improvement of its thermal conductivity could be succeeded with a network and well-oriented and -aligned CNT in between joints. The thermal conductivity in the iron/cupper composite materials was slightly improved in the experimental condition, but lower than a theoretical value. The differences between experimental and theoretical value would be a poor interface adherence.

Nanostructured oxide dispersion-strengthened (ODS) ferritic alloys show remarkable high-temperature strength and irradiation resistance. This behavior is attributed to the nano-size oxide precipitates within the microstructure. Mechanical alloying, ball milling in particular, is typically employed in order to produce nanostructured ODS alloys using nano-particle of yttrium oxide. It is believed that oxide particles are dissolved in the microstructure or amorphized during the ball milling, and re-precipitate at high temperatures during annealing or hot consolidation.
This work was motivated by the need of more evidence to clarify the real sequence of events which results in the formation of nano-scale precipitate. In this study, Fe-14Cr-0.4Ti alloy was ball-milled with different amounts of Y2O3 particles up to 10 wt. %, and then annealed at temperatures up to 1100 °C. Micron-size Y2O3 particles were substituted with the nano-size counterpart to study the mechanism of precipitate formation. XRD results revealed that as-milled solid solution does not contain primary Y2O3 particle, and after annealing at high temperatures, precipitation initiates. TEM study reveals that the as-milled microstructure has grain size of less than 20 nm. HRTEM shows that the oxide particles are reduced to nano-size during ball milling. Defect analysis of precipitates on the annealed samples via TEM and HRTEM study disclose that the use of micron-size primary oxide particle can provide the nano-size precipitation, stable in nano-scale up to 1100 °C, and distributed uniformly throughout the microstructure.

<111> crowdions are the most stable single self-interstitial atomic defect in many of the bcc transition metals, and are created in quantity when these metals are irradiated. They aggregate into the prismatic dislocation loops that form much of the microstructure characteristic of radiation damage [1].

Crowdions are quasi-one-dimensional, and generally have a very low barrier (O(meV)) to migration. Essentially, they are strongly delocalized along the <111> direction, and the motion of the defect centre-of-mass proceeds via the almost infinitesimal translation of many constituent atoms. The heights of the migration barriers for the bcc transition metals were determined in [2] via a hybrid analytical-DFT model, and were used to roughly estimate the temperatures (Tmig ~ Emig/kB) at which the defects begin to move, correlating fairly well with the shape of the resistivity recovery curves in [3].
In this work we extend the treatment of [2], and investigate the migration of crowdions at low temperatures. Given the exceedingly low values the defect effective mass can take, quantum effects can be important.
The crowdion is modelled as a quasiparticle of effective mass m* moving in a sinusoidal Peierls potential. The sine-Gordon equation admits kink solutions for the migration trajectory from one Peierls valley to the next (not to be confused with the very similar solution for the atomic displacement field). We identify these kinks with sine-Gordon instantons [4], and evaluate the Feynman path integral for the tunneling rate in the semiclassical limit. This is compared with the classical Kramers escape rate for thermal migration, and the quantum/classical transition temperature is identified.
[1] G. S. Was, “Fundamentals of Radiation Materials Science: Metals and Alloys” (Springer, 2007)
[2] Fitzgerald, S.P and Nguyen-Manh, D. (2008) "Peierls potential for crowdions in the bcc transition metals", Phys. Rev. Lett. 101,115504
[3] P. Ehrhart, P. Jung, H. Shultz and H. Ullmaier, in “Atomic Defects in Metals”, eds. H. Ullmaier et al Landolt-Bornstein New Series, Group III, 25, (1991).
[4] R. Rajaraman, “Solitons and Instantons” (Amsterdam: North Holland, 1987).

Recent advances in combining aberration-corrected scanning transmission electron microscopy (AC-STEM) with high-collection-angle energy dispersive X-ray spectrometer (EDS) systems allow unprecedented data acquisition rates and compositional detail in local elemental analysis. These techniques were used to study a 14YWT nanostructured ferritic alloy (NFA) that was heavy-ion irradiated up to a maximum dose of 300 displacements per atom (dpa) at temperatures of -100, 650, and 750°C. AC-STEM applied high currents (6-7 nA) in a ~0.5 nm probe, combined with the large-area EDS detector, allowed the compositional distribution across the entire ~2000 nm depth of the heavy-ion-irradiation damaged zone and detailed scans with ~0.5 nm resolution allowed individual nanoclusters and grain boundary segregation features to be chemically mapped in times < 1 hour. The results indicate that high dose ion irradiation at -100°C homogenizes the as-fabricated NFA structure, but high-temperature irradiation (650-750°C) results in growth of grain boundary precipitates, a slight reduction in grain boundary solute levels, and partitioning of impurity aluminum to nanoclusters and Y-Ti-rich precipitates.
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. FEI Titan STEM usage courtesy NCSU Analytical Instrumentation Facility.

Highly alloyed systems based on Nb, Ti, V and Zr have been modeled using a range of techniques and scales including simple thermodynamic methods using empirical data and ab-initio methods. Selected samples of these alloys, relevant to the nuclear industry, have been produced by magnetron sputtering. XRD and electron microscopy characterization of the produced phases was carried out to guide the modeling. We begin our alloy design with the ternary system of Zr, V and Ti, each of which is relatively neutron-transparent. Although Zr, Ti and V are each BCC at elevated temperatures, their ternary system contains an intermetallic compound (V2Zr), two allotropic transformations α-Ti <->β-Ti, and α-Zr<->β-Zr and a miscibility gap (α-Ti / V). We examine strategies by which these transformations can be suppressed or controlled by addition of more elements leading to the formation of a solid solution. Solid solution strengthening, due to the small mismatches in atomic size, should provide enhanced yield strength and hardness. The alloys should, in addition, possess good resistance to radiation damage due to their inherent disorder. The crystal structure of these alloys - termed high entropy alloys - was predicted and further analysis including their mechanical and elastic properties reported. The experimentally produced alloys support the predictions made on these systems.

Self-interstitial-atom (SIA) defect cluster with C15 Lave phase structure in body-center-cubic (bcc) iron was recently proposed to be important for the micro-structural evolution of irradiated bcc iron. Although previous computational research indicates that this type of defect cluster, when considered alone, is stable, highly immobile and grows easily, there is yet no experimental observation of the C15 defect cluster in irradiated bcc iron. Atomic simulations under a more realistic defect environment surrounding the C15 structure could provide a more comprehensive understanding of this type of defect cluster. The effects of a self-interstitial, an a/2 <111>{112} edge dislocation, and shear strain field on the evolution of C15 defect cluster are studied by the atomistic activation-relaxation algorithm, Autonomous Basin Climbing method. An additional self-interstitial near the C15 cluster reduces the activation barrier of the unfaulting process from 1.40 eV to 0.43 eV, along with changing the unfaulting mechanism. A 3% pure shear strain can lead to only a 0.1 eV lower activation barrier, while the dislocation can absorb the cluster with a barrier of 1.1 eV. These results show that the defect environment around the C15 cluster can destabilize it and accelerate its unfaulting to mobile defect configurations, and reconcile the controversy between previous simulations and the lack of experimental evidence.

While driven by a combination of irradiation, stress and aggressive water environment, the mechanistic origin of irradiation-assisted stress corrosion cracking (IASCC) remains poorly understood. In this research, a two-fold approach is used to study the micromechanistic process that may trigger IASCC. The first characterizes dislocation channel interactions with grain boundaries in proton irradiated and deformed material in a simulated boiling water environment. The second uses in situ TEM deformation experiments to directly observe these interactions in ion irradiated material. Correlating these studies reveals how the interaction of dislocations with grain boundaries may lead to disruption of the protective surface oxide scale such that the steels are exposed to the water environment.

Until now, austenitic steels (316Ti, 15/15Ti, AIM1) were used to clad the nuclear fuel of fast reactors (1) but currently new oxide dispersion strengthened steels (ODS) are considered as promising cladding materials because of their superior temperature strength, their thermal properties and their resistance to higher irradiation doses.
A specific R&D program is carried out by CEA in order to obtain data on the ODS clad corrosion during the fuel dissolution step in concentrated nitric acid (2,3). Indeed, an increase of iron or chromium coming from the clad corrosion could impact the reprocessing plant equipment lifetime by oxidizing reactions and could potentially impact the waste loading of the nuclear glass. Consequently, the corrosion phenomenon has to be carefully studied and predicted to avoid these potential detrimental effects on the integrity of the reprocessing plant and it performances.
This study will present the ODS corrosion behavior in nitric acid solutions in presence of oxidizing actinide species as Pu(VI) and Np(VI). Corrosion tests were carried out by mass loss measurements (immersion tests) and electrochemical techniques (with linear sweep voltametry). The release of major clad constituents (Fe and Cr) is quantified by ICP analysis during the immersion. Actinide concentrations and HNO2 are also monitored by UV-Vis measurements. The corroded surface was also characterized by SEM.
Three ODS grades were selected depending of their chromium content: 9, 14 and 18%Cr. The dependency of the ODS corrosion rate with the chromium content in the steel or with the actinide concentration in solution is discussed to prevent corrosion problems during the fuel reprocessing process. Moreover the neptunium seems to have an important role on the corrosion mechanism. At the same time, corrosion resistance of a reference steel, 15-15Ti steel, will be compared with ODS in original and thermally aged conditions. The effect of the thermal aging will be investigated on the corrosion resistance point of view.
(1) P. Fauvet et. al; Corrosion mechanisms of austenitic stainless steels in nitric media used in reprocessing plants, Journal of Nuclear Materials, 375 (2008) 52-64
(2) B. Gwinner, et. al; Impact of the use of ferritic/martensitic ODS steels cladding on the fuel reprocessing PUREX process, Journal of Nuclear Materials, 428 (2012) 110-116
(3) Y. de Carlan, et al.; CEA developments of new ferritic ODS alloys for nuclear applications, Journal of Nuclear Materials, 386-388 (2009) 430-432

Hafnium hydrides (HfH_X) have been developed and studied as a control rod material for fast reactors. HfH_X have several advantages over B₄C which has been mainly used in fast reactors as a control rod material. The lifetime of B₄C control rods is restricted by a pellet-cladding mechanical interaction failure due to helium gas swelling. On the other hand, HfH_X produce no helium gas during neutron absorption. The nuclear reaction of Hf is expressed as ¹⁷⁷Hf (n,γ) ¹⁷⁸Hf (n,γ) ¹⁷⁹Hf (n,γ) ¹⁸⁰Hf. Since hydrogen atoms in HfH_X are able to moderate neutrons, fast neutrons become thermal neutrons around Hf atoms. Thus, HfH_X have high neutron-absorbing capability in fast reactors. On the other hand, it is important to understand the thermal properties, especially heat capacity of HfH_X when HfH_X are used as a control rod in the reactors. Therefore, in the present study, we evaluate and analyze the heat capacity of HfH_X and HfD_X.
Fine bulk samples of HfH_1.63 and HfD_1.60 were prepared and their heat capacities were investigated in the temperature range from liquid helium temperature to 673 K. The Debye temperature and γ coefficient of the electronic heat capacity were evaluated from the low-temperature heat capacity data. The heat capacities of HfH_1.63 and HfD_1.60 were analyzed in terms of various contributions, i.e., the vibrational term for acoustic mode, the dilatational term, the electronic term, and the vibrational term for the optical mode. It was revealed that, at low temperatures, the acoustic mode of lattice vibration mainly contributes to the heat capacity, while above room temperature, the hydrogen and deuterium vibration, that is the optical mode, remarkably increases the heat capacity.

Ab-initio atomic scale computer simulations have been used to predict the solution enthalpies of H in Zr and Zr-M intermetallic phases (M=Cr,Cu,Fe,Mo,Nb,Ni,Sn,V).
In Zr metal, H preferentially dissolves into the β phase (BCC) compared to α phase (HCP). However, addition of the element Nb is predicted to reduce this preference to the point that the accommodation energy of H in the two phases is comparable. Therefore binary Zr-Nb alloys are predicted to have no strong H sinks.
Of the intermetallic second phase particles (SPPs) studied, all the Nb or V containing Laves phases accommodate H more readily than α-Zr, whilst all the SPPs containing either Mo or Cr provide very unfavourable solution sites. Cu, Ni and Sn additions may form a number of intermetallic phases, but tend to stabilise as Zr-rich phases, all of which provide lower energy sites for H accommodation compared to α-Zr. Finally, Fe additions form Zr-rich SPPs with high affinity for H as well as Fe-rich (Laves) SPPs with low H affinity.
Notably, intermetallics that relate to the Zr2(Ni,Fe) SPPs found in Zircaloy-2 exhibit favourable (but small) solution enthalpies for H, whilst the intermetallic phases that relate to the Zr(Cr,Fe)2 SPPs, found predominantly in Zircaloy-4, do not offer favourable sites for interstitial H.

Nuclear fuels based on composite materials are alternatives for nuclear energy generation. The dispersion type nuclear materials consist of particles of uranium - based material (or other nuclear material) dispersed into a metallic, ceramic, or graphitic matrix. These composites are the center of the fuel element and are protected by claddings of metallic plates, thick enough to avoid the escape of fusion products. Uranium alloys have been considered promising nuclear dispersing materials. In the present study, we investigated the microstructures of U-Mo alloys processed by two different melting techniques, i.e. arc melting and induction melting. The investigated alloys have the following compositions: U-7%wMo, U-8,5%wMo and U-10%wMo. The as-cast alloys have their chemical composition characterized by atomic absorption spectroscopy to verify their chemical composition. The alloys&’ microstructures were characterized by optical microscopy, scanning electron microscopy, X-ray energy dispersive spectroscopy, and X-ray diffractometry. These analyses indicate that the alloys melted by arc-melting are mainly single phased with their microstructure consisting of γ-(U,Mo) solid solution. The alloys processed by induction melting have microstructures composed by a matrix of γ-(U,Mo) phase and α-(U,Mo) as a second phase. Therefore, to obtain a fully single phased γ-(U,Mo) alloy by induction melting, a high temperature annealing is necessary to dissolve the α-(U,Mo). Moreover, it was observed the presence of an oxide phase , indicating that oxygen contamination play an important role on the quality and homogeneity of the arc-melted alloys. Finally, the influence of Mo content and heat treatment procedures are also depicted in this study.

Zirconium-based alloys, known under the name of Zircaloy, are used as cladding materials in light water reactors because zirconium has a small neutron macroscopic absorption cross section. It consists of zirconium with small quantities of such metals as tin, iron, chromium, and nickel in order to improve the corrosion resistance and mechanical properties of the cladding. Among these additives, tin dissolves into the alpha phase of zirconium (αZr) substitutionally. Oxygen also dissolves into αZr interstitially. The solubility limits of tin and oxygen are 1-2 and 29 at%, respectively. This alpha phase of zirconium, containing Sn and O (αZr-Sn(O)), is possibly formed during loss-of-coolant accidents (LOCA). Under LOCA situations, Zircaloy cladding would be exposed to steam at high temperatures over 1000 degrees, and Zircaloy transfers from α to β phase. Steam reacts with Zircaloy to form a superficial layer of ZrO₂ and an intermediate layer of αZr-Sn(O). It is well known that the oxide film has the lower thermal conductivity than β-Zircaloy, and a number of studies have been conducted to reveal the effect of the formation of the oxide layer. Although the formation of αZr-Sn(O) may also affect the thermophysical properties of the cladding materials, little is known about αZr-Sn(O). The purpose of this study is to evaluate the thermophysical properties of αZr-Sn(O) solid solutions. Samples of Zr-Sn(O) (Zr_0.9-xSn_xO_0.1, x = 0, 0.01, 0.02 and 0.04) were prepared by solid-state reaction of Zr, ZrO₂ and SnO₂ at 1420 K. The X-ray diffraction patterns of the samples of x = 0, 0.01, and 0.02 were well consistent with that of αZr hexagonal crystal. However, the sample x = 0.04 had the impurities such as SnO₂, ZrO₂, and Zr₅Sn₃. The Scanning Electron Microscope/ Energy Dispersive X-ray Spectrometer analysis confirmed that Sn and O were uniformly-distributed in the samples of x = 0, 0.01 and 0.02. The thermal diffusivity of the samples of x = 0 and 0.01 are 0.06 and 0.03 cm²/sec, respectively. This result indicates that the dissolved Sn reduce the thermal diffusivity by enhancing phonon scattering.

Zirconium based cladding tube reacts with water vapor and fuel pellets during LWR severe accidents, and mainly forms (U,Zr)O₂ as debris. The fuel debris are considered to be treated by wet processing, however, it was reported that zirconium-rich (U,Zr)O₂ does not dissolve in nitric acid. In the present study, ball milling process is focused for improvement of the (U,Zr)O₂ solubility. It is known that the ball milling process promotes amorphization and chemical reaction at room temperature depending on the conditions. The powders of UO₂ / ZrO₂ mixture (UO₂ :ZrO₂ = 1:9) were milled for 20 hours, 40 hours and 60 hours respectively, and immersed in 6N nitric acid aqueous solution for 1 hour at 353 K. The degree of dissolution was estimated by measuring weight of the residue. The powders were observed by a scanning electron microscope before and after the immersion. The ratio of Zr and U of the residue were quantitatively analyzed by EDS. The solubility increased with the ball milling time. On the other hand, only UO₂ dissolved for not milled sample. SEM and XRD analysis confirmed that the particle size reduced to several nano-meters by ball milling. The atomic ratio of U and Zr of residue is 1:9, indicating that the samples dissolved uniformly for milled samples. The increased solubility is probably caused by accumulated strain generated by the ball milling, which destabilizes the crystal structure.

The unique abilities of atom probe tomography (APT) have been used to characterize the response of a 14YWT nanostructured ferritic alloy to He ion irradiation and post irradiation annealing (PIA) at 750 °C for 10 and 100 h. The light element capability of APT was used to estimate the He concentration in the ferrite matrix. For these measurements, all precipitates, nanoclusters, and He bubbles were removed from the data with the use of iso-concentration and iso-density surfaces, respectively. For a dose of ~2.7 displacements per atom (dpa), the He concentration in the ferrite matrix for the as-implanted and 10 h PIA treatment was ~75 appm, but decreased to approximately half that value after the 100 h PIA treatment. No evidence for He clusters in the ferrite matrix was found in the APT atom maps. The distributions of He bubbles in the matrix, grain boundaries, and on the surfaces of nanoclusters and Ti(N,C,O) precipitates were determined for the above mentioned conditions. More than 90% of the He bubbles were present as isolated bubbles in the ferritic matrix and less than ~5% were on the surfaces of the nanoclusters in the ferritic matrix. The remainder were associated with the grain boundaries, with a small percentage on the surfaces of the Ti(N,C,O) precipitates. An approximately 10-nm-wide He-bubble-depleted zone was observed adjacent to some grain boundaries. The average size of the He bubbles was similar for the as-implanted and 10 h PIA treatment, and a small increase in the size of the He bubbles was observed after the 100 h PIA treatment. Number density increased significantly from the as-implanted to 10 h PIA state, with small changes from 10 to 100 h PIA. The amount of swelling in the high dose region was estimated from the additional volume generated by the He bubbles in the APT data and was found to increase from ~1% in the as-implanted and 10 h PIA conditions to ~3% after the 100 h PIA treatment. Concomitantly, the number of He atoms per unit volume in the He bubbles decreased by ~3-5 times. These experimental APT results indicate that additional nucleation of He bubbles occurred during the PIA treatment concurrent with the coarsening process.
Research sponsored by the Materials Sciences and Engineering Division, Office of Basic Energy Sciences, US Department of Energy, and by ORNL&’s Shared Research Equipment (ShaRE) User Program, which is sponsored by the Scientific User Facilities Division, Office of Basic Energy Sciences, US Department of Energy.

The stability of grain boundaries in nanostructured ferritic alloys is a major factor in their excellent response to high temperature creep and high dose radiation tolerance under extreme conditions. In these ultra-fine grained (20-400 nm) materials, the grain boundaries exhibit complex non-equilibrium shapes and curvatures as well as many triple lines. A novel atom probe tomography method based on a 3D Hough transformation has been developed that enables the full five degrees of freedom of the orientation relationship between the individual grains to be measured. In addition, the extent of solute segregation for all elements over the surface of the grain boundary may be estimated with 1 nm by 1 nm spatial resolution. This approach enables variations in the solute excess for the elements with the habit plane and curvature of the grain boundary to be evaluated. The new method also enables low levels of solute segregation, such as observed after low temperature ion irradiation, to be quantified. The method has been applied to mechanically-alloyed flakes of a 14YWT nanostructured ferritic alloy after isothermal aging at different temperatures as well as high temperature and high dose heavy ion irradiated material. The results confirm previous atom probe tomography and transmission electron microscopy results indicating that solute segregation of chromium and tungsten and preferential precipitation occurs on the grain boundaries. The innovative high resolution two-dimensional results provides additional indications that the solute excesses across the surface of the grain boundaries are not uniform and correlate with changes in its local curvature. These features pin the grain boundary against grain growth and provide the stability for excellent creep properties.
Research sponsored by the Materials Sciences and Engineering Division, Office of Basic Energy Sciences, US Department of Energy, and by ORNL&’s Shared Research Equipment (ShaRE) user facility, which is sponsored by the Scientific User Facilities Division, Office of Basic Energy Sciences, US Department of Energy.

Zirconium hydride precipitates in zirconium-based fuel cladding in light water reactors. The hydride is brittle and known to deteriorate mechanical strength of the cladding. The hydride has crystal habit planes with zirconium matrix, such as (111)δ-ZrHx//(0002)α-Zr. Therefore, relation between the crystal orientation and the mechanical properties of zirconium hydride is important for evaluation of hydride-precipitated claddings. In the present study, bulk δ-zirconium hydride samples were synthesized and the mechanical properties were measured with EBSD observation.
Bulk zirconium hydride was prepared using modified Sieverts&’ apparatus from arc-melted zirconium ingot. The ingot was annealed at 1073 K-1173 K at high vacuum before the hydrogenation. The surface was polished by argon ion milling. The indentation test was performed on the surface using dynamic Vickers hardness tester. The crystal orientation of the indentation area was observed by EBSD analysis.
The microstructure of zirconium hydride depends on the fabrication condition. Slowly cooled and hydrogenated samples have strain-free grains. On the other hand, quickly fabricated hydride includes many twinned crystals in the structure. The indentation hardness and modified Young&’s modulus of quickly fabricated samples are lower than those for the strain-free samples. It can be one of the reasons for differences in reported mechanical properties of zirconium hydride.

Uranium dioxide is currently the most widely used fuel in nuclear power plants. The fission of uranium atoms creates irradiation defects and produces large quantities of volatile fission products, which is a limiting factor for the efficiency of the fuel.
Vacancy type defects in UO₂, empty or containing fission products, can be studied by positron annihilation spectroscopy (PAS), a non-destructive method allowing one to probe the electronic structure of materials. This experimental technique is based on the principle that when a positron gets trapped in a vacancy-type defect, its average lifetime increases, what can be measured. In PAS the neutral and negative defects can be distinguished, while positive ones cannot be detected, as they repel positrons. Since the positron annihilation spectroscopy experiments themselves do not provide the information on the types of defects, the results must be complemented with those of other experimental techniques or electronic structure calculations.
We present positron lifetimes calculated for UO₂ in the density functional theory (DFT) framework using self consistent calculation schemes and taking into account the relaxation effect. We used the DFT+U formalism to describe the strong correlations between the 5f electrons. Additionally, we calculated the defects formation energies to predict their detectability in PAS. We compare our results with the experimental observations.

The use of ion irradiation to form nanoceramic materials and evolve functional properties has not yet been fully explored. In the past, the effect of ion-induced ballistic mixing and chemical segregation at heterointerfaces has been observed. Yet still lacking is a fundamental understanding to connect evolving material properties under irradiation to simultaneous changes in material structure and radiation tolerance. Beyond the current understanding is the tailored use of ion beams to substantially lead to, enhance, or degrade material properties remains aloof.
Key to interpreting and eventually developing a predicative capability of composite interfaces is studying heterointerfaces at the atomic scale and providing vital chemical information following synthesis and subsequent ion irradiation. Pivotal to the formation of composite ceramic interfaces is interpreting the evolution of microstructure and possible formation of nanoceramics on the basis of point defect energetics and thermodynamic instabilities. Within this fundamental framework, the concomitant role of interface structure and chemistry is thereby studied in connection with the formation of composite interfaces and evolutionary radiation damage.
In this work, we examine the microstructural and chemical evolution of STO-MgO and (Fe,Cr)-TiO2 in connection with the formation of nanoceramic composites following synthesis and/or ion irradiation. We have used aberration corrected scanning transmission electron microscopy (STEM) and STEM based spectral imaging to study the formation of composites at ceramic interfaces. Based on the results, we hypothesize the use of interfacial composites on the degradation or enhancement of functional and physical properties based on first-principles based calculations.
This work was supported by Center for Materials at Irradiation and Mechanical Extremes, an Energy Frontier Research Center funded by the U.S. Department of Energy, Office of Basic Energy Sciences under Award Number 2008LANL1026, the Laboratory Directed Research and Development program, and ORNL&’s Shared Research Equipment (SHaRE) User Facility, sponsored by the U.S. Department of Energy, Office of Basic Energy Sciences.

Tungsten is a candidate material for the divertor and for first wall armour of future thermonuclear fusion reactors (ITER and DEMO). In such irradiation conditions it is well known that the microstructure and thus the properties of materials will evolve.
In this perspective the fate of irradiation induced defects (helium atoms, vacancies, self interstitials and the complexes they can form) has to be understood. This means that the properties of each species as well as the way they interact with each other have to be determined.
To this end, we have implemented an experimental approach which allows to investigate different effects by choosing the conditions of damage and helium introduction: introduction of He atoms only, co- introduction of He and damage with various He/dpa ratio and finally the creation of pure damage only. The evolution of the tungsten microstructure and the helium behavior are investigated. In adaptated conditions the experimental data are used to optimize the parametrization of a modeling code based on Object Kinenetic Monte Carlo.
Helium is introduced at various energies using implantation with either ion accelerators or plasma reactor (0.32 - 800 keV) and damage is induced by the He ions themselves or independently by irradiation with 20 MeV W ions. Nuclear Reaction Analysis is used to mesure He profiles and desorption, Positron Annihilation Spectroscopy to probe vacancy defects and Transmission Electron Microscopy to observe large defects such as cavities or bubbles. The OKMC simulations are performed using the LAKIMOCA code parameterized on Density Functional Theory data. The binary collision approximation code Marlowe is used to calculate the implantation induced defects and He profiles which are introduced in the OKMC code at the appropriate flux. These simulations allow to predict the depth profile and the type of the irradiation induced defects after implantation and annealing in appropriate experimental conditions.
The results obtained will be discussed with a special emphasis on how simulations can help to interpret the experimental data and to discriminate between the different possible mechanisms : vacancy migration, agglomeration, and the role of He in these phenomena as well as He desorption, trapping and clustering.

Helium and small amounts of other noble gases will be present in fusion reactor materials, such as the candidate divertor material tungsten. The effect of, in particular, helium bubble formation on the tungsten surface morphology and mechanical and thermal properties will be vital for long term reactor operation. At a temperature of 1000-2000 K, a low density surface fuzz is formed at a tungsten surface subject to low energy He or He+H plasma. The underlying mechanisms and temperature dependence are still not well understood, though helium bubble formation and bursting are likely to play an important role. Fuzz has so far not been experimentally observed from exposure to other noble gas plasmas, and an atomistic study of the properties of small clusters can provide insight into the differences between helium and other noble gases.
While helium in tungsten and other metals has been, and continues to be, extensively studied using ab initio, molecular dynamics (MD) and longer length and time scale simulations, much less attention has been paid to other noble gases. Using density functional theory (DFT) we have compared small clusters of helium, neon and argon in tungsten. Similarly to helium in tungsten, neon and argon are strongly self-binding and have low migration barriers. A new tungsten-neon interatomic pair potential was constructed based on the DFT data, to allow MD simulations of neon in tungsten. Previously the formation, trap mutation and mobility of helium clusters in tungsten have been studied in detail using MD and noting and understanding the differences between neon and helium will help explain the special role of helium in the fuzz formation process.

High thermal conductivity is a key performance metric for nuclear fuels. In oxide-based fuel systems, which are electronic insulators, heat is carried by phonons; their dynamics dictate the thermal conductivity of the material. The microstructure of a UO₂ fuel pellet evolves constantly during burnup, resulting in a progressive degradation in the thermal transport properties. We present the results of thermal conductivity calculations within the lattice dynamics framework in UO₂ using electronic-structure calculations for the interatomic force constants. Comparison of the phonon lifetimes with the recently measured lifetimes is provided. We have also performed large scale molecular dynamics (MD) simulations to determine the effects of microstructure - grain boundaries and dislocations - on the thermal conductivity of UO₂. We show how insight from such simulations can be integrated into the FRAPCON fuel performance code. This work is supported by the Center for the Materials Science of Nuclear Fuel, a DOE-BES Energy Frontiers Research Center.

The structure of CrPuO₃ has been predicted and confirmed by modelling and experimental techniques. The formation of the CrPuO₃ precipitate in nuclear fuel has been studied in relation to how it will affect the high burn-up structure and act as a getter for minor actinides, including Cm3+ and Am3+. The thermal conductivity has also been predicted and compared to the host UO₂ lattice. CrPuO₃ was found to take up a disordered perovskite structure, where the cations occupy the normal perovskite sites but the oxygen ions are shifted from their site forming a super-structure. The intrinsic defect properties were then studied using both empirical and ab-initio techniques to confirm the compound&’s response to radiation. Solution of trivalent oxides was then studied, as well as trivalent cation partition from stoichiometric UO₂. The thermal conductivity of the perovskite phase was found to be far lower than that of the UO₂ phase having large implications on fuel performance.

A solution to decrease the radiotoxicity and heat load of ultimate spent nuclear fuels would be to transmute Am produced by neutron captures during fuel irradiation. In this context, U-Am oxides are promising fuels for transmutation. They remain however little-known among nuclear fuels. As a first approximation, they can be considered similar to U-Pu oxides, based on the closeness between Pu-O and Am-O phase diagrams. The particularities of Am, especially in terms of oxygen potentials and 241Am self-irradiation effects, limit this comparison. A quite unexpected cationic charge distribution was for instance reported for Am/(U+Am) ratios between 10 and 20 at.%: Am is only present as Am(+III), while U(+IV) is partially oxidized to U(+V). The U(+V) content seems to be related to that of Am(+III), meaning that a charge compensation mechanism may occur between U(+IV) and Am(+IV). It remains unclear what would happen if the Am content is increased, as no data were reported for Am/(U+Am) ratios above 20 at.%. Moreover, the existence of a miscibility gap in the U-Am-O system is expected (based on the comparison to the U-Pu-O system) but was never observed due to the lack of data concerning high Am/(U+Am) ratios. Consequently, no ternary U-Am-O phase diagram has ever been reported.
In this study, we report the synthesis of new U-Am mixed-oxide samples with high Am contents and their study by XRD (X-ray diffraction) at both RT (room temperature) and HT (high temperature) and XAS (X-ray absorption spectroscopy). The synthesis of samples with five different Am/(U+Am) ratios comprised between 7.5 and 50 at.% was performed from UO2+δ and AmO2-δ precursors through the UMACS (uranium minor actinide conventional sintering) process. Based on XRD characterization, the samples are monophasic whatever the Am content, exhibiting a single fluorite-type structure (Fm-3m). The results obtained by XANES (X-ray absorption near edge structure) confirm the stability of Am(+III) in these compounds whatever the Am content. A significant influence of the composition on U valence is however noted. The oxidation state of U increases with Am content, the 50%-Am spectra at the ULIII edge being between those of U(+IV/+V)4O9 and U(+V/+VI)3O8, hence the possible presence of U(+VI) in the high Am-content compounds. Through EXAFS (extended X-ray absorption fine structure), interatomic distances consistent with the presence of reduced Am and oxidized U are obtained. The behavior of these compounds in terms of oxidation is also affected by this charge distribution. As was observed in-situ by RT- and HT-XRD, phase transitions occur during the oxidation of samples with Am/(U+Am) ratios above 30 at.%, which may be associated with the presence of the expected miscibility gap. These results, which will be further described during the oral presentation, thus confirm the particularities of U-Am mixed oxides and will be useful for thermodynamic modeling of the U-Am-O ternary phase diagram.

Compound energy formalism (CEF) Gibbs energy models were developed forU_1-yGd_yO_2±x and U_1-yThyO_2±x and are extensions of the successful representation of the UO2±x phase. Gd and Th readily dissolve in the fluorite matrix of urania; Gd is important since it can be used as a burnable poison while Th is fertile material. Therefore, a CEF model for U_1-yGd_yO_2±x and U_1-yThyO_2±x are useful for better understanding oxide fuel with the Th+4 allowing benchmarking of fluorite structure behavior with multiple cations. Oxygen potential measurements for known compositions of urania thoria solid solutions are taken by thermogravimetric analysis (TGA) and compiled with existing data from the literature. The body of work reporting equilibrium oxygen dissociation pressures for U_1-yGd_yO_2±x was judged sufficient. These data are then used to fit the interaction and other selected parameters in the CEF model. The models for the U_1-yGd_yO_2±x and U_1-yThyO_2±x solid solutions are used to calculate ternary phase equilibria and can be combined with other representations of actinide and fission product containing fluorite urania phases to develop multi-component models within the CEF framework.
Research supported by the US Department of Energy, Office of Nuclear Energy, Fuel Cycle R&D Program.

The microstructure of a UO2 fuel pellet evolves considerably during burn-up. Such microstructural evolution has considerable effects on the thermal transport properties. To better understand the phonon scatterings at grain boundary (GB), we have performed simulations of phonon scattering from grain boundaries in UO2 using phonon wave packet dynamics. The results enable the connection between the energy transmission and grain boundary energy to be elucidated. Moreover, by comparing the Kapitza conductance calculated using the energy transmission results with that from non-equilibrium molecular dynamics, the role of anharmonicity at the interfacial thermal transport is revealed. This work is supported by the Center for the Materials Science of Nuclear Fuel, a DOE-BES Energy Frontiers Research Center.

The formation enthalpies and defect ordering tendencies of trivalent fission products are calculated using atomic-scale simulations. Low-energy defect structures are first screened with calculations based on core-shell ionic-potential models. The most stable structures are then studied by density functional theory with Hubbard-U corrections (DFT+U), and by hybrid DFT methods. The calculations consider compositions of ordered compounds where the trivalent cations are charge-compensated by oxygen vacancies and holes. Calculated formation enthalpies and cation-ordering tendencies for vacancy-compensated systems are compared with recent calculations for doped thoria.

Based on density functional theory (DFT) and empirical potential calculations, the diffusivity of Xe atoms in UO₂ nuclear fuel has been calculated for a range of stoichiometry (i.e. UO_2±x), under both out-of-pile (no irradiation) and in-pile (irradiation) conditions. This was achieved by first deriving an expression for the activation energy that accounts for the type of trap site that the fission gas atoms occupy, which includes the corresponding type of mobile clusters and the charge state of these defects. In the next step DFT calculations were used to estimate migration barriers and internal energy contributions to the thermodynamic properties, while calculations based on empirical potentials were used to estimate defect formation and migration entropies (i.e. pre-exponentials). From these calculations, the fission gas diffusivity is predicted to depend strongly both on the UO_2±x stoichiometry and the irradiation conditions, which govern the concentration of uranium vacancies. The results agree favorably with available experiments for both intrinsic and radiation-enhanced diffusion. Nevertheless, the models and the calculations contain several approximations that deserve additional attention. The predicted fission gas diffusion rates are implemented in the BISON fuel performance code and fission gas release from a fuel rod irradiation experiment is simulated. A reasonable agreement was observed between the fission gas release predicted using the theoretical model and the available experimental data. The agreement is the same or better than existing empirical models, but the close connection to physical diffusion mechanisms should make it more transferable.

During reactor operations, a large thermal gradient is built up radially in UO2 fuel pellets due to the low thermal conductivity of the fuel. Some extended defects such as grain boundaries (GB) and voids can migrate under the thermal gradient leading to fuel restructuring. To elucidate the thermal gradient driven GB migration mechanisms, molecular dynamics simulations with the Basak potential were conducted to investigate how GB character affects the GB mobility. A number of GBs of different misorientation angles and energies were studied, all having either tilt or twist character. It was found that these GBs migrate due to the thermal gradient at very different rates, depending on their characters. Some GBs, such as tilt Σ5 and tilt Σ29, migrate very fast, while others, such as tilt Σ25 and Σ3 twin boundary do not migrate within the simulation time. Analysis of the displacement field showed that GBs migrate via the diffusion mechanism, which is different from the dislocation gliding mediated mechanism in shear-driven coupled motion. The trend of GB mobility calculated under thermal gradient is consistent with that calculated under zero driving force using the random walk method. The correlation between GB mobility and GB self-diffusion is also discussed.

Atomistic understanding the mechanisms of oxygen defect clustering in uranium dioxide (UO2) is an important step towards modeling of microstructure evolution in this material under extreme conditions such as high temperature and irradiation. In this work, Parallel kinetic Monte Carlo (KMC) and thermodynamic analysis are used to investigate the dynamic clustering of oxygen defects in UO2 under different hyperstoichiometric conditions, where the predominant defects are oxygen interstitials. In particular, the influence of temperature on oxygen clustering dynamics will be exploited. The results such as the diffusion coefficients, and cluster population distribution and evolution will be evaluated and compared to available experimental data. This work is supported by the Center for Materials Science of Nuclear Fuel, an Energy Frontier Research Center (EFRC) funded by the U.S. Department of Energy, Office of Science, Office of Basic Energy Sciences under Award Number FWP 1356.

The behavior of oxide fuel under irradiation is extremely complex involving generation of over 60 additional elements with resulting major changes in chemistry/defects, void formation and microstructural evolution, and resultant changes in thermal properties. All these effects are interrelated, with many of them driven by changes in fuel phase composition and formation of secondary compounds. In particular, substantial progress has been made in understanding and simulating oxygen behavior as a result of the non-uniform generation of fission and transmutation products, including gradients and the potential for oxidizing clad inner surface. In addition, improved representations of zirconium clad hydriding have been developed, including 2-D simulations. This paper will report on the results of modeling fuel thermochemical behavior and future challenges.
Research supported by the US Department of Energy, Office of Nuclear Energy, Nuclear Energy Advanced Modeling and Simulation Program.

Tungsten is a leading candidate material for the diverter in future nuclear fusion reactors. Previous experiments have demonstrated that surface defects and bubbles form in tungsten when exposed to helium and hydrogen plasmas, even at modest ion energies. In some regimes, between 1000K and 2000K, and for He energies below 100eV, “fuzz” [1] like features form. The mechanisms leading to these surfaces comprised of tungsten “tendrils” which include visible helium bubbles are not currently known. The role of helium bubble formation in tendril morphology could very likely be the starting point of these mechanisms.
Molecular dynamics (MD) simulations are well suited to describe the time and length scales associated with initial formation of helium clusters that eventually grow to nano-meter sized helium bubbles. MD simulations also easily enable the modeling of a variety of surface such as single crystals, grain boundaries or “tendrils”.
We investigate the effects on such tungsten surfaces and morphologies of the cluster formation process resulting from the implantation of a flux of He, with an implanted depth corresponding to a 60eV ion energy. We not only analyze the atomic retention and atom depth distribution as a function of time but more importantly the cluster size distribution and tungsten surface evolution.
We investigate the temperature dependence by analyzing the clustering process at 500K, 1200K and 2000K. We also look into the effect of tungsten morphology by comparing the results of single crystal (100) and (110) tungsten, to those of tungsten with a grain boundary, then to those of tungsten with a tendril at the surface.
We find that during the bubble formation process, He clusters create self-interstitial defect clusters in W by a trap mutation process, followed by the migration of these defects to the surface that leads to the formation of layers of adatom islands on the tungsten surface. As the helium clusters grow into nanometer sized bubbles, their proximity to the surface and extremely high gas pressures leads them to rupture the surface thus enabling helium release. Helium bubble bursting induces additional surface damage and tungsten mass loss which varies depending on the nature of the surface.

Reference:
[1] M.J. Baldwin, Nucl. Fusion 48, 035001 (2008)

Tungsten is a leading candidate material for the divertor in future nuclear fusion reactors. Previous experiments have demonstrated that surface defects and bubbles form in tungsten when exposed to helium and hydrogen plasmas, even at modest ion energies. In some regimes, between 1000K and 2000K, and for He energies below 100eV, “fuzz” [1] like features form. The mechanisms leading to these surfaces comprised of tungsten “tendrils” which include visible helium bubbles are not currently known. The role of helium bubble formation in tendril morphology could very likely be the starting point of these mechanisms.
We investigate the effects on such tungsten surfaces and morphologies of the cluster formation process resulting from the implantation of a flux of He, with an implanted depth corresponding to a 60eV ion energy. We investigate the effect of tungsten morphology by comparing the results of single crystal (100), (110) and (111) tungsten, to those of tungsten with a grain boundary, then to those of tungsten with a tendril at the surface.
Previous MD simulations [2] have shown that during the bubble formation process, He clusters create self-interstitial defect clusters in W by a trap mutation process, followed by the migration of these defects to the surface that leads to the formation of layers of adatom islands on the tungsten surface. Now, for single crystals, we show that as the helium clusters grow into nanometer sized bubbles, their proximity to the surface and extremely high gas pressures leads them to rupture the surface thus enabling helium release. Helium bubble bursting induces additional surface damage and tungsten mass loss which varies depending on the nature of the surface. Furthermore, we find that in the case of helium implantation in tendrils, small helium clusters frequently escape the tendril leading to an equilibrium state with stable amounts of helium bubbles in the tendril and no major tendril surface damage. Unlike with single crystal surfaces, the tendril's high surface to volume ratio enables it to resist to surface damage caused by helium bubble evolution and bursting.
Then, to further understand the bubble bursting process in single crystals, MD simulations have been used to systematically study the pressure evolution and bursting behavior of sub-surface helium bubbles and the impact on tungsten surface morphology. We investigate how parameters such as shape and size of the bubble, temperature, tungsten surface orientation and ligament thickness above the bubble influence bubble stability and surface evolution. The tungsten surface is roughened by a combination of adatom “islands”, craters and pinholes. This study provides insight into the mechanisms and conditions leading to various tungsten topology changes, which we believe are the initial stages of surface evolution leading to the formation of nanoscale fuzz.
[1] M.J. Baldwin, Nucl. Fusion 48, 035001 (2008)
[2] F. Sefta, Nucl. Fusion 53, 073015 (2013)

Tungsten is a candidate material for the divertor in fusion reactors. The divertor will be subject to intense, low energy (1-100 eV) hydrogen isotope and helium bombardment from the plasma. He and H in a material can cause changes in thermal and mechanical properties, such as swelling, ductile to brittle transition temperature, bubble formation and nanofuzz formation. Fuel retention is a serious issue, in particular due to limits on tritium allowed in reactors.
Molecular dynamics is a valuable tool to study the energetics and structures of H and He clusters and many radiation damage phenomena that happen on short time and length scales, up to nanoseconds and millions of atoms. We have previously shown that hydrogen tends to decorate helium bubbles in tungsten, quickly being pushed to the edges from inside helium bubbles, as well as being trapped while diffusing in the tungsten matrix. A single hydrogen atom is bound within the first couple of tungsten layers surrounding a helium bubble with a binding energy of about 1 eV.
The amount of hydrogen that can be bound at the edge of a bubble, as well as the energy landscape of a hydrogen atom near and in the hydrogen layer, has been studied for different bubble sizes and compositions, comparing to voids with no helium. Under helium plasma exposure, bubbles grow as helium diffuses into the bubble and tungsten interstitials are ejected to relieve pressure. A hydrogen layer surrounding a bubble can affect both processes and bubble growth with and without a surrounding hydrogen layer was investigated. We have compared different inter-atomic potentials to assess the validity of the simulations.

In fusion reactors, plasma facing components (PFC) and in particular the divertor will be irradiated with high ion fluxes of low energy (~ 100 eV) helium and hydrogen. Tungsten is one of the leading candidate divertor materials for ITER and DEMO fusion reactors. However, the behavior of tungsten under high dose, coupled helium/hydrogen exposure remains to be fully understood.
The PFC response and performance changes are intimately related to microstructural changes, such as the formation of point defect clusters, helium and hydrogen bubbles or dislocation loops. Computational materials modeling has been used to investigate the mechanisms controlling microstructural evolution in Tungsten following high dose, high temperature helium exposure.
The aim of this study is to understand and predict primary defect production and defect diffusion, clustering and interaction of tungsten surface exposed to low energy helium irradiation at high fluences (~ 1026 He/m2) and temperatures ( ~ 1000 K).
We report results from a spatially-dependent cluster dynamics model based on reaction-diffusion rate theory. The model was improved to be able to include very large helium vacancy clusters expected to form under these irradiation conditions. The key parameter inputs to the model (Diffusion coefficients, migration and binding energies, initial defect production) are determined from a combination of atomistic materials modeling and available experimental data.
We find good agreement between the model and analytical work. We also compare results with molecular dynamics simulations and existing experimental results.

Tungsten is a leading candidate material for the divertor in future nuclear fusion reactors. Previous experiments have demonstrated that surface defects and bubbles form in tungsten when exposed to helium and hydrogen plasmas, even at modest ion energies. In some regimes, between 1000 K and 2000 K, and for He energies below 100 eV, “fuzz” [1] like features form. The mechanisms leading to these surfaces comprised of tungsten “tendrils” which include visible helium bubbles are not currently known. Experiments indicate that tungsten tendrils do not form under other noble gas plasma exposure such as neon or argon. Understanding why tungsten “fuzz” does not form under neon exposure may help isolate the specific role of helium in “fuzz” growth.
Molecular dynamics (MD) simulations are well suited to describe the time and length scales associated with initial clustering and bubble formation mechanisms of helium and neon on tungsten. Initial noble gas clusters form and eventually grow to nano-meter sized bubbles. MD simulations also easily enable the modeling of a variety of surfaces such as single crystals, grain boundaries or “tendrils”.
Previous MD simulations [2] have shown that during the bubble formation process, He clusters create self-interstitial defect clusters in W by a trap mutation process, followed by the migration of these defects to the surface that leads to the formation of layers of adatom islands on the tungsten surface. As the helium clusters grow into nanometer sized bubbles, their proximity to the surface and extremely high gas pressures leads them to rupture the surface thus enabling helium release. Helium bubble bursting induces additional surface damage and tungsten mass loss which varies depending on the nature of the surface.
Here, we compare the effect of helium implantation in tungsten to that of neon. We investigate the cluster formation process resulting from the implantation of a flux of Ne, with an implanted depth corresponding to a 60 eV ion energy. We not only analyze the atomic retention and atom depth distribution as a function of time but more importantly the cluster size distribution and tungsten surface evolution, as a function of temperature, surface orientation and grain boundary effects.
References:
[1] M.J. Baldwin, Nucl. Fusion 48, 035001 (2008)
[2] F. Sefta, Nucl. Fusion 53 073015 (2013)

Helium from linear plasma devices and tokamak plasmas is known to cause the formation of microscopic features, termed “fuzz” or “coral,” on the surface of plasma-exposed materials after only a few hours of plasma exposure. The precise details of such surface modifications are as yet uncertain. This study examines the initial and intermediate stages of fuzz formation by large-length-scale molecular dynamics (MD) simulations of helium-implanted tungsten over time scales of up to microseconds using single-crystalline and polycrystalline supercell models of tungsten. The large-scale MD simulations employ state-of-the-art many-body interatomic potentials and implantation depth distributions for the insertion of helium atoms into the tungsten matrix constructed based on MD simulations of helium-atom impingement onto W surfaces under prescribed thermal and implantation conditions. The large-scale MD simulations reveal surface features formed via the sequence of helium implantation, diffusion of helium atoms and their aggregation to form bubbles, growth of bubbles and consequent production of tungsten self-interstitial atoms, organization of those atoms into prismatic loops, glide of those loops to the surface, and bubble rupture. The crystallographic orientation of the surface and the grain microstructure of the tungsten matrix, namely, the presence or absence of grain boundaries, are found to have strong effects on the resulting surface features. In particular, grain boundaries serve as sinks for helium atoms, resulting in more rapid growth of surface features in the vicinity of grain boundaries. Targeted calculations of helium bubble energetics as a function of distance from grain boundaries in tungsten and from the free surface provide thermodynamic and kinetic interpretations of the helium bubble formation processes observed in the simulations. Our findings have significant implications for the surface morphological evolution and the near-surface structural evolution of plasma-facing components in nuclear fusion reactors.

Beryllium is a leading candidate for plasma facing and neutron multiplier components in fusion reactor designs. Understanding the interaction of fusion products with impurities and second phase precipitates is crucial for the development of high tritium release, radiation resistant Be-alloys.
In the current work the partitioning and trapping of extrinsic species, including H and He, at second phase particles formed within the beryllium alloys are investigated by means of DFT simulations. It is predicted that additions of Fe react with any Al impurities (which are otherwise expected to decorate grain boundaries) to form stable AlFeBe4 precipitates. Further additions of Fe would form FeBe2 particles, whilst the FeBe5 phase was found to be metastable
These intermetallic compounds are predicted to getter other undesired impurities such as Si and Mg, whilst not posing a significant sink for H migration. Similarly, Li and O are expected to remain in bulk Be as a substitutional and interstitial species respectively. Finally, C and He show no significant difference in solution energy between accommodation in the intermetallics and in the bulk metal.
Implications with regards to in-reactor performance of Be-alloys are discussed.

Tungsten is a candidate material for first wall tiles, and particularly the divertor, in future fusion reactors, owing to its extremely high melting point and favorable transmutation characteristics under neutron irradiation. Understanding the evolution of radiation-induced damage is critical in providing insight into the mechanical properties of such material, particularly when IFMIF and other experiments capable of reproducing a similar environment do not exist [1]. In this large and complex problem, the thermal annealing of ½<111> and <100> prismatic dislocation loops is of interest. Experiments capturing the dynamics of such events are limited in bulk [2,3] and in-situ TEM [4]. This has limitations in the annealing temperatures it can reach before compromising the performance of the TEM, as well as potentially introducing artefacts due to the presence of the thin film surfaces, which act as strong dislocation loop sinks.
In this work we develop, validate, and employ a spectral methods code based on the method presented in [5] to efficiently calculate the image stresses experienced by dislocations in the thin foils used for in-situ TEM. We examine the image forces on ½<111> and <100> loops in such films, along with simulations of more complex structures of 3 to 30 mixed-type loops. The results are compared with simulations of the same initial microstructure in a bulk material, and the impact of the free surfaces is quantified.
[1] M. Reith et al, Journal of Nuclear Materials 432 (2013) 482-500.
[2] F. Häusermann, Phil. Mag. 25, 537 (1972); 25, 561 (1972); 25, 583 (1972)
[3] W. Jäger, M. Wilkens, Phys. Stat. Sol. (a) 32, N.1 (1975)
[4] X. Yi, M. L. Jenkins et al, Phil. Mag. 93, 14 (2013)
[5] C. Weinberger et al, Modelling. Simul. Mater. Sci. Eng. 17 (2009) 075007

The implantation of helium (He) atoms has significant implications for the surface morphological evolution and the near-surface structural evolution of plasma-facing components in nuclear fusion reactors. In tungsten (W), such interstitial He atoms are very mobile and aggregate to form clusters of different sizes; the smaller of these clusters also are mobile and their diffusional transport mediates the evolution of surface morphology and the structural evolution of the near-surface regions of the plasma-exposed material.
In this presentation, we report results of a systematic, comprehensive computational study of mobile He-cluster transport in W. Our analysis bridges atomistic and continuum modeling and simulation for a rigorous and quantitative description of drift and diffusional transport of such He clusters in near-surface W regions. In addition to Fickian diffusion, the analysis takes into account drift fluxes for cluster transport driven by surface segregation forces; such thermodynamic driving forces become significant and induce substantial drift fluxes in near-surface regions. Our modeling approach links hierarchically atomic-scale computations with continuum drift-diffusion models for the evolution of the cluster concentration fields in the near-surface region; the atomic-scale modeling consists of molecular-dynamics (MD) simulations of mobile cluster diffusion, molecular-statics (MS) computations of the energies of structurally relaxed He-cluster configurations as a function of their distance from the surface, and computations of the corresponding optimal cluster migration pathways employing the climbing-image nudged elastic band (NEB) method. The cluster size n (ranging from 1 to 7 He atoms) and the surface crystallographic orientation are important parameters in the study. In addition to the fundamental and quantitative understanding of mobile He-cluster mass transport in near-surface W regions, this study provides the required closure relations for the drift and diffusive fluxes in continuum cluster mass transport (drift-diffusion) models through predictions of diffusion coefficients and segregation potentials as a function of distance from the surface. This transport modeling also is extended to the analysis of cluster transport near sinks other than surfaces, such as grain boundaries (GBs), with emphasis on GBs in near-surface regions and the corresponding combined (multi-sink) drift effects due to segregation forces.

High-Cr ferritic/martensitic steels are strong candidates for future fission and fusion reactors. However these materials can undergo a decomposition between Fe- and Cr-rich phases, as well as radiation induced segregation at point defect sinks. We present here atomistic kinetic Monte Carlo (AKMC) simulations of these phenomena, based on diffusion by vacancy and self-interstitial mechanisms. The point defect migration barriers are computed with a simple broken-bond model, using composition and temperature dependent pair interactions fitted on ab initio calculations. This model reproduces the asymmetrical thermodynamic properties of the alloy, especially the change of sign of the mixing energy at low concentrations. It also takes into account the acceleration of diffusion that occurs at the transition between ferromagnetic and paramagnetic states. AKMC simulations, and the comparison with 3D atom probe and small-angle neutron scattering experiments, show that this acceleration strongly affects the kinetics of decomposition during thermal ageing, especially in concentrated alloys. Simulations of segregation at grain boundaries under irradiation have been performed. The segregation behavior and the evolution of chromium concentration profiles are analyzed by measurements of the Onsager coefficients that control the coupling between fluxes of atoms and point defects.

The mobility of screw and edge dislocations that control plasticity in metals depends strongly on the alloy composition and impurity interactions with dislocations. We use density functional theory to study segregation of Cr and Ni atoms to the dislocation core and its
impact on mobility of 1/2<111> screw and 1/2[111](110) edge dislocations in bcc Fe. Periodic arrays of dislocations were used to model the dislocation core structures; quadrupoles in the case of screw dislocations and dipoles for edge dislocations. The influence of magnetism on the dislocation core structure and interaction of Cr and Ni impurity atoms with the core is demonstrated. Effects of Cr and Ni atoms on the dislocation core structure and mobility are discussed.
This work was supported by the Center for Defect Physics, an Energy Frontier Research Center funded by the US Department of Energy, Office of Science, Office of Basic Energy Sciences.

Reduced-activation ferritic/martensitic steels have been developed as the prime candidate materials for experimental fusion reactors. To estimate irradiation effects in fusion reactor components, multiple-scale modeling has been studied. Modeling activities for irradiation induced microstructural change is quite effective to enhance the capability to predict mechanical properties of the materials during irradiation. Defect activation energies such as vacancy and interstitial migration energies should be estimated to obtain fundamental parameters for the modeling. In this study, pure iron and Fe-based model alloys including minoe elements have been irradiated by electron and ion beams using a high voltage electron microscope in order to clarify the effect of minor elements on microstructure evolution, especially loop formation and defect activation energies.
Growth rate and saturated number density of dislocation loops have been measured and Arrhenius plotted to estimate the migration energies of point defects. Electron irradiation experiment indicated that the net migration energy of vacancy in a high purity iron tended to be lower compared to that in low purity pure iron and model alloys. Furthermore, vacancy migration energies in all the specimens including more carbon was slightly higher than that including more nitrogen.

High chromium ferritic-martensitic (F-M) steels are one of the promising structural materials for future nuclear power plants, combining corrosion resistance, swelling resistance, adequate toughness and elevated-temperature strength during service. Some studies have shown that above 12% Cr, F-M steels can exhibit hardening and embrittlement after thermal ageing due to α-α' phase separation. However, precise knowledge of the phase separation at operating temperature range is key for selecting optimal alloy compositions.
Six Fe-Cr alloys (3 to 18 at.% Cr,), neutron irradiated at 553 K at 1.6 dpa in the INL ATR reactor, were characterized by atom probe tomography (APT). Cluster analysis confirms the presence of α' for Cr concentration greater than 9 at.%. Apart from α' precipitates, different family of clusters were observed for different alloys, including Ni-Si clusters, clusters containing Ni, P, Si and Cr, and Si-enriched dislocation loops. Along with these microstructural changes occurring in the matrix, Cr segregation analyses were performed on selected grain boundaries in all six alloys. The results will be discussed in terms of possible dependences with Cr concentration and grain boundary orientation.

Iron-Nickel alloys are perspective alloys as nuclear energy structural materials because of their good radiation effect tolerance and mechanical properties. Experimentally observed features such as effect of Ni content to radiation effects evolution are not understood yet but they are essential for creation of predictive models for accounting radiation effects. Extensive atomic-scale modelling of defects in dilute Fe-Ni alloys performed in this work has revealed new features related to solute effects in defects and mass transport. Thus, it was found that mobility of point defects, vacancy and self interstitial atom (SIA), is not affected by up to 1.6at.% of Ni whereas it affects dramatically to SIA clusters mobility. Mobility of SIA clusters suppressed by their attractive interaction with Ni solute atoms and the overall effect depends on the cluster size and Ni concentration. An expression for estimation the breaking effect was revealed based on calculated features of cluster - solute interaction. The results obtained allow to explain effects of Ni to the microstructure evolution and swelling observed experimentally in Fe-Ni alloys.

Materials submitted to continuous radiation are maintained in nonequilibrium conditions that may accelerate their microstructure evolution and/or and induce unexpected structural bifurcations.The understanding and quantification of such transformations remain an important challenge from a fundamental point of view and constitutes major issues for nuclear reactor safety. In metallic alloys, radiation produces large quantities of vacancies and self-interstitial atoms which are absorbed by microstructural defects called sinks (dislocation lines, dislocation loops, voids, grain boundaries, heterophase interfaces, etc.). The sink ability to capture point defects depends on various parameters like its geometry, its capacity to absorb point defects partially or totally, and on the existence or not of long range elastic interactions. In this work, we developed a phase-field model idoneous to quantify these kind of effect since it can easily incorporate the microelasticity theory associated to any dislocation network, and couple it to the diffusion of the migrating species. The numerical results are compared with analytical models in selected reference cases to prove its validity. The method is then applied to various complex configurations for which a numerical approach is required (interacting dislocations in anisotropic crystals). It is shown that a correct evaluation of the sink efficiency of a dislocation network in presence of radiation flux leads to values significantly higher than those obtained by analytically solving the diffusion equation.

Alloy 617 is selected as the primary candidate material for the Gen IV High Temperature Gas Cooled Reactor (HTGR) intermediate heat exchanger (IHX) design due to its superior oxidation resistance and creep properties. However, selective internal oxidation along the alloy grain boundaries is a growing concern in IHX helium containing CO, CO2, H2, H2O and CH4 in ppm levels, at the IHX service temperature range of 750 - 850°C. Alloying elements such as Al, Ti and Si fail to form continuous sub-scales despite the formation of surface Cr2O3. Studies investigating the multicomponent gas-alloy interactions at high temperatures revealed that the surface scales such as Cr2O3 are indeed permeable to molecular oxidants. Permeation of gas molecules may alter the equilibrium at the alloy/oxide interface resulting in exacerbated oxygen dissolution in the alloy.
The objective of this study is to assess the gas permeability of the surface Cr2O3 scale by comparing the internal oxidation kinetics of Alloy 617 in IHX helium equivalent He-CO-CO2 environments with that at the dissociation oxygen partial pressure of Cr2O3 at 850°C. The latter condition represents the thermodynamic equilibrium attained at the alloy/oxide interface if a perfectly dense and well-adhered Cr2O3 would form at the surface of the alloy. To achieve this, samples were placed on one side of evacuated dumbbell shaped quartz capsules containing a mixture of Cr and Cr2O3 powders on the other side (henceforth will be called as Cr - Cr2O3 Rhines Pack samples).
Cr - Cr2O3 Rhines Pack sample tested at 850°C for 500 h exhibited approximately 5 times lower weight gain and 60% less internal Al2O3 depth than the IN617 samples did upon exposure to the He-CO-CO2 environments at the same temperature and for the same time. These preliminary results indicated that the surface Cr2O3 scale may be permeable to the molecular oxidants.

Long range order in Ni-Cr alloys is a concern for long time, elevated temperature applications in the nuclear power industry. High Cr alloys such as 690 and its weld metals have compositions near the Ni2Cr ordered phase. Ordering kinetics are poorly defined in Ni-Cr alloys and can change with cold work or alloy content. In this work, a combined experimental and computational approach is used to understand the role of solute type (Fe, Mo, Nb, etc.) and concentration on the rate of long range ordering. Model binary and ternary alloys (Ni-Cr-X) have been isothermally aged and characterized by x-ray diffraction to quantify the rate of ordering via lattice contraction. First principles calculations were performed using special quasi-random structures to determine the ordering energy for the model alloys. Lastly, cluster expansions and kinetic Monte Carlo simulations were employed to identify rate controlling diffusion mechanisms and independently assess the activation energy.

The MIAMI* facility at the University of Huddersfield permits the ion irradiation of thin foils in-situ in a transmission electron microscope (TEM). The facility has been developed with a particular focus on enabling the in-situ implantation of helium and hydrogen into TEM foils, necessitating ion energies in the range 1 - 10 keV. In addition, however, ions of a variety of species can be provided at energies of up to 100 keV, enabling studies to focus on the build up radiation damage in the absence or presence of implanted gas. Recent results on the development of He bubbles will be presented for SiC, W and (as a model fcc system) Cu; with this latter work including some insights gained into the formation of bubble super-lattices in metals. A detailed study will also be presented of a series of experiments aimed at shedding light on the origins of the dimensional changes known to occur in nuclear graphite under irradiation with both neutrons and ions. Single crystal graphite foils have been irradiated with 60 keV Ar ions in order to create a non-uniform damage profile throughout the foil thickness. This gives rise to varying basal-plane contraction throughout the foil which results in almost macroscopic (micron scale) deformation of the graphite. These observations will be presented and discussed with view to reconcile underlying point defect behaviour with recent ab-initio calculations of point defect mobilities in graphite.
*Microscope and Ion Accelerator for Materials Investigations

We investigate the contribution of the equation of state of helium to the thermodynamic driving force for preferential nucleation and subsequent growth of bubbles along grain boundaries. We present an effective equation of state for Helium gas in bubbles with atomistically derived parameterizations for both single crystalline bulk iron and copper, as well as selected twist boundaries for both material. We show that there are differences in pressure for bubbles of the same density and radius when they are located at grain boundaries or in bulk, and we relate these differences to the nature of the dislocations making the boundary. The question we want to answer in this work is why do bubbles prefer interfaces, and what does differentiate interfaces with respect to He trapping capacity? By doing this we are able to find the dependence of the critical radius, a relevant parameter that determines the onset of swelling, in terms of the material and GB properties.

The aim of our work is to examine how elastic stress, arising from nanovoids, and gas pressure in these voids influence the diffusion vacancy fluxes using approach [1,2], developed by us earlier, and to obtain a kinetic equation for the growth rate of voids (bubbles) in cubic metals. Our approach takes into consideration, that the strains can alter the surrounding atom configuration near the jumping one and consequently the local magnitude of the activation barrier. Knowing this change, we can calculate the atomic jump rate and obtain an equation for the vacancy flux. In this case, the vacancy flux depends on matrix of diffusion coefficients. Each of these coefficients depends on strain tensor components in the nonlinear way. In corresponding nonlinear equations, the functional dependence on strain is determined by coefficients, which are the main characteristics of the strain influence on diffusion (SID coefficients). These coefficients are very sensitive to atomic structure in the nearest vicinity of defect and still more to atomic structure of the saddle-point configuration. We have built advanced model to evaluate them. SID coefficient simulation is the first step of this work.
Then we use the results of theory of elasticity concerned with the displacement field around a sphere and obtain the equation for vacancy flux, where the influence of elastic strain near the void on this flux is taken into account. After substitution of the obtained expression for flux in the continuity equation, we have a diffusion equation in which the influence of the elastic strain near the void is taken into account. As it was done in LS theory [3], the growth of the voids is considered by using the mean-field approximation and steady-state solution of the diffusion equation around each void. The diffusion equation for vacancies in which the influence of elastic stress near the void on flux was taken into account is linearized and resolved. The obtained kinetic equation for the growth rate of voids contains the additional terms conditioned by strains, arising from voids. This feature distinguishes this equation from one that is known in LS theory [3] and changes the kinetic of void growth. Using this kinetic equation, we simulate void growth in fcc and bcc metals for different temperatures and vacancy supersaturation. Similar way we examine kinetic growth of gas bubble for different pressure in the bubbles.

1. A.V. Nazarov and A.A. Mikheev, J. Phys.: Condens. Matter 20 (2008) p. 485203.
2. Andrei Nazarov, Alexander Mikheev, Irina Valikova and Alexander Zaluzhnyi, Solid State Phenomena, 172-174 (2011) p.1156.
3. P.G. Cheremskoy, V.V. Slesov, V.I.Betehtin: Voids in Solid State (Energoatomizdat, M. 1990).

In the intense neutron irradiation environment of a reactor core, it is well known that inert gases can be produced, particularly in the fuel rod cladding materials. As the concentration of inert gas increases, a high density of nanoscale bubbles can form. These bubbles have a serious impact on the mechanical and physical properties of the material, especially when they form on grain boundaries, where they can lead to life-limiting embrittlement.
Under certain conditions, these nanoscale bubbles have been observed to self-assemble into a regular, periodic lattice - a bubble lattice [1,2]. Moreover, the structure of this bubble lattice imitates the lattice structure of the host material, albeit on a much larger scale, with three-dimensional lattices being formed in fcc and bcc crystals, and two-dimensional lattices formed in hcp structures (with the bubble lattice parallel to the basal plane). Confining the gas in such a stable manner keeps it away from the grain boundaries, offering the potential to extend the service lifetime of the components and reduce the risk of accidents due to material failure. However, bubble lattices only seem to form for certain specific values of the governing parameters, temperature and dose rate. Whilst there has been significant effort to unravel the fundamental mechanisms leading to bubble lattice formation in the past, there still is much debate as to their nature. Recent computational phase field models have had some success [3,4], but a fully predictive three-dimensional model remains elusive. The anisotropic diffusion of host metal self-interstitial atoms (SIAs) [5] is the most favoured mechanism in the literature.
To further elucidate the bubble lattice formation mechanisms, we have performed a systematic study by irradiating model metallic systems with low energy He ions in situ of a transmission electron microscope, using the MIAMI facility at the University of Huddersfield. This unique facility allows the bubble lattice formation to be examined dynamically, and under a variety of different conditions (temperature, dose rate). These results will be discussed in the context of the mechanisms described in the literature. Progress towards predicative three-dimensional stochastic simulations will also be presented, and compared with the available experimental data.
[1] J. H. Evans, Nature 229 (5284) 403 (1971)
[2] S. E. Donnelly, Radiation Effects 90 1 (1985)
[3] P.C. Millet and M. Tonks, Cur. Op. in Sol. State Mater. Sci. 15 125 (2011)
[4] Y. Li et al, J. Nucl. Mat. 407 119 (2010)
[5] J. H. Evans, J. Nucl. Mat. 132 (2) 147 (1985); Phil. Mag. 85 (11) 1177 (2005)

Nucleation and growth of helium bubbles in irradiated materials is a process that leads to changes in the physical and mechanical properties of crystalline solids. In particular, a large amount of helium is generated in high nickel alloys used in structural components of thermal fission reactors such as the heavy-water moderated CANDU reactors. Moreover the rate of He generation is dose dependent since it depends on the transmutation of Ni isotopes. In the case of fast breeder reactors, the generation of He is relatively low but the rate of radiation-induced defect generation is higher. Use of currently available experimental data provides an opportunity to develop a mechanistic description of the phenomena in the framework of reaction rate theory based on the mean field approximation. The most accurate description of gas bubble is given in terms of the evolution of their size distribution (SD). The SD follows a master equation describing the growth and dissolution of bubbles due to different reactions with mobile defects. The objective of this work is to study gas bubble evolution in stainless steels under different irradiation conditions through the use of a numerical solution of the master equation based on an accurate grouping method.