In this talk we discuss two topics. The first one is oxidation of UO₂ and the second one is diffusion of fission gases in UO_2±x and its implications for fission gas release models. Formation of hyperstoichiometric uranium dioxide compounds, UO_2+x, derived from the fluorite structure was investigated by density functional theory (DFT) calculations. Oxidation was modeled by adding oxygen atoms to UO₂ fluorite supercells. A similar approach was applied for studying reduction of U₃O₈. In agreement with the experimental phase diagram we identify stable line compounds at the U₄O_9-y and U₃O₇ stoichiometries. Additionally, we also found a new compound of the U₃O_7.3333 stoichiometry to be stable between U₃O₇ and U₃O₈. The calculated low-temperature phase diagram indicates that the fluorite-derived compounds are favored up to the UO_2.5, i.e. as long as the charge-compensation for adding oxygen atoms occurs via formation of U⁵⁺ ions. Once U⁶⁺ ions are required to achieve overall charge neutrality, the U₃O_8-y phase becomes more stable. According to our calculations the most stable fluorite UO_2+x phases at low temperature (0 K) are based on split quad-interstitial oxygen clusters. This cluster contains four excess and two displaced regular fluorite oxygen ions. It shares some features with the cuboctahedral cluster that is used in existing crystallographic models of U₄O₉ and U₃O₇, but the details are different. In order to better understand these discrepancies, the new structure models obtained from our simulations are analyzed in terms of existing neutron diffraction data. Finally, we discuss the importance of cluster formation for oxygen diffusion in UO_2+x. In order to better understand bulk fission gas behavior in UO_2±x we calculate the relevant activation energies using DFT techniques. Here we focus on Xe, since it is the most important fission gas in UO₂ nuclear fuels. By analyzing a combination of Xe solution thermodynamics, migration barriers and the interaction of dissolved Xe atoms with U vacancies, we demonstrate that Xe diffusion predominantly occurs via a vacancy-mediated mechanism. Next we investigate Xe transport on the (111) UO₂ surface, which is motivated by the formation of small voids partially filled with fission gas atoms (bubbles) in UO₂ under irradiation. Surface diffusion could be the rate-limiting step for diffusion of such bubbles, which is an alternative mechanism for mass transport in these materials. As expected, the activation energy for surface diffusion is significantly lower than for bulk transport. These results are further discussed in terms of engineering-scale fission gas release models.

The microstructure of UO₂ is altered significantly by irradiation during its operation as nuclear fuel. Being an oxide, defect disorder, vis-agrave;-vis off-stoichiometry, in this important material is a major aspect of the microstructure change process under irradiation. In general, defect disorder is controlled by the thermodynamic state variables, i.e. temperature and stress, as well as the composition or the oxygen partial pressure in the surrounding environment. We investigate the off-stoichiometric variation in UO₂ in terms of atomic defects and electronic carriers. A defect model that describes the defect disorder in terms of temperature and oxygen partial pressure is developed. The underlying parameters are drawn from density-functional theory literature. We extend the model to include the stress and electrostatic fields. Such extension relates the spatial off-stoichiometric variation to the microstructure elements in the oxide. In this regard, the impact of free surfaces and voids forming inside the bulk of UO₂ are investigated. The details of the surface interaction with the environment are understood using the ionosorption theory. The results establish the ground for the study of the oxide's time response to transients and irradiation conditions, thus employing the inventory of results from atomistic methods for a better understanding of the oxides structural behaviour. 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.

Understanding the long term behaviour of the UO₂ spent fuel in terms of alpha radiation damage and oxidation is a very important issue for the safety aspects of storage or disposal. Actinides contained in the spent fuel being alpha emitters, often with a long half life, will generate large quantities of helium and possibly cause chemical and physical modifications of the spent fuel matrix. Furthermore, many properties at the atomic scale like defect mobility and self diffusion are strongly dependent on the stoichiometry of UO₂. Although helium behaviour in stoichiometric UO₂ has been studied since the 1960s, there is still a lack of experimental studies for helium in the fcc lattice of non stoichiometric UO₂. In order to assess the effect of the stoichiometry on helium solubility, infusion experiments were performed on hyper and also on hypo-stoichiometric UO₂ samples to provide a comprehensive picture of helium behaviour in non-stoichiometric UO₂ matrix. Since the UO_2-x structure exists only at high temperature, La-doped UO₂ samples with various La-content were used. Supplementary work on the oxidation behaviour of UO₂ and changes in its local structure were studied by various spectroscopic and microscopy techniques. Helium behaviour was also investigated in 0.1 wt. % ²³⁸Pu-doped UO₂ samples. The radiation damage build-up and recovery processes were investigated together with the helium behaviour to better understand the possible state of the spent fuel after long storage times. This study will thus give new insights into the relevant aspects to consider for the behaviour of helium in stoichiometric UO₂, and defected systems such as nonstoichiometric UO₂ and Pu-doped UO₂.

Light ion implantations have generated a lot of interest over the years since they have major technological applications. In nuclear materials studies, they offer the prospect of understanding radiation effects in detail or developing new materials with enhanced radiation resistance properties. Indeed without using costly remote handling and characterization facilities, ion implantation techniques enable the study of effects resulting from neutron irradiations that make samples highly active. The primary effect of loading the surface of a material with foreign elements is to generate swelling of the crystal structure. However, the sample is generally not bulk irradiated but presents an implanted layer the thickness of which typically ranges between a few nanometers and a few microns. The question of how to relate expected swelling in a bulk or surface irradiated sample is therefore essential and we discuss here the first step towards understanding this relationship. Characterization of this swelling effect is usually performed using monochromatic high resolution X-Ray diffraction. However, it does not enable a comprehensive characterization of the strain field in the surface layer loaded with foreign elements for polycrystals. Also, the mechanical models adopted to interpret experiments are usually either simplified (eg. isotropic model) or apply to simplified situations (eg. textured materials) which fails to highlight the more general case in which grain orientation has a major contribution. As a consequence both extensive characterization and accurate modeling of the mechanical state of the implanted layer are required. In this communication, the selected characterization technique (micro-XRD in Laue mode) is first shown to be an efficient method to obtain the strain tensor in the implanted layer at several points within each grain of the polycrystalline samples. Then the strain tensor is demonstrated to be strongly dependent upon crystal orientation. Finally an anisotropic elastic mechanical model involving a free swelling is used to rationalize all the experimental data.

Conventional sintering of UO2 requires high sintering temperature at around 1700oC in a hydrogen atmosphere for several hours. However, this method requires long production cycle and consumes large amount of energy. Spark plasma sintering (SPS), also known as field assisted sintering technique (FAST), has now become a popular sintering method in various fields due to its merits including rapid processing and low energy consumption. However, very few reports are available in literature on sintering UO2 by SPS. Thus, in this manuscript, a systematic investigation of densification behavior and microstructure of UO2 sintered by SPS is presented. Urania powder of particle size 2.3 microns and grain size 100 nm was used in a graphite die in the SPS machine. Three major sintering parameters (maximum sintering temperature, heating rate and hold time) were varied to investigate their effect on densification, microstructure evolution, Vickers microhardness and Young&’s modulus. The result revealed that the major densification range for UO2 is between 720oC to 1000oC. The sintered material achieved a 96% theoretical density at a maximum sintering temperature above 1050 oC with heating rate of 200oC/min and a hold time of 30 seconds. The entire sintering run duration was around 10 minutes. The effect of hold time and heating rate on final density is different depending on the range of the maximum sintering temperature. At low-temperatures around 850oC, increasing the hold time to 20 minutes and the lowing heating rate to 50 oC increase the final density more significantly than sintering at high temperature range above 1350 oC. The average grain size is increased with the maximum sintering temperature and the hold time. There was no significant effect of the heating rate on the average grain size. Typical grain size varied from 2 microns to 6 microns depending on the processing conditions. The mechanical properties of the SPS sintered pellets have been measured and compared with the value reported in the literature. The average hardness of the pellets is around 6.4±0.4GPa and shows a Hall-Petch correlation with the average grain size. The measured Young&’s modulus using ultrasonic measurements is 204±18GPa and it increased with the theoretical density. Both of these results are in agreement with literature for UO2 produced by conventional methods.

Nanocomposite materials provide many opportunities for synthesizing materials with improved or unique properties. The challenge for exploiting this potential resides in the difficulty in elaborating such materials at the nanometric scale. An attractive way to overcome this difficulty is to employ the self organization of the composition field generated by ion beam mixing at the atomic scale. This composition field is mainly driven by thermal spikes induced by the slowing down of incident particles. Despite some attempts were made to describe the effect of a thermal spike on the composition field, no clear analysis of the effect of a displacement cascade on this field has been established. Using a simple Cahn Hiliard model, we present a first attempt to describe this effect. In this presentation, we focus our attention on the concept of “effective temperature” to describe the ion beam mixing.

In nuclear energy systems, structural and fuel materials suffer from high energy neutron bombardment, which causes material&’s microstructure changes, resulting in the degradation of materials&’ performance. This may lead to serious problems in the integrity of reactor performance. Not only for development of superior radiation-resistant materials but also for establishment of the methodology of efficient and effective reactor maintenance, the degradation of material&’s performance due to irradiation should be well understood, enough monitored, precisely predicted, sufficiently controlled, and reasonably suppressed, on a basis of reliable radiation damage physics. In the present study, multiscale radiation damage processes were theoretically investigated using such several computational modeling methods as ab-initio calculations, molecular dynamics (MD) calculations, kinetic Monte-Carlo (KMC) calculations, and kinetic rate theory based calculations. Nucleation and growth of point defect clusters in nuclear metals during irradiation was simulated using KMC and rate equation calculations, in which defect energetics was obtained by ab-initio and MD calculations. Simulated microstructural evolution in irradiated metals was obtained as a function of dpa/s. It shows that the nucleation and growth of voids depends much on dpa/s, while that of SIA loops does not. Also, in order to understand the oxidation process of Zr alloy, the fuel cladding material in commercial light-water reactor, the migration behavior of an oxygen atom in ZrO2 was investigated by ab-initio calculations and diffusion equation analysis as a function of the stress and temperature gradient applied to ZrO2. Our results indicate that the oxidation rate of Zr is in approximately proportion to cubic root of time, which is consistent with experiments.

The grain size in nuclear fuel has a large effect on critical phenomena such as fission gas release and creep. Thus, our ability to predict these important phenomena will be improved by a better understanding of grain growth under irradiation. In this study, a phase field model is used to study the interaction between pore and grain boundary migration under a high temperature gradient. We begin with a detailed study of the interaction between a single pore and a bicrystal GB. We then do a more complicated simulation of many pores migrating in a polycrystal. We also discuss how this information will be used to develop a model of the average grain size as a function of temperature for a fuel performance code.

Atomic collisions in solids induced by ion beam are often associated with the concept of disorder. In fact, the mobility induced in solids by ion irradiation at appropriate temperatures leads to the production of a wealth of phases which may (or not) be related to the equilibrium phase diagram. Despite, many attempts are made to understand the phase stability and the enhanced mobility of defect under irradiation even at the atomic scale, no clear picture of ion beam mixing exists. The major problem associated with ion beam mixing comes from the fact that it remains quite difficult to accurately measure a concentration over few nanometers. The X Ray Reflectometry (XRR), extensively used in micro electronics, appears to be a useful technique to overcome this difficulty. In this work, we apply the XRR technique to study the nanostructuration of Cr/Si layers induced by 80 keV Kr ion beam at room temperature, a textbook example of ion beam mixing. The analysis of XRR profiles allows computing accurate profiles of Si and Cr concentrations with a resolution equals to 1 nanometer. From these experimental profiles, we point out that the ion beam mixing appears to be a complex process which can not be only described as a diffusion controlled process.

This presentation will review some of the current and emerging strategies to develop high-performance materials with simultaneous high radiation resistance, high strength, good toughness and corrosion resistance, and moderate fabrication cost. There are three general approaches for designing radiation resistance: Nanoscale precipitates or interfaces to produce high point defect sink strength; purposeful utilization of immobile vacancies; and utilization of radiation-resilient matrix phases. In the future, utilization of advanced manufacturing processes to produce near-net shape parts with precise microstructural control will be of increasing importance to control fabrication costs and to create high-performance fabrication architectures that could not be achieved using conventional fabrication methods. Recent progress on development of high-performance steels designed using computational thermodynamics will be summarized. These steels are designed to incorporate a high density of highly stable nanoscale precipitates that could serve as efficient point defect recombination centers during irradiation, and also provide good thermal creep strength at high temperatures. Some of the options under consideration for potential accident tolerant cladding and fuel systems for commercial fission light water reactors will be discussed. Desirable attributes for the cladding under postulated loss of coolant accident and station black out conditions include oxidation resistance to steam and air and good thermal creep strength at temperatures in excess of 1200oC, along with standard cladding requirements for fabricability, low parasitic neutron absorption, hermetic containment of fission products, good compatibility with the fuel and fission products, and acceptable cost.

Internal structure of pressurized water reactors are made of austenitic materials. Under irradiation, the microstructure of these concentrated alloys evolves and solute segregation on grain boundaries or irradiation defect such as dislocation loops are observed to form. In order to model and predict the microstructure evolution, a multiscale modeling approach needs to be developed, which starts at the atomic scale. Atomic Kinetic Monte Carlo (AKMC) modeling is the method we chose to provide an insight on defect mediated diffusion under irradiation. In that approach, we model the concentrated commercial steel as a FeCrNi alloy (γ-Fe70Cr20Ni10). As no reliable empirical potential exists at the moment to reproduce faithfully the phase diagram and the interactions of the elements and point defects, we have adjusted a pair interaction model on DFT calculations. The point defect properties in the Fe70Cr20Ni10, and more precisely, how their formation energy depends on the local environment will be presented and some AKMC results on irradiation defect formation and solute segregation will be presented.

The overlap of displacement cascade is believed to be important in the development of visual defect clusters in thin film, in-situ ion irradiation studies. In this work, we use molecular dynamics simulations to investigate how impurities and damage induced by displacement cascades affect the mobility of a pre-existing interstitial-type dislocation loop in BCC iron. It is well known that impurities, such as helium, carbon, and nitrogen affect the ability of interstitial dislocation loops, and are likely responsible for difference in loop diffusivities between computer simulations and experimental observations by transmission electron microscopy. We have used molecular dynamics simulations to evaluate whether a displacement cascade could result in the de-trapping of an interstitial cluster from interstitial impurity atoms. By varying the energy and directional velocity of the primary knock on atom (PKA), we are able to observe how the trapped defect reacts with the cascade damage. Our initial simulation results reveal that cascades caused PKAs of energy greater than 10 KeV can cause the loop to de-trap from impurities, but that it may rapidly become trapped in the cascade debris. Furthermore, on several occasions, the cascade has induced a change in orientation or Burgers vector of the dislocation loop. The presentation will summarize the molecular dynamics simulation results as a function of PKA energy, distance from the trapped loop and direction, as well as the effect of loop size on the probability for de-trapping and subsequent diffusion. These simulation results will be used to inform cluster dynamics models of dislocation loop evolution in irradiated ferritic/martensitic alloys.

Ferritic Stainless steels are commonly used as structure materials in the nuclear industry. The prediction of the lifetime of current power plans and the development of new generations of nuclear reactors raise the question of the possible decomposition of iron-chromium alloys. Atomistic simulations are especially useful to study the kinetics of decomposition during thermal ageing and under irradiation. We propose here a modeling of precipitation in binary iron-chromium alloys during thermal ageing by Atomistic Kinetic Monte Carlo simulations in a rigid lattice approximation. The system evolves by vacancy diffusion. Pair interaction energies are fitted on the thermodynamic and diffusion properties of the alloy. Both of these properties are highly influenced by the magnetic evolution of the alloy with the concentration of chromium and the temperature. On the one hand, the mixing energy of the alloy faces a change of sign according to the concentration which lead to an asymmetrical phase diagram. To take this characteristic into account, a dependency on the local chromium concentration is introduced in interaction energies. By this fit, we obtain solubility limits in good agreement with most recent review studies. On the other hand, the ferro to paramagnetic transition strongly accelerates the diffusion near the Curie temperature. As this evolution affects the kinetic evolution of the alloy, we reproduce this increase of the diffusion coefficients by introducing a correcting parameter on the migration barriers depending on the local concentration and the temperature. In order to evaluate our model, we compare the Monte Carlo simulations with existing Small Angle Neutron Scattering kinetic experiments at various compositions and temperatures. Some of these experiments are in the ferromagnetic configuration, others in the paramagnetic one. We obtain a good agreement between experiments and simulations for both magnetic configurations.

Void swelling and irradiation creep can be life-limiting processes for components of both fast reactors and light-water reactors. Predictive equations are therefore required to forecast safety and economic limits of operation. It is not generally recognized, however, that such equations are almost always developed from rather limited numbers of specimens, all of thickness 0.3-1mm. In such configuration there are no significant variations in temperature, stress or neutron flux-spectra across the specimen thickness. However, in actual reactor components, especially in PWR internals or fast reactor reflector assemblies, thicknesses can be larger with significant gradients not only in environmental variables (temperature, dpa rate, stress) but in the resulting distribution of macroscopic stresses and strains. There are currently no benchmark data fields that allow confident incorporation of such "thin" equations in design codes for "thick" and complex shapes, especially since swelling-creep interactions determine the local stress field and feed back into both swelling and creep strains. We have examined a series of five hexagonal cross-section reflector blocks (annealed 304SS, 50mm flat-to-flat, ~250mm length) that were vertically stacked in a thin-wall (1mm) hexagonal 304SS wrapper can in Row 8 of EBR-II in flowing sodium. During their residence in core the blocks accumulated 0.5-33 dpa depending on their axial position in the assembly. Over the stack there were significant axial and radial gradients in both dose and temperature with gamma heating leading to significant internal temperature increases, producing a complex spatial distribution of void swelling and creep strains. Four of these blocks have been subjected to non-destructive examination, and two of these to extensive destructive examination thereafter. Measurements involved profilometry, ultrasonic time-of-flight measurements to map the internal distribution of swelling, backed up by density change measurements and electron microscopy. Asymmetrical internal peaks approaching ~4% swelling were found in the center of the blocks with average swelling of ~1.5% over the block cross section. It was found that the complex internal distribution of microscopic strains arising from void swelling and carbide densification can be related to the macroscopic deformation of the blocks, producing both bulging and bowing of the blocks. The strains of the blocks could also be correlated with the measured strains of the hexagonal can that housed them. These data not only provide unique insights on the interrelationships between swelling and irradiation creep but can also serve as a benchmark for computer code calibration for prediction of distortion of thick components.

The influence of temperature, hold time, and cooling rate on the Charpy impact properties and the copper level in the matrix has been investigated on a weld fabricated from the same weld wire used for HSSI Weld 73W and a Linde 80 flux before and after irradiation to 0.8x10¹⁹ neutron/cm² (E>1MeV). This weld has a relatively high bulk copper content, 0.32% wt, in as-welded condition. The heat treatment consisted of heating the material to the desired temperature, holding at the post-weld heat treatment (PWHT) temperature, and then cooling down to room temperature. Except for special cases, all PWHTs were performed with a heating and cooling rate of 15^oF/h (8^oC/h) to simulate the heating/cooling rate of a real vessel. In two special cases, material was heated with 15^oF/h (8^oC/h) rate but water quenched after holding at the PWHT temperature. The highest PWHT temperature was 650^oC/24h, while the other PWHTs were 610^oC/24h, typical PWHT of reactor pressure vessels (RPV), and 580^oC/100h. In addition, a part of the material after 610^oC/24h PWHT was heat treated at 454^oC/168h to simulate a post-irradiation annealing. Charpy impact properties were measured using sub-size 3x4 mm specimens and matrix Cu content was measured by atom probe tomography before and after irradiation. Small-angle neutron scattering was used to measure number densities and size of copper-rich precipitates in the irradiated specimens. Charpy specimens were irradiated in the Ford Reactor at 288^oC. It was found that the higher PWHT temperature resulted in higher Charpy upper-shelf energy (USE) with little effect on the ductile-to-brittle transition temperature (DBTT). The lower PWHT temperature and slower cooling rate were found to be beneficial in reducing the matrix Cu content. The matrix Cu content after irradiation to 0.8x10¹⁹ neutron/cm² was approximately the same for all three welds measured regardless of their different matrix Cu contents in the unirradiated condition. Consequently, the weld with the lowest PWHT temperature (and lowest matrix Cu) exhibited the lowest shift of DBTT and drop in USE. These results are proving the postulate about impotence of copper in solution (matrix Cu) rather than bulk Cu content in radiation embrittlement of RPV materials. Additional annealing at 454^oC/168h after 610^oC/24h PWHT did not show any additional effects on subsequent radiation embrittlement. Atom probe tomography (MKM) sponsored by the Scientific User Facilities Division (ORNL ShaRE User Facility), Office of Basic Energy Sciences, U.S. Department of Energy.

To estimate irradiation effects such as irradiation hardening or softening both in fission and fusion reactor components, an appropriate equation describing relationship between microstructure and mechanical properties in actual length scale is absolutely required. On the other hand, austenitic stainless steels have been used extensively in fission reactor applications, so that the most extensive database for nuclear applications would be based on the steels. In this study, modified-SUS316 austenitic stainless steels irradiated in a fast reactor were investigated to construct an equation connecting their micro- and macro-structures and mechanical properties with utilizing TEM, tensile test and hardness test. A modified-SUS316 stainless steel (PNC316), which has been developed as a cladding tube material with superior high temperature strength and swelling resistance for a liquid metal cooled fast reactor, was examined in this study. The representative chemical composition of PNC316 is Fe-16Cr-14Ni-2.5Mo-0.25P-0.004B-0.1Ti-0.1Nb. It is, in most cases, used in the 20% cold-worked condition. Neutron irradiation was performed using the experimental fast reactor JOYO. The range of irradiation temperature and dose were 750-1000 K and 16-103 dpa corresponding to irradiation time from 4,000 to 20,000 h, respectively. Correlations of yield strength and Vickers-hardness with irradiation-induced microstructure were investigated. Irradiation softening in yield strength occurred in whole temperature range in the present study, probably due to a recovery of existing dislocation structure. Large precipitates like Laves phase, M₆C and/or M₂₃C₆ were observed by TEM, especially in high temperature irradiation condition. The estimation of radiation-induced change in yield strength, Δσ_y, based on barrier hardening model developed by short-range and long-range obstacles showed small value compared to actual Δσ_y. It is assumed that concentration of carbon in the matrix was decreased due to precipitation of irradiation-induced large carbides, resulting in a loss of solution hardening.

Diffraction is well suited to the characterization of microstructures. Neutron and high energy x-ray diffraction, in particular, have some undeniable advantages for studying activated samples. Neutrons and high energy x-rays penetrate millimeters into most materials, providing a statistically relevant probe of the microstructure in the bulk of the material. Moreover, little or no hazardous and costly sample preparation is necessary, which enables repeat tests on the same sample or even in-situ measurements under simulated operating conditions. Finally, neutron detectors are often insensitive to background gamma radiation emitted from activated samples. The SMARTS diffractometer at the Lujan Center was designed to study engineering materials under their operating conditions and, as such, has sophisticated sample environments enabling in-situ ND studies during deformation and at non-ambient temperatures. This talk will use the example of in-situ diffraction measurements during annealing and deformation of irradiated HT-9 steel to highlight the capabilities of the instrument, in particular, related to the study of activated materials.

Structural materials used in nuclear environments are subjected to significant doses of radiation at elevated temperatures. This gives rise to a large number of point defect clusters, dislocation loops, and complex dislocation networks in the irradiated metals. Irradiation-induced defects hinder the glide of mobile dislocations (that carry inelastic deformation) through the material. An increase in yield strength, followed by flow localization and lower engineering strain to fracture is observed during quasi-static tensile loading. The post-deformation microstructure reveals that the inelastic strain is localized along narrow dislocation channels which are ‘cleared&’ of majority of the irradiation-induced defects. The present work uses a continuum constitutive crystal plasticity framework to model the mechanical behavior and deformed microstructure in irradiated bcc metals, with emphasis on capturing the aforementioned localization phenomena. Material defects generated due to irradiation (interstitial loops in bcc) and dislocations (mobile and immobile) are used as substructure variables in this model. The substructure evolution equations are based on physical mechanisms of dislocation-dislocation and dislocation-defect interactions. The framework is used to simulate the localization phenomena for a model ferritic/martensitic steel subjected to irradiation and then loaded in tension. Events leading to flow localization along the dislocation channels are studied. Qualitative effects of varying the grain size, initial crystallographic orientations, and degree of cross-slip on the localization behavior are also studied.

Synergistic effects of H and He on radiation damage of Fe-16Cr-4.5Al-0.3Ti-2W-0.37Y2O3 ODS ferritic steel ODS steel under triple-ion irradiation have been investigated using high-resolution transmission electron microscopy (HRTEM) techniques. The frequent observations of partially crystallized Y4Al2O9, YAlO3, and Y2TiO5 complex-oxide nanoparticles/clusters indicate that the nanoparticles/clusters in as-fabricated MA/ODS steel are not only structurally defective but also chemically deviate from equilibrium. Simultaneous triple-ion-beam consisting of Fe8+, H+, and He+ were employed for irradiation of the ODS steel at 600C. Results of the experiment reveal that a constructive effect of defective nanoparticles/clusters on the suppression of radiation-induced swelling can be revealed through the observations of helium-filled cavities (bubbles) trapped preferentially at the interfaces between the matrix and nanoparticles/clusters. An adverse effect of hydrogen implementation on the triple-ion irradiated ODS ferritic steel is readily observed through the formation of hydroxide compound in association with large facetted voids. The formation of HFe5O8-base hydroxide compound (space group: P63mc) with lattice parameter: a = 0.598 nm and c = 0.937 nm was identified presumably for the first time in triple-beam irradiated ODS ferritic steel. This work was performed under the auspices of the U.S. Department of Energy by Lawrence Livermore National Laboratory under Contract DE-AC52-07NA27344.

Generation of helium in materials through transmutation nuclear reactions under high-energy neutron irradiation, giving rise to radiation swelling and grain boundary embrittlement, is a major factor limiting the lifetime of structural materials in fusion and fission power plants. Chemically, helium is similar to other noble gases, for example neon and argon, and the development of an accurate predictive model for defects formed due to accumulation of helium and other noble gases in the crystal lattice of iron, steels and non-magnetic body-centred cubic (bcc) refractory metals and alloys, offers a way of quantifying the effect of transmutation reactions on the structural integrity of reactor components. So far, experimental and modelling effort has been focused exclusively on helium defects and the combined synergetic effects associated with simultaneous accumulation of helium and hydrogen in materials. It remains unclear to what extent ion-irradiation experiments, involving other noble gas atoms, could be used to explore the radiation-induced phenomena resulting from accumulation of transmutation products in materials. Reliable experimental information on noble gas defects is scarce, and application of first-principles density functional methods is probably the fastest and most accurate presently available method of acquiring quantitative information about noble gas defects. We have carried out a comparative study of defects, resulting from the incorporation of noble-gas atoms (He, Ne, Ar, Kr, Xe) into bcc transition metals, using first-principles density functional theory (DFT) calculations. The formation energies for noble-gas atoms in various substitutional and interstitial configurations have been calculated to understand the trends and quantify the local lattice distortion effects as a function of the noble-gas atom size. Helium is a relatively small atom and He defect energies are lower than those corresponding to other noble gas atoms. The size effect results in the change of the relative stability of tetrahedral and octahedral interstitial sites for Ne, Ar, Kr and Xe in comparison with He. Furthermore, the most stable double-gas-atom configuration changes from the <110> in bcc-Fe and <111> in bcc-W configurations for the He-He case into the <100> one for other noble-gas atoms. We also investigate the stability of bound vacancy-noble-gas configurations which are expected to promote void nucleation under irradiation. Finally, the dependence of vacancy-noble gas binding energies in iron and other bcc transition metals has been investigated as a function of vacancy/noble-gas atom ratio to provide input for kinetic Monte-Carlo simulations of microstructural evolution under irradiation.

In addition to intrinsic defects, large amounts of He and H are also produced by nuclear transmutation under high energy neutron irradiation. For instance, structural materials of future fusion devices may suffer from swelling, intra- and inter-granular embrittlement due to the accumulation of He. As a first and indispensable step to understand these macroscopic properties, the energetic and mobility of He atoms in bcc iron need to be accurately determined. Both first-principles and classical molecular dynamics methods are employed to investigate the lowest-energy sites and migration mechanisms of He in various grain boundaries (GBs) with different characteristics. We have first optimized the GB structure by adding either vacancies or self-interstitial atoms close to the interfacial plane. Then, He formation energies for all the possible sites are evaluated. The obtained values are indeed much lower (~1.7 eV) near the GBs than in the bulk, indicating a strong He segregation tendency. In addition, isolated He atoms reveal to reduce significantly the GB cohesion, although without forming bubbles. We show that the He formation energies are determined by the interplay of two competing contributions: the available volume for He insertion, and the distortion of the Fe lattice. Also, both 0K and finite temperature migration barriers for He and vacancy diffusion in the various GBs are calculated. Interestingly, an interstitial He always requires a larger energy barrier for diffusion along the GBs than in the bulk, at variance with the case of vacancy migration. The present results are used as input data for larger-scale models in order to study the kinetic of He segregation and bubble formation at GBs.

The structure of He bubbles in bcc Fe is discussed from a simulation point of view. Formation energies are calculated for different sizes of vacancy-He clusters using classical interaction potentials. Molecular dynamics simulations are then performed to investigate the effect of low energy collision cascades on bubbles of different sizes. The results show that it is possible for both vacancies and Fe interstitials to be attracted to the bubbles depending on their relative size and stability. Some preliminary long time scale dynamics results will also be presented to indicate possible mechanisms by which the He bubbles can form.

Nanomechanical experiments and site-specific microstructural analysis on as-fabricated and helium (He)-implanted 120nm-diameter single crystalline Cu and Fe nano-pillars revealed specific effects of He implantation on the mechanical properties. In particular, the <111>-oriented Cu pillars implanted with 0.35±0.05 at. % He throughout the gauge section were found to yield at 1.2GPa under uniaxial compression. This value is 30% higher than the yield strength of as-fabricated samples with the same diameter. Stress-strain data of the implanted Cu pillars exhibits shorter and more frequent strain bursts, as well as notable strain hardening with a hardening slope of 3.52±0.82GPa. In the case of <110>-oriented Fe pillars, samples implanted with 0.33±0.05 at. % He gave a compressive flow stress of 2.0GPa at 5% strain, a value ~25% higher than the 1.5GPa flow stress of as-fabricated pillars. In both as-fabricated and implanted Fe pillars, the stress-strain behavior was shown to have 3 distinct regimes, starting from elastic loading followed by continuous strain hardening, and eventually reaching a “steady state” where the flow stress remained a constant. Our findings were interpreted in terms of the interaction between mobile dislocations and the implantation-induced defects.

Using atomistic modeling, we find that He is trapped at Cu-Nb interfaces in the form of sub-nanometer sized platelet-shaped clusters at misfit dislocation intersections (MDIs). This behavior occurs due to the spatial heterogeneity of Cu-Nb interface energy: He-vacancy clusters wet regions of high interface energy at MDIs while avoiding low interface energy regions. Below a critical interface He concentration, these platelets are stable even in the presence of high vacancy supersaturation. The consequences of this finding for preventing He damage in fusion applications will be discussed. This material is based upon work supported as part of the Center for Materials at Irradiation and Mechanical Extremes, an Energy Frontier Research Center funded by the U.S. Department of Energy, Office of Science, Office of Basic Energy Sciences under Award Number 2008LANL1026.

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, in particular of tritium, is a serious issue. 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. Using molecular dynamics simulations, we have studied the effect of voids and helium bubbles on hydrogen binding to bubbles and defects, hydrogen retention and hydrogen diffusion. Hydrogen introduced in bubbles of different size and composition quickly diffuse towards the edge of the bubbles, but stay bound around the first atomic layer of the surrounding W matrix. Only at high temperatures, 1500 K and above, a significant amount of hydrogen escapes further than a few Å from the bubble. Helium, on the other hand, stays strongly bound within the bubble. By comparing the energetics and positions of hydrogen defects in the tungsten matrix, near the edge of the bubble and inside the bubble, we note that the edge is about 1 eV more favorable than the ground state position in the matrix. A bubble near the surface can burst, expelling the gas atoms, depending on bubble pressure and distance to surface. As the hydrogen is preferentially bound to the edge of the bubble, a significant amount of the hydrogen is retained in the surface, while the helium is expelled and a crater is formed. We study the effect this has on gas atom composition in the surface and surface morphology. By implanting H and He in surfaces with pre-existing high pressure He-H bubbles close to the surface, we can study both surface roughening and the different roles of hydrogen and helium during plasma exposure.

Helium atoms formed in steels by transmutation reactions under neutron irradiation lead to He-filled bubble formation that affect strongly microstructure evolution and mechanical properties. We present here the results of atomic-scale modeling study of properties of He-filled bubbles in Fe. We have investigated equilibrium state as a function of bubble size, ambient temperature and He content as well as effect of He-bubbles to dislocation motion. The results obtained on bubble stability allow us to formulate an equation of state for He-bubbles that can used in theoretical and computational models for prediction of microstructure evolution in steels under neutron irradiation. The results on bubble-dislocation interaction allow the prediction of mechanical property change in irradiated steels.

Helium-3 buildup is an important mechanism of aging in metal tritides. Accumulation eventually leads to formation of bubbles of 3He gas, causing changes in material properties and, ultimately, uncontrolled release. It has been hypothesized that this problem would be mitigated in nanoporous Pd, since all points in the metal lattice should be near a gas/solid interface, allowing 3He to diffuse harmlessly out of the solid. We have shown that mesoporous Pd and Pd alloy powders can be synthesized in a scalable fashion using soft templates. We have produced batches of mesoporous Pd on the gram scale, with regular arrays of pores having diameters that are tunable between 3 and 13 nm by chemical reduction of Pd2+ salts around hexagonally packed cylindrical micelles of surfactants or block copolymers. This control of pore size is effected by varying the composition and/or size of the molecular template. The pore size has effects hydrogen storage capacity and kinetics, as well as thermal stability. Preliminary results of helium implantation experiments show that helium is not retained in these materials. Sandia National Laboratories is a multi-program laboratory managed and operated by Sandia Corporation, a wholly owned subsidiary of Lockheed Martin Corporation, for the U.S. Department of Energy&’s National Nuclear Security Administration under contract DE-AC04-94AL85000.

During the initial stages of the deceleration of a fast atom in a bulk material, interactions of the moving atom with the electronic system of the target material are the dominant mechanism. For several decades the interest in understanding these processes has been growing and has prompted both experimental and theoretical studies, ultimately aiming at producing novel materials with a high radiation tolerance. In recent years predictions from parameter-free computational studies are becoming a useful tool in this context: Predictive results can now be achieved due to the use of real-time time-dependent densityfunctional theory for modeling the interaction of the projectile with the electronic system of the target, overcoming the problematic adiabatic Born-Oppenheimer approximation. Our newly developed first-principles implementation of Ehrenfest dynamics is based on explicit integration of the time-dependent Kohn-Sham equations in real time. The wave functions are expanded in plane-waves and we achieve an excellent scalability on high-performance computers. Being able to perform large-scale simulations involving several hundreds of electrons we systematically study the stopping of fast H and He atoms in bulk Al in order to address various scientific and computational issues. These include the influence of core electrons on the stopping and also how different charge states (e.g. He vs. He⁺⁺) of the projectiles can play an important role. Prepared by LLNL under Contract DE-AC52-07NA27344.

Reduced-activation ferritic/martensitic steel F82H has been developed as one of prime candidate materials for experimental fusion reactors. Some of the key issues are the effects of helium and hydrogen on the microstructure evolutions such as swelling, and on the mechanical properties such as fracture toughness or embrittlement. In this study, iron-base alloys and its electron-beam-welded joint have been irradiated by using an ion accelerator and a High Voltage Electron Microscope (HVEM) to examine synergistic effect of displacement damage and hydrogen or helium atoms on microstructure evolution, especially swelling behavior. Pre-injection of hydrogen (0~2000appm) and/or helium (0~1000appm) at RT, followed by electron irradiation in matrix at 400oC, were carried out for pure iron, F82H IEA and electron-beam-welded F82H IEA. Irradiation-induced cavities were observed in all the alloys. He-implanted alloys tended to exhibit higher number density and smaller mean size of the cavities compared to H-implanted alloys. In addition, in-situ electron irradiation experiment for H-implanted pure iron showed change in morphology and shrinkage of cavities. From these results, it is suggested that He effect on enhancement of cavity formation would be greater than H, while H could affect stability of cavities during irradiation. Multi-beam ion irradiation experiment with Fe+, He+, and H+ were also carried out at 400oC up to 10 dpa. Dual beam irradiation with Fe+ and H+ in pure iron resulted in lower number density and larger mean size of cavities compared to H-implanted and electron irradiation condition. Additionally, F82H and EB-welded F82H dual-beam -irradiated with Fe+ and H+ exhibited little formation of cavities. Therefore, it seemed that H effect on cavity formation in F82H would be much less than that in pure iron.

In order to assess the potential for cavity swelling at end-of-life doses in LWR internal components fabricated from austentic stainless, it is necessary to develop a validated computational model that incorporates the relevant physical mechanisms. The basis for the current modeling activity is a comprehensive microstructural model that was developed to assess the swelling behavior of similar materials under fast neutron irradiation at elevated temperatures. The model employs the well-known reaction rate theory description of radiation damage formation and damage evolution, and accounts for the role of helium produced by nuclear transmutation in the nucleation and growth of voids. The model has been updated to account for recent advances in our understanding of primary damage production and point defect diffusion and reaction behavior obtained from atomistic simulations. Model validation has focused on the 250 to 325°C temperature range. The presentation will focus on the influence of critical irradiation and material parameters which would lead to the prediction of a level of swelling that is of engineering significance.

Thermally or radiation induced transport properties impact practically all engineering aspects of nuclear oxide fuels, whether at the manufacturing stage, during in-reactor operation, or under long-term repository conditions. From a more fundamental standpoint, measuring transport properties is also a means of probing point or complex defects that are responsible for atomic migration. Although many studies relating to self-diffusion in UO2 have been carried out over the past forty years, these have not generally focussed on characterising these properties as a function of all the physical variables which determine it, i.e. temperature, the oxygen partial pressure and the impurity content, usually present in the form of bi- or tri-valent action impurities. In this talk, we show how electrical conductivity and self-diffusion property measurements may be combined in order to determine fundamental data relating to the nature of defects responsible for the property and their formation or migration energies. These data may then be compared to those obtained from first principles electronic structure calculations. The first part of the talk is dedicated to the development of a point defect model which captures the basic dependence of the electrical conductivity of UO2 upon temperature and oxygen partial pressure. The formation energies derived from this exercise are then compared to charged defect calculations using Density Functional Theory in the LDA+U approximation. The second part is concerned with oxygen self-diffusion and chemical diffusion coefficient measurements. We show that the point defect model is also compatible with the experimental data available. In the third part of the talk we examine uranium self-diffusion properties. The point defect model is specifically developed to account for uranium defects in a composition range close to stoichiometry. An analytical apparent activation energy is derived in this composition region and a numerical application is carried out based on basic formation and migration energy estimates obtained from first principles. The results compare favourably to existing data but highlight the need for additional experimental and theoretical work.

Within the Fuel Cycle Research & Development program, six mixed oxide fuel compositions were fabricated and irradiated in the Advanced Test Reactor at the Idaho National Laboratory. Irradiation to 262 effective full power days resulted in fuel burnups ranging from approximately 5.8-8.4 at.% heavy metal. Fuel compositions included standard mixed oxide (MOX) fuel, (U_0.80,Pu_0.20)O₂, and MOX fuel with minor actinides, (U_0.75,Pu_0.20,Am_0.03,Np_0.02)O₂. These compositions are being studied in order to understand how fuel incorporating constituents from recycled light water reactor fuel would behave in fast reactors, with particular focus on transmuting actinides. Initial destructive postirradiation examinations of these compositions have been completed, including fission gas release, optical microscopy, microhardness testing, and burnup analysis. Initial calculations of fission gas release indicate that the fractional release of krypton and xenon fission gases ranges between 30% and 100% (complete release). Select postirradiation examination results, with a focus on fission gas release, will be presented and compared to results from the historical fuel performance database of fast reactor oxide fuels.

Oxygen interstitials in UO_2+x significantly affect thermophysical properties of the oxide nuclear fuel. Based on our previous work, we have analyzed the effect of mono-, di-, and other larger interstitial clusters on the oxygen ion diffusivity. First principles calculations are used to estimate the stability and energetic of different interstitial clusters. The estimated energies (migration, dissociation, etc.) with first principles are used to estimate the oxygen ion diffusivity in the kinetic Monte Carlo. We will discuss the effect of temperature and stoichiometry on the overall diffusivity of UO_2+x. The computed diffusivities are compared with available experimental data. In addition we will present the sensitivity analysis of the associated energies on the diffusivity of oxygen interstitials. This work is funded by the DOE Office of Nuclear Energy&’s Nuclear Energy University Programs.

Xenon (Xe) is one of the major gas elements produced during nuclear fuel burning. It has long been known that Xe accumulation in the fuel pellet and Xe release to the plenum are detrimental to the fuel performance and safety and therefore must be well understood and controlled. Experimental data concerning description of Xe diffusion in UO2 -based nuclear fuels have a wide range of disparity such that it is difficult to verify modeling results for development of predictive nuclear fuel performance codes. Obtaining conventional UO2 samples containing even a few percent Xe accumulated in-pile is not possible without years of irradiation in a test reactor. An equally challenging problem, which relates to obtaining quantitative parameters governing Xe migration, is synthesis of reference UO2 samples with a uniform and controllable Xe concentration, separated from the various and complex other effects attendant to the fission process. The goal of this work was to fabricate and characterize such reference samples to isolate and quantify the inherent transport properties of Xe in UO2. We utilized ion beam assisted deposition (IBAD) to fabricate depleted UO2 (DUO2) films with embedded Xe atoms. The films were annealed at 1000 oC to induce Xe atom redistribution, and characterized before and after the annealing. Microstructural and chemical composition changes were examined by transmission electron microscopy (TEM), energy dispersive X-ray spectroscopy (EDXS), atomic force microscopy (AFM) and Rutherford backscattering spectroscopy (RBS). A detailed study of the deposition condition&’s influence on Xe content and morphology as well as the DUO2 film microstructure will be presented. Preliminary results detailing Xe-filled bubble growth and Xe atom diffusion in the DUO2 films will be presented.

The interaction of ions with solids results in energy loss to both atomic nuclei and electrons. At low energies, nuclear energy loss dominates, and irradiation damage occurs primarily by ballistic collisions. At high energies for fission products and swift heavy ions, electronic energy loss dominates, and the intense ionization can lead to latent track formation or recovery of existing irradiation damage. At intermediate ion energies, including energies of primary knock-on created by fast and fusion neutrons, ballistic and ionization energy losses are of similar magnitude and can lead to synergistic or competitive processes that that affect the evolution of irradiation damage. We have integrated experimental and computational approaches to investigate the separate and combined effects of nuclear and electronic energy loss on damage formation and recovery in several materials. Experimentally, we have shown that that there is a synergy between the nuclear and electronic energy loss on damage evolution in amorphous SiO₂ at intermediate ion energies. Large scale molecular dynamics simulations, which include ballistic collisions and local heating based on the inelastic thermal spike model, have been employed to investigate the separate and combined effects of nuclear and electronic energy loss on damage production. These simulations demonstrate conclusively the additive effect on nuclear and electronic energy loss on damage production. On the other hand, ionization in Ca₂La₈(SiO₄)₆O₂ from intermediate energy ion irradiation leads to competitive damage recovery processes that decrease damage production. In SiC, irradiation with intermediate energy ions leads to defect formation and amorphization; however, it has been shown that swift heavy ions can induce some recovery of such irradiation damage. Large scale molecular dynamics simulations confirm that swift heavy ions induce defect recovery and recrystallization in SiC that are well described by an inelastic thermal spike phenomenon.

The CePO₄ monazite, with Ce as a surrogate for Pu, is considered as an important candidate to incorporate actinides for potential nuclear waste form applications. In this study, we synthesized nanostructured CePO₄ monazite with different grain sizes ranging from 20 nm to over 100 nm. Separate and simultaneous displacive (1 MeV Kr²⁺) and ionizing (200 keV electrons) irradiations were conducted on the different sized CePO₄ in order to investigate their radiation response. In situ TEM observation revealed that CePO₄ nanoparticles can be amorphized by 1 MeV Kr²⁺ irradiation with better radiation tolerance than their bulk counterpart, whereas 200 keV electron-beam irradiations can induce recrystallization of the pre-amorphized CePO₄. An abnormal behavior with a strong size effects on the radiation response of nano-sized CePO₄ was identified. The 20 nm sized CePO₄ exhibits lower radiation tolerance than the larger 40 nm sized ones, probably due to excess surface energy of reduced sized CePO₄ which may alter the energy difference between the amorphous and crystalline phases and consequently affect the radiation stability. A higher recrystallization rate was observed in smaller sized CePO₄ under ionizing irradiations, which may result from the higher specific surface area that can provide preferential recrystallization sites for CePO₄ grains. When irradiated with simultaneous ionizing electrons and displacive ions, CePO₄ displayed high tolerance against amorphization, implying a desired radiation performance for its nuclear waste form application. The remarkable size dependence of the displacive radiation-induced amoprhization and ionizing radiation-enhanced recrystallization processes for nanostructured CePO₄ indicates the existence of a critical grain size for optimized radiation tolerance.

Radiation-induced defect formation and accumulation due to high-density electronic excitation is one of the essential issues to understand the radiation damage processes induced by fission fragments. In the present talk, we will give an overview of our recent transmission electron microscope (TEM) studies on defect formation and accumulation in ceramic oxides irradiated with swift heavy ions. It is well known that high-density electronic excitation induced by irradiation with swift heavy ions results in formation of columnar defects called as ion tracks in oxide ceramics. High resolution TEM observation of the ion tracks formed in MgAl₂O₄ has shown that the crystalline lattice survives irradiation with 340 MeV Au ions even in the central core regions along the ion trajectories. Here the electronic stopping power was evaluated to be 34 keV/nm. The TEM contrast of ion tracks appearing in conventional bright-field (BF) images depends on excitation of Bragg reflections. Dark columnar contrast clearly appears along the ion tracks in an inclined view with g=220, whereas it becomes faint when the Bragg position is on g=440. The contrast variation suggests structural deviation from the spinel-type crystal structure inside the ion tracks. The strain field contrast stands out under the latter condition, indicating the formation of extended defect aggregations in the matrix. TEM BF-imaging with defocus and HAADF-STEM have shown reduction of atomic density, namely formation of a high concentration of vacancies at the cores of ion tracks. The number density of ion tracks increases proportionally with ion fluence in the earlier stage of accumulation but it saturates to keep a constant after reaching 10¹⁶ ions/m². This result is discussed in terms of a balance between the formation and the annealing of the ion tracks under irradiation. The annealing range to induce the recovery is evaluated to be 7-9 nm in radius. Both electron channeling X-ray analysis (HARECXS) and electron diffraction have revealed that cations tend to evacuate from the tetrahedral sites and reside preferentially on the octahedral sites after significant overlaps of radiation damages. The results on CeO₂ with the fluorite structure also will be referred in the talk.

The disposal and safe storage of nuclear waste is a significant challenge for the global community. Several of the radionuclides generated through the nuclear fuel cycle, such as ²³⁹Pu and ²³⁵U, have long half lives (24,100 years and 7x10⁸ years respectively) and careful choice of suitable immobilisation matrices is crucial to prevent any environmental contamination. Such an immobilisation material must be able to withstand prolonged heavy ion particle bombardment while maintaining structural integrity. Pyrochlore-type compounds have been proposed as suitable host matrices for this purpose, and great attention has been paid to members of the series Gd₂(Zr_xTi_2-x)O₇ (0le;xle;2). The radiation tolerance of this series increases with increasing zirconium content, and the healing process in the zirconate is expected to be faster than for the titanate as it does not undergo an amorphous transition upon radiation damage and is a fast ion conductor. We propose a new set of Buckingham potentials, specifically tailored for looking at this Gd₂(Zr_xTi_2-x)O₇ series. The accuracy and robustness of these new potentials is demonstrated by calculating and comparing values for a selection of point defects with those calculated using a selection of other published potentials and our own ab initio values. Frenkel pair defect formation energies are substantially lowered in the presence of a small amount of local cation disorder. The activation energy for oxygen vacancy migration between adjacent O_48f sites is calculated for Ti and Zr pyrochlores using an improved tangent nudged elastic band method. This energy is lower for the non-defective Ti than for the Zr pyrochlore by ~0.1 eV, consistent with the majority of the potentials tested. The effect of local cation disorder on the VO_48f → VO_48f migration energy is minimal for Gd₂Ti₂O₇, while in contrast the migration energy is lowered typically by ~43% for Gd₂Zr₂O₇. Since the healing mechanisms of these pyrochlores are likely to rely upon the availability of oxygen vacancies, the healing of a defective Zr pyrochlore is predicted to be faster than for the equivalent Ti pyrochlore.

Ceramics have been shown to have vastly different responses to radiation damage, in some cases recovery can be rapid, in others it can be slow. What are the likely contributors to the recovery rate is the subject of much research, both experimental and simulation in nature. Using model systems it has been possible to examine the contributions of both bonding/chemical composition and structure. Primarily this through the modification of either composition or structure, i.e. modifying one with minimal, or zero, change in the other. To fully appreciate the effects of radiation damage on a system, the undamaged, or equilibrium, structure itself must be understood. The information derived from such studies can be integrated into understanding further recovery mechanisms from damage. For example, why are certain structures adopted for a given composition, when others are possible, will impact on the recovery, and more importantly, the rate from damage. For example in some systems the recovery from damage has been found to be non-linear with changes in composition, whereas in others the reverse is true. In such systems it is found that the recovery from damage initially increases with a change in composition, whereas with continued change the recovery rate begins to decrease. What are the competing drivers within the system that modifies how a system responds to damage? This work uses two examples to outline how recovery from damage can be predicted by examining structures/bonding of materials prior to irradiation. It will also show one in which the process of predicting damage recovery is not only complex, but in some cases unexpected.

Mesoporous silicas are highly potential materials for applications in the nuclear field for separation, recycling or confinement of nuclear wastes. They present a porous network that may be organized (SBA, MCM) or disordered (Vycor glass). For these intended applications it is essential to know the behavior of these materials under irradiation, as it might induce modifications of the porous properties as for example closure or destruction of the micro- or meso-porosity. For this purpose different ion irradiation experiments, with different regimes of energy deposition (electronic, ballistic), were implemented on these mesostructured materials, playing with ions energy, fluence and nature (Xe, Au, He, C, Ar,hellip;). The specific case of mesoporous thin films was investigated with the adapted characterization techniques as X ray Reflectivity and cross sectional Scanning Electron Microscopy, and compared to radiation effects on bulk mesoporous materials as Vycor glass (analysed with X rays and Nitrogen adsorption measurements), to access complementary informations on surface and volume effects. After an overview of the preliminary results and taking into account the irradiation effects described in the litterature for dense and porous silica materials and the available models, the presentation will attempt to answer the question of the existence of a relationship between the rigidity of the silica network and its mesoporous structure, with the radiation damages that are induced.

In the UK, alkali borosilicate glasses are used to vitrify high level waste (HLW) produced by reprocessing of spent nuclear fuel. HLW contains fission products and minor actinides which continue to undergo radioactive decay in the wasteform for up to 1 million years. Other cations such as Fe and Zr are also present in the waste stream and require incorporation into the final wasteform. Actinides undergo α-decay events with the formation of α-particles (He nuclei) and energetic (~100KeV) daughter recoil nuclei. Energy is transferred from the energetic recoil nuclei to other atoms in the glass via elastic interactions, resulting in atomic displacements which form collision cascades. Accumulation of this ballistic damage can lead to migration of alkali ions, resulting in changes in glass network polymerisation, and changes in cation valance state. These changes can affect wasteform durability and since the wasteform acts as the final barrier against radionuclide release into the environment, it is important that the effects of α-decay on the structure of the glass are understood. In this study, heavy ion implantation (e.g. 450KeV Kr irradiation) was used to provide an analogue for the α-recoil damage and samples were irradiated at room temperature with a dose relevant to vitrified product lifetimes (0.1 - 1.0 displacements per atom (dpa)). The effects of simulated α-recoil damage were investigated by probing the speciation and valence of Fe as a network intermediate, in analogue nuclear waste glasses, as a function of glass composition. In this contribution we report on the effect of heavy ion implantation on the Fe oxidation states and co-ordination environment, in various model glasses, using X-ray absorption and Raman spectroscopies.

In the development of current and future nuclear materials, it is key to determine the functionality of a material in the environment of its intended use. Central to this is principle is the requirement to understand how a material's irradiation response is affected by temperature. Using computer simulation it is possible to capture the fundamental atomic-scale processes responsible for the accumulation of damage within a material. In this work, we systematically study the initial phase of radiation events to determine the threshold displacement energy (E_d) using extensive molecular dynamics simulations. To quantify the effect of temperature on defect production, the simulations involve rigorous sampling of the impact energy and direction of Primary Knock-on Atoms (PKAs) at a range of temperatures from 50 to 1200 K. In application to rutile TiO₂, defect production on the oxygen sublattice is found to be significantly affected by increases in temperature. This leads to a marked reduction in the number of residual defects from both PKA species across the energy range studied. In addition, a shift in the oxygen value of E_d is observed from 18 eV at 300 K to 53 eV at elevated temperatures. This is attributed to oxygen Frenkel pair recombination mechanisms that are thermally activated during the simulations. Significantly, transitions of this kind would occur well within the timescale of experimental techniques of determining E_d. These techniques report a higher value of E_d for oxygen indicating the short-lived defects present in the simulations are recombining before experimental detection takes place. This work highlights the importance of understanding the link between temperature and timescale when calculating values of E_d. If comparisons are to be made between experiment and simulation or if E_d is to be used in models of radiation damage, it is imperative to know the context in which the value of E_d was determined.

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 CERMET fuels. 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.

Sodium zirconium phosphate (NZP) type ceramics accommodate approximately 42 elements of the periodic table including most fission products derived from nuclear power plant fuel. As such, NZP-structure type ceramics have considerable potential as host materials for the immobilization of radioactive waste as well as candidate inert matrices for minor actinide burning. It is therefore important to investigate the behaviour of this material under irradiation conditions in order to verify its long-term stability. In this study strontium zirconium phosphate (an NZP-type structure ceramic) has been irradiated with gold and helium ions to simulate the consequences of alpha decay. The effects of the irradiation on the structural as well as macroscopic properties (e.g. density and hardness) are investigated using grazing-incidence X-ray diffractometry, Raman spectroscopy, scanning electron and atomic force microscopy, and nano-indentation. Irradiation by gold ions results in significant changes to the crystalline structure and hardness. After a fluence of 10¹⁵ gold ions/cm², strontium zirconium phosphate undergoes structural amorphization, a volume reduction, and an increase in hardness. These results as well as the results from He-ion irradiation are discussed with regard to the application of NZP-structure type ceramics as inert matrices for minor actinide burning or as host materials for the immobilization of radioactive waste.

Fergusonite and scheelite-structured ABO4 ternary oxides are an important class of materials owing to their technological applicability and geological significance. In spite of their growing interest as potential wasteform ceramics, only very little is known about their behaviour under irradiation in regards to other ABO4 analogues such as zircon and monazite. To this purpose, we have studied and compared the effects of ion-beam irradiation on compounds LaVO4, YNbO4 and CaWO4 by 1 MeV Kr+ ions as a function of irradiation temperature (50 - 600K). Resulting critical temperatures for amorphisation (Tc) differ slightly for LaVO4 and YNbO4 each with a Tc of 400K and 450K respectively. CaWO4 shows stronger amorphisation ‘resistance&’ and has a Tc of 200K. The susceptablity toward amorphisation and disorder in each structure is discussed in terms of their structural parameters as well as the stopping powers, displacement energies, and defect energies of the materials. The phase transitions that occur between tetragonal scheelite and monoclinic fergusonite will also be highlighted.

This study presents structural analysis of crystalline and radiation-damaged zirconolite, CaZrTi₂O₇, and pyrochlore, Gd₂Ti₂O₇, both potential actinide-accommodating nuclear waste materials, using molecular dynamics (MD) and connectivity topology analysis - a powerful method for describing both crystalline structures and their metamict or amorphous analogues, because it places no reliance on symmetry operators or periodic translation, both of which vanish upon introduction of disorder to a material. The work establishes characteristic topological differences in the connectivity of each structure and finds evidence that amorphization induced by alpha-recoil displacement cascades still retains certain short- and intermediate-range ordered configurations, particularly for Ti atoms. [TiO_x] polyhedral edge-sharing chains are observed in the metamict state in both materials, which may act to stabilize the radiation-damaged structure and prevent recovery of the initial crystalline phase. We also present an assessment of the predicted amorphizability of zirconolite based on the topological constraints imposed by its structure, finding that the varying structural rigidity of the layers in the structure is crucial to its amorphizability potential. The hexagonal tungsten bronze structure [TiO_x] layer in particular provides weak constraints that are responsible for zirconolite&’s comparative ease of amorphization.

Molecular dynamics simulations using thermal spikes have been carried out to investigate the effect of pressure on the time-dependent generation of defects under irradiation in the three common polymorphs of TiO₂: rutile, anatase and brookite. The effect of crystal structure on the tollerance to radiation damage was first investigated, highlighting the experimental observation that the rutile phase is the most tollerant to damage. The density of the phases was then varied and the same thermal spike methodology repeated with some interesting results suggesting a strong correlation between density and radiation tollerance.

In the realm of radioactive nuclear materials, a major challenge to the development of materials is the measurement of the properties for which the material is being developed. For example, the phenomenon of microstructure evolution of a nuclear fuel in reactor is well known but the details of the effects of the change on the behavior of such important issues as thermal conductivity, mechanical properties, and phase formation have not been quantified at the grain size level. There is a strong need to develop or adapt advanced instrumentation for measurements on radioactive materials. Idaho National Laboratory has an ongoing effort to develop or adapt a variety of measurement techniques to highly radioactive materials. A laser based device termed the Scanning Thermal Diffusivity Microscope, conceived and developed over the past few years, has recently been installed in a hot cell where examinations of fresh and irradiated fuel samples have begun in order to profile the thermal diffusivity of fuels and materials at 50µm spatial resolution. A second generation instrument close to implementation will soon give thermal conductivity values at 5-10 µm. The unique application of dual-beam focused ion beam (FIB) to the preparation of highly radioactive material samples has become an exceedingly useful tool for determining 3D grain orientation (EBSD), mechanical properties by nano/micro indentation or compression testing, microstructure through transmission electron microscopy, and nano-scale element distribution by atom probe tomography. This contribution will present the current state of the implementation plan of these instruments to highly radioactive fuels and materials and examples from ongoing irradiated fuels and materials studies will be given.

Under ion irradiation, all crystalline materials display some degree of dynamic annealing when defects experience evolution after the thermalization of collision cascades. The exact time and length scales of such defect relaxation processes are, however, unknown even for Si at room temperature. Here, we propose a method to measure effective diffusion lengths and relaxation times of mobile defects that dominate the formation of stable post-irradiation disorder. A defect lifetime of about 5 ms and a characteristic defect diffusion length of about 30 nm are measured for Si at room temperature, essentially independent of the average density of ballistic collision cascades. Defect relaxation appears to be dominated by a second order kinetic process. We discuss implications of these findings for the development of predictive models of radiation damage buildup in solids. This work was performed under the auspices of the U.S. DOE by LLNL under Contract DE-AC52-07NA27344.

Low temperature irradiation of crystalline materials is known to result in hardening and loss of ductility, which limits the usefulness of candidate materials in harsh nuclear environments. In bcc metals, this mechanical property degradation is caused by the interaction of in-grown dislocations with irradiation defects, particularly small dislocation loops resulting from the microstructural evolution of displacement cascades. In this work, we present a multi scale model encompassing dislocation dynamics (DD) simulations, crystal plasticity, and finite element (FE) simulations of bcc Fe containing various concentrations of dislocation loops produced by irradiation in an attempt to gain insight into the processes that lead to hardening and embrittlement. The DD simulations reveal a transition from homogenous to highly localized deformation at a critical loop density. Above it, plastic flow proceeds heterogeneously, creating defect-free channels in its wake. These simulations are then used to calibrate a tensorial crystal plasticity model capable of reaching strains in excess of 10%. The calibrated crystal plasticity model is used as the constitutive relation in FE simulations of polycrystalline irradiated Fe systems.

One of the challenges in the design of future nuclear power plant is to develop materials capable to resist in the hostile environment of a fusion reactor. Because of its low sputtering yield, low-activation, high melting point, high thermal conductivity and low thermal expansion, tungsten is one of the most attractive materials proposed for first wall applications in nuclear fusion reactors [1-3]. Even when W is assumed to be the best candidate as plasma facing material (PFM), some limitations have been identified that have to be defeated in order to fulfil specifications i.e. an important point of concern to the light species behavior (H, D, T and He). Nowadays some strategies to overcome these limitations are being investigated [4]. In this work we focus on the study of the capabilities of nanoW as PFM. Firstly, we report about DC magnetron sputtering deposition procedure, presenting the dependence of sample microstructure on deposition parameters. Microstructural characterization studies by XRD,TEM and SEM evidence that nanostructured samples are polycrystalline and are composed of columns with a diameterbetween 50 and 200 nm. Then, the thermal properties (conductivity and stability) are studied. These results illustrate that the column diameter does not significantly increases in the temperature range up to 400 C. For temperatures higher than 800 C the adhesion between W and the steel substrate has identified to be a major problem. Finally, the light species behaviour is characterized as a function of sample microstructure and implantation conditions. The role of the synergetic effects,when the samples are simultaneously exposed to different particle irradiation, in the light species behaviour is addressed. For this purpose resonant nuclear reaction (RNRA) experiments were performed using the H(15N,α)12C nuclear reaction in nanostrcutred (nW) and polycrystalline (pW) samples implanted with (i) H at an energy of 170 keV, (ii) sequentially implanted with C at an energy of 665 keV and H at 170 keV and (iii) simultaneously implanted with C and H at the above described energies. Implantations were carried out at a fluence of 5e16at/cm2 and at two different temperatures RT and 400 C. RNRA data evidence that the highest H retention is observed for the C and H co-implanted samples, being the lower one measured for those samples implanted only with H. In general, the H retention is higher for nW than for pW samples. Moreover, increasing the irradiation temperature up to 400 C drives the H to completely out diffusion in nW as well as, in pW samples. The role of microstructure and radiation-induced damage on light species behaviour will be discussed. [1] C. H. Wu et al. J. Nucl. Mater. 220-222 (1995) 860 [2] G. Federic et al. J. Nucl. Mater. 266-269 (11999) 14 [3] M. Kaufmann et al. Fusion Engineering and Design 82 (2007) 521-527 [4] Alvarez J., Rivera A., Gonzalez-Arrabal R., Garoz D., Del Rio E., Perlado J.M. Accepted in Fus. Sci. and Technol.

It has now been firmly established that the formation of relatively large <001> prismatic dislocation loops in iron, irradiated at temperatures exceeding 500°C, is associated with anisotropy elasticity effects, which dominate the dislocation microstructure at temperatures close to the temperature of the α-γ phase transition. Iron irradiated at lower temperatures shows no anomalies and the majority of the loop defects formed have ½<111> Burgers vector. An elasticity argument also suggests that only ½<111> loops should form in elastically isotropic tungsten under irradiation. Surprisingly, our recent observations of the microstructure of self-ion-irradiated tungsten show that, at low temperatures, a significant proportion of the prismatic dislocation loops formed have the anomalous <001> Burgers vector. The experiments also reveal a strong temperature dependence of the preferred Burgers vector orientation of both vacancy and self-interstitial prismatic dislocation loops. For example, at 300°C and1 dpa dose, 36% of the loops are of <001> type, while at higher temperatures the ratio changes, and by 800°C almost all of the loops observed after irradiation are of the ½<111> type. Atomistic simulations performed using several recently developed potentials for W predict that mesoscopic <001>-type prismatic loops will form in tungsten, and that there is a critical loop size below which loops with a <001> Burgers vector are more favourable than those of ½<111> type. Using finite-temperature simulations, we compute the free energy of the loops and show that the <001> loops become less favourable at high temperatures - in agreement with observations. Furthermore, simulations predict that an increase in the average loop size should shift the equilibrium in favour of the ½<111>-type loops, also in agreement with observations. We explore the origin of this anomalous stability of mesoscopic <001> loops, which signifies a striking failure of the elasticity argument in the mesoscopic loop-size limit, and highlight the role played by the dislocation core energy in determining the structure of mesoscopic dislocation loops formed in tungsten under irradiation. This work was funded partly by the RCUK Energy Programme under grant EP/I501045 and the European Communities under the contract of Association between EURATOM and CCFE. The views and opinions expressed herein do not necessarily reflect those of the European Commission.

Helium retention in irradiated tungsten leads to swelling, pore formation, sample exfoliation and embrittlement with deleterious consequences in many applications. In particular, the use of tungsten in future nuclear fusion plants is proposed due to its good refractory properties. However, serious concerns about tungsten survivability stems from the fact that it must withstand severe irradiation conditions. In magnetic fusion as well as in inertial fusion (particularly with direct drive targets), tungsten components will be exposed to low and high energy ion (helium) irradiation, respectively. A common feature is that the most detrimental situations will take place in pulsed mode, i.e., high flux irradiation. There is increasing evidence on a correlation between a high helium flux and an enhancement of detrimental effects on tungsten. Nevertheless, the nature of these effects is not well understood due to the subtleties imposed by the exact temperature profile evolution, ion energy, pulse duration, existence of impurities and simultaneous irradiation with other species. Kinetic Monte Carlo is the technique of choice to simulate the evolution of radiation-induced damage inside solids in large temporal and space scales. We have used the recently developed code MMonCa (Modular Monte Carlo simulator to study He retention (and in general defect evolution) in tungsten samples irradiated with high intensity helium pulses. The code simulates the interactions among a large variety of defects and impurities (He and C) during the irradiation stage and the subsequent annealing steps. In addition, it allows us to vary the sample temperature to follow the severe thermo-mechanical effects of the pulses. In this work we will describe the helium kinetics for different irradiation conditions. A competition is established between fast helium cluster migration and trapping at large defects, being the temperature a determinant factor. In fact, high temperatures (induced by the pulses) are responsible for large vacancy cluster formation and subsequent additional trapping with respect to low flux irradiation.

Although tungsten is considered the best candidate as a plasma facing component (PFC) in the divertor region in the International Thermonuclear Experimental Reactor (ITER), [1] severe morphology changes such as cavities, blisters, bubbles and nanostructure formation are expected to take place as a result of high fluence exposure to burning tokamak fusion plasmas. [2]. Increasing defect sinks in the tungsten microstructure by increasing the grain boundary area through the formation of ultrafine and nanocrystalline grain tungsten is one of the possible solutions to mitigate the irradiation damage. In this work, detailed irradiation study was performed on fine and ultrafine grained tungsten materials of multimodal grain size distributions and very high Vickers hardness (up to 600 kg/mm2) prepared by Spark Plasma Sintering (SPS) [3] and Severe Plastic Deformation (SPD) techniques. The irradiations were performed in the Nuclear Engineering at Purdue University and in the Dutch Institute for Fundamental Energy Research (DIFFER). The samples were irradiated with helium at different temperatures (200-950 °C), fluxes (1x1014-1x10^20 cm-2 s-1) and fluences (1x1016-1x10^23 cm-2). Scanning Electron Microscopy (SEM) and Transmission Electron Microscopy (TEM) studies were performed on the samples after irradiation. Several phenomena such as bubble, pores and nanostructure formation as well as recrystallization and grain boundary grooving were observed. Difference between SPS and SPD samples regarding the response to helium irradiation are discussed. Electron Backscatter Diffraction (EBSD) was performed on the SPD samples to investigate any relation between structure modification after irradiation and grain boundary orientation. All the results above are compared with a similar study on commercial tungsten samples of 15-25 micrometers grain sizes. Conclusions are presented regarding grain size refinement and irradiation tolerance of these materials to helium irradiation. [1] Lipschultz B, et al. Nucl. Fusion 47 (2007) 1189-1205. [2] Zinkle SJ, Ghoniem NM. Fus. Eng. Design, 51-52 (2000) 55. [3] El-Atwani et al., Materials Science and Engineering A 528 (2011) 5670-5677

Tungsten based alloys are being considered as likely candidate materials for the plasma facing components in a nuclear fusion reactor that have to withstand the most extreme environmental conditions including high temperatures, high fluxes of 14MeV neutrons and a large degree of helium implantation. There are still however major challenges to be faced with the fabrication of complex components due to the brittleness of tungsten and the effects of radiation damage on plasma-facing materials. A study has been performed on a range of tungsten-tantalum and tungsten-rhenium alloys to understand the effect of radiation damage on the mechanical properties. Tungsten-rhenium alloys are produced in reactor conditions by the transmutation of tungsten to rhenium due to the 14MeV neutrons: Additionally rhenium alloying has already having been shown to lower the brittle to ductile transition temperature of tungsten. Tungsten-tantalum alloys are of interest as tantalum is fully soluble in tungsten and may also have beneficial effects on mechanical properties. These alloys have been produced in various compositions (1, 2, 5, 10 & 25wt%), by the arc-melting of high purity powders, in order to analyse the effects of both alloying additions and radiation damage on the mechanical properties. Due to the limitations of current facilities to simulate the levels of neutron irradiation a first-wall reactor material will experience, tungsten ion and helium implantations have been used as a facsimile for the lattice damage. Implantations have been carried out in a range of conditions at 300°C, 500°C, and 800°C, using both single and multiply charged ions. The ion species are, tungsten only; helium only; tungsten followed by helium; tungsten and helium simultaneously. Damage levels of 0.6, 1.2 & 2.5 displacements per atom (dpa) have been produced. The nature of the small scale damage profile produced by ion implantation necessitates the use of micromechanical testing techniques to measure the changes in mechanical properties. Nanoindentation has been used to obtain indentation moduli and hardness data for a range of alloys, in both the implanted and unimplanted conditions. The results have shown a dramatic increase in hardness for He+ implanted specimens of up to 6.13GPa, with much more modest increases of up to 1.46GPa for W+ ion implanted material. A similar pattern was seen with the change in elastic modulus of up to 79.1GPa for He+ implanted material, but little change is seen in W+ ion implanted material. The results for the sequentially implanted material show a maximum hardness increase of 4.5GPa and a maximum increase in modulus of 40GPa while the simultaneously implanted material shows maximum increases of 4GPa in hardness but little change in elastic modulus. This shows the importance of considering the synergistic effects which can occur during dual ion implantations.

We review recent developments in the field of ab initio electronic structure theory and its application for studies of complex alloy systems. Basic ideas behind state-of-the-art techniques for first-principles theoretical simulations of the phase stabilities and properties of intermetallic compounds and alloys based on multiscale approach are outlined. We concentrate on methods that allow for an efficient treatment of compositional [1] and thermal [2] disorder effects, and illustrate their predictive power for two systems relevant for nuclear energy applications: Fe-Cr and Zr-based alloys. Ferritic low Fe-Cr steels with about 6 at.% Cr are considered as promising structural materials for fast neutron reactors due to their relatively low rate of swelling at elevated temperatures. In the binary Fe-Cr alloy the decomposition occurs either via the nucleation and growth mechanism or as spinodal decomposition, depending on the Cr content. However, at low chromium concentrations the alloys are anomalously stable [3]. We investigate the effect of hydrostatic pressure and ternary additions on the phase stability of Fe-Cr alloys, and show that in the ferromagnetic state both of them reduce the alloy stability at low Cr concentration. On the atomic level, the pressure effect can be explained by the suppression of the local magnetic moments on Cr atoms, which gives rise to a decrease of the Fe-Cr magnetic exchange interaction at the first coordination shell and, as a result, to the observed variation of the ordering tendency between the Fe and Cr atoms [4]. Hexagonal closed packed (hcp)Zr-based alloys are widely used in nuclear energy applications. In particular, Zr-Nb alloys (with about 1 at.% Nb) show strengthening behavior due to polymorphous martensitic αharr;β transition [5]. We show that the transition is associated with pronounced dynamical instabilities of body-centered phase of Zr, and present novel approach to evaluate free energy and describe phase transitions in such systems [2]. [1] A. V. Ruban and I. A. Abrikosov, Rep. Prog. Phys. 71, 046501 (2008). [2] O. Hellman, I. A. Abrikosov, and S. I. Simak, Phys. Rev. B 84, 180301(R) (2011). [3] P. Olsson, I. A. Abrikosov, L. Vitos, and J. Wallenius, J. Nucl. Mat. 321, 84 (2003); P. Olsson, I. A. Abrikosov, and J. Wallenius, Phys. Rev. B 73, 104416 (2006). [4] A. V. Ponomareva, A. V. Ruban, O. Yu. Vekilova, S. I. Simak, and I. A. Abrikosov, Phys. Rev. B 84, 094422 (2011). [5] S. A. Barannikova, A. V. Ponomareva, L. B. Zuev, Yu. Kh. Vekilov, and I. A. Abrikosov, Solid State Commun. 152, 784 (2012).

In a multiscale materials simulation approach mesoscale methods, such as phase field, can be informed by low length scale techniques, such a s molecular dynamics (MD). A particular challenge is the incorporation of nucleation phenomena in phase field. We present a method for the calculation of critical nucleus sizes and nucleation rates demonstrated for alpha prime precipitation in an FeCr alloy. A Monte Carlo based extension of the phase field method is proposed to utilize the nucleation data on the mesoscale level. Only recently has data become available for both the free energy of the solid solution and the interfacial free energy of the FeCr system. Free energy of the solid solution phase was obtained in our group using a switching Hamiltonian method in a molecular dynamics simulation using an empirical concentration dependent embedded atom (CDEAM) potential [1]. The resulting free energy surface extends all the way through the miscibility gap and covers a temperature range up to 700K well below the Curie temperature. The interfacial free energy was obtained using variance constrained grand canonical Monte Carlo calculations by Sadigh and Erhart [2]. The authors observe nearly spherical precipitates, indicating an orientation independent interfacial energy. Ab initio calculations by Lu et al [3,4] on the (001) and (110) interfaces support the observed isotropy and offer excellent agreement on the energy values. To calculate the total change in free energy we explore the two limits of a localized depletion zone around the precipitate and an infinitesimal non-local depletion as given by classical nucleation theory. This yields a bracketing of the critical radius of a nucleus in the system at any given temperature and matrix concentration, where the actual critical radius will be determined by the kinetics of the system and the resulting depletion profile. This work was supported by the nuclear energy advanced modeling and simulation (NEAMS) program, under the FMM project. [1] A. Caro et al, JOM 59 (2007) 50 [2] B. Sadigh and P. Erhart, in press http://arxiv.org/abs/1111.1880 [3] S. Lu et al, Phys. Rev. B, 82 (2010) 195103 [4] S. Lu et al, Phys. Status Solidi B, 248 (2011) 2087-2090

The key to perfect radiation endurance is perfect recovery. Since surfaces are perfect sinks for defects, a porous material with a high surface to volume ratio has the potential to be extremely radiation tolerant, provided it is morphologically stable in a radiation environment. Experiments and computer simulations on nanoscale gold foams reported here show the existence of a window in the parameter space where foams are radiation tolerant. We analyze these results in terms of a model for the irradiation response that quantitatively locates such window that appears to be the consequence of the combined effect of two length scales dependent on the irradiation conditions: (i) foams with ligament diameters below a minimum value display ligament melting and breaking, together with compaction increasing with dose (this value is typically sim;5 nm for primary knock on atoms (PKA) of sim;15 keV in Au), while (ii) foams with ligament diameters above a maximum value show bulk behavior, that is, damage accumulation (few hundred nanometers for the PKA's energy and dose rate used in this study). In between these dimensions, (i.e., sim;100 nm in Au), defect migration to the ligament surface happens faster than the time between cascades, ensuring radiation resistance for a given dose-rate. We conclude that foams can be tailored to become radiation tolerant. Nano Letters July 2012 dx.doi.org/10.1021/nl201383u

There is considerable interest in Fe-Cr steels reinforced with nanoscale oxides for structural components in the nuclear industry. Good creep resistance of such materials is ensured by the fineness and stability of the complex Y-Al-Cr-Ti-W oxides. Fe-base alloys containing 14-20% Cr have been selected due to their good oxidation-corrosion resistance, although there may be a susceptibility to low temperature embrittlement at high Cr contents, perhaps associated with the formation of Cr-rich precipitates. Alternatively, iron aluminides containing large amounts of Al and often small amounts of Cr have been examined as potential structural materials where oxidation resistance is ensured by the high Al content, but such alloys are believed to suffer low ductility due to the ordered matrix. The introduction of stable Y-base oxides to such aluminides has been shown to lead to good creep resistance. A study has recently begun to examine the structural stability and creep resistance of Fe-Al-Cr alloys, with Al and Cr contents intermediate between that of typical Fe-Cr and Fe-Al alloys. A range of such alloys has been prepared by mechanical alloying in order to introduce stable oxides, and studies of microstructure and high temperature strength and creep resistance have begun. This presentation will show some preliminary results of typical microstructures and oxide particles found in these materials, as well as some results of high temperature mechanical testing. It is of particular interest to compare the fineness of the oxide particles found here, and their thermal stability, with equivalent particles in the well-known Fe-Cr steels, and also to compare the high temperature mechanical behaviours.

The preparation of site-specific atom-probe (AP) samples containing localized features such as surface oxides and oxidized grain boundaries has become possible with the use of focused ion beams (FIBs) and will be described in this paper. Transmission electron microscopy (TEM), providing microstructural and chemical characterization of the same features, has also been used, revealing crucial additional information. The study of grain boundary oxidation in stainless steels and nickel alloys is required in order to understand the mechanisms controlling stress corrosion cracking in nuclear reactors. Samples oxidized under simulated pressurized water reactor conditions were used and FIB lift-out TEM and AP specimens containing the same oxidized grain boundary were prepared and fully characterized. The results from both techniques were found fully consistent and complementary. Chromium-rich spinel oxides grew at the surface and into grain boundaries. Nickel was rejected from the oxides and accumulated ahead of the oxidation front. Lithium, which was present in small quantities in the aqueous environment during oxidation, was incorporated in the oxide. All phases were accurately quantified and the effect of different experimental parameters analysed. Implications to SCC mechanisms will be discussed, including the possibility of detecting trapped Hydrogen.

Radiation induced segregation (RIS) is the process by which concentrations are altered at defect sinks due to a flux of defects produced during irradiation. RIS occurs when species are transported at different rates by the dominant radiation induced defects, typically vacancies and interstitials. Of particular importance is the RIS of Cr in steels of interest in the nuclear industry, as changes in Cr concentration at sensitive regions like grain boundaries can alter corrosion resistance or lead to precipitation of new phases. Cr RIS in fcc structured steels has been well established to show Cr depletion, while Cr RIS in bcc systems has proven more complex and can perhaps show both enrichment and depletion. While RIS in steels has been studied for over 40 years it has been difficult to establish the fundamental mechanisms underlying its manifestations. In particular, the relatively low temperature where RIS is observed (Cr RIS typically peaks around 700K) has meant diffusion data must be extrapolated significantly from high-temperature measurements. Furthermore, the absence of species dependent interstitial diffusion data has meant that interstitial contributions were estimated qualitatively or simply ignored. However, over the last 10 years work by our group and others has integrated ab initio energetics and diffusion modeling into RIS analysis and established that RIS is a balance of both vacancy and interstitial contributions. Ab initio based models can now simulate RIS that agrees well with a number of experimental results, including RIS profiles, total extent of RIS, and RIS vs. grain boundary type. We will discuss this updated understanding of RIS mechanisms and the modeling approaches that have enabled it.

Understanding of mechanisms of segregation of solute atoms at grain boundaries (GBs) in iron alloys is crucial for predicting the properties of irradiated materials. The presence of a high density of GBs in nanocrystalline (nc) materials has been found to play a dominant role in the mechanical properties of such materials and indicates that nc materials should have increased radiation resistance due to the large number of sink sources present in the samples. Recently, several computer simulations have been employed to yield important understanding of the GB structure and behavior under irradiation in bcc and fcc metals (see, for example [1]). While molecular dynamics studies are limited to small times and Monte Carlo simulations have the disadvantage of using a predefined ideal atomic lattice, a new powerful method that overcomes this obstacles has arisen in recent years. It is referred to by different authors whether as the phase-field-crystal [2] or atomic density function (ADF) model [3]. The ADF theory incorporates in a natural way elastic and plastic properties of crystalline materials and describes processes on diffusion time scales. Results for 3D GBs simulated using ADF model are presented. The symmetric tilt GBs obtained during simulation have form of arrays of dislocations. The GB energy as a function of tilt angle is in good agreement with the prediction of Read and Shockley [4] derived from the isotropic elasticity theory. The simulation results can be directly compared to the images obtained by HRTEM. Along with the simulation results theoretical predictions for the model phase diagram and elastic constants are given. [1] M. A. Tschopp et al. Scripta Materialia 64, 908 (2011); [2] K. R. Elder and M. Grant, Phys. Rev. B 70, 051605 (2004); [3] Y. M. Jin and A. G. Khachaturyan, Jour. of Appl. Phys. 100, 013519 (2006); [4] W. T. Read and W. Shockley, Phys. Rev. B 78, 275 (1950);

The dependence of the width of void-denuded zones (VDZs) on grain boundary (GB) characters was investigated in Cu irradiated with Helium ions at elevated temperature. Dislocation loops and voids formed near GBs during irradiation were characterized by transmission electron microscopy, and GB misorientations and plane normals were determined by electron backscatter diffraction. The VDZ widths at Σ3 <110> tilt GBs ranged from 0 to 24 nm and increased with the GB plane inclination angle. For non-Σ3 GBs, VDZ widths ranged from 40 to 70 nm and generally increased with misorientation angle. Nevertheless, there is considerable scatter about this general trend, indicating that the remaining crystallographic parameters also play a role in determining the sink efficiencies of these GBs. Voids were also observed within GB planes and their density and radius also appeared to depend on GB character. We conclude that GB sink efficiencies depend on the complete GB character, including both misorientation and GB plane orientation. This research is supported by US DOE, Office of Science, Office of Basic Energy Sciences.

Nanocrystalline and nanostructured metals have demonstrated significant promise as radiation tolerant materials. In MD simulations, nanocrystalline Ni and Fe have shown reduced interstitial survival and excess vacancy production immediately following the cascade event. Experimental work in Pd, Au, and Ni alloys has demonstrated a decrease in radiation damage with smaller grain size, with an absence of defect clusters found below a given grain size proportional to their respective self-diffusion lengths. Each case has demonstrated that grain boundaries in nanocrystalline materials provide effective annihilation sites for point defects during cascade events and subsequent diffusion. Our previous work has shown that dislocation loops in nanocrystalline iron also display a strong size dependence with grain size resulting from a strong sink contribution from grain boundaries across temporal scales. This work will explore the contribution of grain boundary character to the annihilation of individual point defects and dislocation clusters at boundary sinks. For this purpose, free-standing nanocrystalline iron films were deposited at Los Alamos National Laboratory using physical vapor deposition to create microstructures with predominantly high angle or low angle boundaries. The nanocrystalline film samples were irradiated in-situ using the Hitachi H-9000NAR TEM at Argonne National Laboratory&’s IVEM Tandem facility using 1MeV Kr²⁺ ions.

This talk discusses the progresses made for improving high temperature fracture characteristics of nanostructured ferritic alloys (NFAs). The NFAs, or often called the advanced oxide dispersion strengthened (ODS) steels, have been developed for a decade and considered to be the only Fe-based materials expected to maintain high temperature strength and radiation resistance in advanced fission and fission reactors. Although the NFAs have many desirable characteristics for high temperature, high dose applications, their excessive crack sensitivity is a serious drawback for such applications. Earlier studies showed that the fracture toughness of high strength NFAs was unacceptably low above 200°C; an easy decohesion at boundaries was responsible for the poor fracture resistance at high temperatures. Therefore, this research has been focused on improving grain boundary bonding. Two base materials (heat-treatable materials in as-extruded condition) with very fine grains and nanoparticles have been successfully produced: the design compositions of the two heats were Fe-9Cr-2W-0.4Ti-0.2V-0.12C-0.3Y2O3 (wt.%) and Fe-9Cr-2W-0.4Ti-0.2V-0.05C-0.3Y2O3. These alloys were designated as 9YWTV-PM1 and 9YWTV-PM2, respectively, and are considered as the base materials for further process development. Annealing & controlled rolling treatments are used to strengthen the weak boundaries and vicinity areas. The thermo-mechanical treatments (TMTs) were guided by the detailed results of calculation and microscopy. The minimum goal of 100 MParadic;m up to 700°C has been already achieved with controlled rolling processes, and some treatments resulted in higher fracture toughness, which is comparable to those of non-ODS ferritic-martensitic steels. Further process refinement and detailed characterization are underway.

To meet our growing energy needs, 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 ability to manage very high level of helium generated by transmutation reactions. It is believed that most of the radiation tolerance and He management properties in NFAs occurs at the metal/oxide interface. Thus an 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 behavior of the interface between the iron matrix and Y₂O₃ using density functional theory. In particular, the atomic structure of the interface will be presented. It was observed that, depending on the external partial pressure of oxygen, a critical number of defects -- iron-vacancies and/or interstitial oxygens -- are essential in stabilizing the metal/oxide interface. Importantly, the accommodation of these defects is very sensitive to the atomic structure of the interface, being enhanced at misfit dislocations at the interface. We discuss the implications of He storage at the interface in presence of such interfacial defects. Finally, we show the role of alloying elements, orientation relationship and interface misfit dislocations on the atomic and electronic structure of the metal/oxide interface, and segregation energies of the alloying elements. These results will form the basis of a phase-field model that will examine the nucleation and growth of Y₂O₃ precipitates in Fe. The insight gained in this research provides the fundamental science-based understanding needed to develop new NFAs tailored to meet challenges in fission and fusion applications, including safer operation of the current fleet of light water reactors.

Nanostructured ferritic alloys (NFAs) are attractive materials for nuclear applications due to their excellent high temperature creep properties and radiation tolerance. These properties result from the high densities of Y-Ti-O nanoclusters that lie along the grain boundaries and within the matrix. In nuclear applications, the NFAs can be subjected to nuclear reactions that result in the generation of α-particles, e.g. the (n,α) transmutation reaction. Over the commercial lifetime of a reactor (either fission or fusion), the continued increase in α concentration can lead to the formation of He bubbles due to the low solid solubility of He in metals. The large surface area and number density of nanoclusters may influence the formation and growth of the bubbles. Therefore a detailed understanding of the nucleation and evolution of He bubbles in NFAs is required. A transmission electron microscopy and atom probe tomography characterisation study has been conducted of a 14YWT NFA that has been irradiated with 335 keV He at a temperature of 400 °C to fluences up to 6.75x1017 ions cm-2. Bubbles were observed to form at both intra- and inter-granular locations. High number densities of ~2-nm diameter bubbles decorated the grain boundaries; were observed in apparent contact with large Ti(N,C) precipitates, and also occasionally within the matrix. Larger, ~5-nm diameter intra-granular bubbles were observed to be either randomly distributed or aligned along linear-type features within the grains. A sample irradiated to 6.75x1016 ions cm-2 was thermally treated, post irradiation, at a temperature of 650 °C for up to 100 hours. Analysis of this sample revealed that all the bubbles had undergone growth with the sole exception of those lying along dislocations. The role of the Y-Ti-O nanoclusters, oxygen-vacancy complexes, and extended defects such as dislocations, grain boundaries and large Ti(O,C,N) particles on the bubble nucleation and growth mechanisms will be discussed. Research sponsored by the Division of Materials Sciences and Engineering, Office of Basic Energy Sciences, U.S. Department of Energy. Research at the Oak Ridge National Laboratory SHaRE User Facility is sponsored by the Scientific User Facilities Division, Office of Basic Energy Sciences, U.S. Department of Energy.

Analysis of microstructure evolution in irradiated Zr and Zr alloys based on the modern radiation damage theory suggests some important features of self-interstitial atoms (SIAs). Alignment of vacancy loops and voids along basal planes requires anisotropic interstitial transport with a dominant contribution along the basal plane. Under neutron irradiation this can be explained by one-dimensional mobility of SIA clusters but experiments under electron irradiation indicate unambiguously that even the single SIA should exhibit anisotropic diffusion. No experimental information is available on SIA properties in Zr and the data obtained by ab initio calculations within the last decade reported stable SIA configurations that should provide essentially three-dimensional diffusion. To clarify this issue, an extensive investigation of SIAs in Zr has been performed from first principles using two different codes. It was demonstrated that simulation cell size, type of pseudopotential, the exchange-correlation functional, and the c/a ratio are crucially important for determining the properties of interstitials in hcp Zr. The most stable SIA configurations lie in the basal plane, which should lead to SIA diffusion mainly along basal planes. The results provide a confirmation of basic mechanisms for microstructural evolution under irradiation.

The 2011 Fukushima Daiichi nuclear accident involved the hydrogen explosion due to an oxidation of Zry-4 cladding tube in the high temperature steam leaving the excess hydrogen gas. In this work an effort has been paid to explore a method to reduce the oxidation by a highly adhesive protective SiC coating on the Zry-4 tube. SiC is an excellent candidate material for high temperature oxidative environment applications. In the coating of SiC on Zry-4, the prime concern is the adhesion at the elevated temperature due to the difference in the thermal expansion coefficients. In this work an ion beam was irradiated onto the thin SiC coating layer (~ 50nm) deposited by an e-beam evaporative deposition and then an additional coating is followed to an aimed thickness. As a result, the ion beam mixed SiC film was still intact on the Zry-4 substrate after heating to ~1000 °C although the difference in the thermal expansion coefficients is more than 5 times, but the crack was formed in the SiC coating layer, confirming the enhanced adhesion by ion beam mixing. Hence, the oxidation protection was greatly improved. An effort is being paid to healing the crack by a multiple coating/IBM process in order to further improve the oxidation protection. One of the merits of the ion beam mixing (IBM) is that a pretreatment for the contamination free surface is not imperative prior to the coating on the metallic substrate.

In Pressurized Water Nuclear Reactors (PWR), Reactor Cooling System (RCS) gets corroded with time, and corrosion products start depositing on fuel rods. The deposition of corrosion products on fuel rods, which is called CRUD, results in Axial Offset Anomaly (AOA), and eventually in fuel failure. Several ionic, molecular, and cluster forms of Fe, Cr, B, and Ni play an important role in CRUD deposition and growth. The evaluation of the deposition rates of these different species is necessary to understand the growth of CRUD layers, and develop a model. We have developed an experimental test facility that simulates the operating environment of a Light Water Reactor. The experimental chamber which is filled with the coolant has a Zircaloy tube in its center. A cartridge heater is placed inside the Zircaloy tube throughout its length. This design replicates the thermal conditions of an actual fuel rod inside the reactor. Initial experiments are focused on investigating the kinetics and mechanism of the deposition of Fe, Ni, and B in ionic and molecular forms on a Zircaloy tube surface, in LWR conditions. In particular, the effect of temperature on sub-cooled boiling and Fe2O3 deposition on Zircaloy tube are studied. The samples of the deposits are recovered from the different regions of the Zircaloy tube at different times. SEM/TEM-EDS characterization techniques are used to evaluate the deposition rates.

Despite being one of the original nuclear materials, surprising few molecular dynamics (MD) simulations have been performed to understand radiation response in graphite. The enormous difference between the MD literature for nuclear graphite and that of metals and oxides can be traced to the challenges associated with the description of bonding in carbon, in particular the anisotropic interactions which are central to graphite. Aside from point defect energetics and estimates of threshold displacement energies, very little is known at the atomistic level about cascade behaviour, dynamical effects associated with temperature or recovery from thermal spikes. In a modern context, understanding of radiation processes in graphite is motivated by lifetime extensions associated with Gen-II designs and new Gen-IV technologies such as the high-temperature graphite-moderated design which was recently selected as the next-generation nuclear technology in the USA. We have performed what is perhaps the first systemic study of radiation response in nuclear graphite using molecular dynamics. Chemical bonding is described using the Environment Dependent Interaction Potential (EDIP), while short-range interactions are modelled using the standard ZBL approach. Cascade simulations reveal that nuclear graphite behaves a manner distinct from metals and oxides, with the cascade primarily generating point defects, in contrast to connected regions of transient damage as are familiar from metals and oxides. Cascade simulations in other pure carbon phases highlight the unique attributes of radiation damage in graphite. Comparison of defect energetics computed with EDIP against values taken from density-functional-theory show that the underlying description of defect behaviour is sound. At the level of defect creation itself, the MD simulations quantify threshold displacement energies for which a broad range of values have been reported in the literature.

Future high temperature nuclear reactor cores with potential temperature excursions to 1200 °C will present new challenges for hot fuel element and structural materials in a high neutron irradiation environment. Graphite, silicon carbide (SiC), and ceramic matrix composites (CMCs) of these materials are potential candidates because of their very-high-temperature strength, corrosion, and neutron irradiation resistance. C _f/C and SiC_f/SiC composites present a matrix reinforced with polycrystalline continuous fibers that can impose an asymmetric geometry on the material properties. Monolithic graphite and SiC may also develop geometric anisotropic features due to swelling and fracturing under non uniform thermal and structural stresses. For example, fractures will form approximately parallel to uniaxial compressive stress [1] or perpendicular to tension [2]. A diagnostic tool that could provide dynamic information of material anisotropy in situ would be valuable for studying and monitoring the anisotropic behavior of materials in extreme temperature and neutron irradiation reactor core environments. Active millimeter-wave (MMW) radiometry can provide direction resolved remote measurement of material emissivity and submillimeter dimension changes that can be directly correlated with material temperature, resistivity, and swelling [3]. MMWs refer to the electromagnetic wavelength range of 10 - 0.1 mm that can be guided at high temperature to remote locations. A dual 137 GHz heterodyne radiometer system with a ceramic waveguide was used to study SGL NGB17 nuclear graphite inside an electric furnace at temperatures up to 1250 °C. The radiometers were oriented with superimposed, orthogonal polarized views over a 22 mm 1/e² diameter spot. Samples with linear 100 mu;m wide grooves and 1.3/mm spacing were evaluate for anisotropic sensitively as an idealized model for fracturing. For a groove depth of 1.6 mm, a ~20% difference in thermal emission was observed uncorrected for waveguide losses. Finite difference time domain analysis of corrugated surfaces predicts that the anisotropy will increase with narrower grooves (smaller fractures) and with dept with cyclic quarter wavelength modulation [4]. The anisotropy is expected to be independent of groove density when it exceeds >2/wavelength. These observations suggest that it may be possible to remotely observe micro fracturing and the direction of the underlying stress in real time. *This research is being performed using funding received from the DOE Office of Nuclear Energy's Nuclear Energy University Programs. 1. Z. P. Bazant, Scaling of Structural Strength, chap. 5, p.121, 2nd edition, Elsevier, Oxford, 2005. 2. A. Hodgkins et al, Materials Science and Technology, vol. 22, p1045, 2006. 3. Paul P. Woskov and S. K. Sundaram, Proc. 2010 MRS Fall Meeting, Symposium R, Boston, 2010. 4. B.Plaum, E. Holzhauer , C. Lechte, J Infrared Milli Terahz Waves, vol. 32, p482, 2011.

Silicon carbide (SiC), as one of the most investigated wide-band gap ceramic materials, has good thermal conductivity, large breakdown voltage, high strength, as well as outstanding chemical and mechanical stability. Research and development in recent years have been devoted to fully utilize this functional material for high-power, high-frequency, and high-temperature electronic and space applications. SiC materials and composites are also being considered as key engineering materials in nuclear applications for operation in harsh environments, such as structural or cladding materials and fuel coatings in fission reactors, structural components in fusion reactors or an inert matrix for transmutation of plutonium. For applications in advanced electronics and nuclear environments, enhancing the radiation-tolerance in SiC is highly desirable, and defect dynamics needs to be understood to predict performance and safe operation. The current work presents clear experimental evidence and microstructural insights on the extraordinary irradiation-resistant behavior in a nano-engineered SiC (NE-SiC) structure consisting of nanograins with high-density stacking faults. While full amorphization at room temperature is achieved at a dose of ~ 0.3 dpa under Si irradiation in single crystals, the crystalline structure of the NE-SiC is retained to much higher irradiation doses, which indicates an order of magnitude increase in irradiation resistance. The enhanced irradiation resistance in NE-SiC under Si irradiation is attributed to more efficient point defect annihilation mechanisms, which is consistent with a defect-stimulated model for amorphization.

Silicon carbide (SiC) is an important material for a wide range of engineering fields, such as a structural and cladding material in fission reactors. Recent research has experimentally revealed that a nano-engineered SiC (NE-SiC), containing nanograins with high-densities of stacking faults, exhibits extraordinary irradiation resistance. In order to gain insights on the underlying mechanisms, we performed theoretical simulation on defect behavior in NE-SiC. Molecular dynamics simulations were performed using the LAMMPS code. A Brenner-type potential, which sufficiently reproduces defect formation and migration energies, was utilized. Three different simulation cells were prepared in order to depict the fundamental nature of the NE-SiC: (1) 3C-SiC single crystal, (2) 3C-SiC with extrinsic stacking faults, and (3) 3C-SiC with intrinsic stacking faults. Initially, defect production created by displacement cascades was simulated in each of the three cell configurations. No significant difference was observed among those three systems, not only in the number but also in the nature of the defects created. Secondly, defect mobility was evaluated, and it was found that the mobility of interstitial Si was enhanced in NE-SiC, while that of interstitial C and vacancies was largely unchanged. The enhanced mobility of the interstitial Si was attributed to the stacking faults because the mobility was clearly dependent on stacking-fault density. Extrinsic and intrinsic stacking faults affected the defect migration in different ways, reflecting the local environment along a migration path. Finally, defect migration paths and corresponding migration barriers were surveyed using the nudged elastic band method. It was shown that the stacking faults created a lower-barrier path for interstitial Si migration. Additional insights on defect production and migration processes obtained from the simulations will be provided, and then the mechanisms for the high radiation resistance in NE-SiC will be discussed.

Understanding the interaction of extended defects such as grain boundaries (GB) with interstitials and vacancies is an active area of research, both from an experimental and computational perspective. While the influence of grain boundaries on the mechanical properties of crystalline materials is reasonably well-understood, much less is known about their role in radiation damage of materials. Molecular dynamics investigations are carried out to understand the role of grain boundaries on radiation damage, particularly with respect to displacement cascades and final distribution of point defects such as vacancies and interstitials for two symmetric tilt grain boundaries in β-SiC - a key structural material in next generation nuclear reactors. Our studies show that the spatial distributions of point defects are influenced by the presence of grain boundaries. Dynamic propensity and directional displacements averaged over an isoconfigurational ensemble, a new analysis approach in cascade simulations, is employed to characterize the statistical nature of atom mobility, and to evaluate possible self-healing of point defects near the grain boundaries.

SiC has been attracting a great deal of attention as a potential fuel cladding material of PWR especially after Fukushima Dai-ich accident due to its superior high temperature oxidation resistance. However, one of the most important issues to be addressed before using SiC as the fuel cladding is to clarify how SiC would behave in a normal reactor operation condition. The corrosion behavior in a normal operation condition would be one of the main criteria to determine the application of SiC in the current PWR nuclear fuel even though SiC can be more desirable to protect the fuel claddings from the oxidation in the accident condition. In this study, the corrosion behavior of SiC was investigated in several corrosion environments which simulated the PWR coolant conditions. The SiC specimens used in the corrosion tests were prepared by a chemical vapor deposition (CVD) and a plasma spraying (PS) to examine the effect of the manufacturing route on the corrosion behavior. Corrosion test for the SiC specimens were performed in a PWR-simulating loop, 360oC static water and 400oC steam conditions with the weight change being measured periodically. It was revealed that the entire specimens exhibited the weight loss irrespective of the corrosion conditions tested. The corrosion behavior of SiC was significantly different depending on the manufacturing route in the same corrosion condition. The surface of the oxide formed on the specimens after the corrosion test was observed by a scanning electron microscopy (SEM) and the cross section of the oxide was analyzed by a transmission electron microscopy (TEM). The effects of the corrosion environment and the manufacturing route on the corrosion behavior of SiC were discussed on the basis of the oxide morphology change.

Materials used in the nuclear power plant components are exposed to irradiation and thermal environments during operation, resulting in the degradation of materials performance. Development of predictive models of such degradation is very important for the safe operation of nuclear power plants, and understanding of the microstructural changes in nano-meter length scales is essential for such a goal. In this paper, we will present our recent activities on characterizing the microstructural changes in nano-meter scale, and on developing models to describe materials behavior. Atom Probe Tomography (APT) is applied to quantitatively characterize the nano-feature formation in neutron-irradiated ferritic low alloy steels, austenitic stainless steels and zircaloys as well as thermally-aged duplex stainless steels. Then we will discuss utilization of APT information for the prediction of mechanical property changes.

There are many interests in prolonging the lifetime of Pressure Water Reactor (PWR). The lifetime of a PWR depends strongly on the integrity of one of the main barrier between the reactor core and the environment: the reactor pressure vessel (RPV). It is known that irradiation results in hardening and embrittlement of the RPV steel. Since the properties of the RPV steel depend of its microstructure, it is crucial to anticipate the evolution of this microstructure. In this context, the European project LONGLIFE started in 2010. A part of this project is dedicated to the characterization of the microstructural evolution of different RPV steels from Eastern and Western Europe under neutron irradiations. Steels and irradiation conditions are selected to identify the effect of neutron flux and dose and the influence of the steel chemical composition. Complementary techniques (Atom probe Tomography, Small Angle neutron Scattering, electron Microscopieshellip;) are used to get an accurate description of the microstructure. In this talk, Atom Probe results will be described and discussed. Different steels will be compared. The evolution of the microstructure (solute clusters, segregationshellip;) with neutron dose will be shown. A comparison with the data from others techniques will be given when available.

SiC and ZrC are promising candidates for applications in nuclear reactors due to their mechanical properties, high melting temperatures, and abilities to contain fission products. However, their irradiation response has still not been fully understood. We have studied the energetics of defect migration and annealing processes by using density functional theory calculation. We propose an approach to find the annealing reaction path effectively for the complicated energy landscapes. The reaction path is first established based on ab initio molecular dynamics simulations. Then the barriers are calculated via climbing image nudged elastic band methods and are determined by removing the constraints gradually. These results are confirmed by the drag method and compared with the literature data for SiC. These annealing processes have important implications in understanding amorphization, damage recovery, and irradiation response in SiC and ZrC.

In the past few decades, first-principles calculations have demonstrated its power in accurately predicting the ground state properties of a vast range of materials at zero temperature. The development of ab initio electronic structure methods has made it possible to calculate force constants with no arbitrary parameters, which allows for accurate calculation of phonon frequencies and the thermodynamic properties derivable from them. In the present study, a systematic study based on first-principles calculations along with the quasiharmonic approximation has been conducted to calculate the thermodynamic properties of CexTh1-xO2 solid solution. The predicted density, thermal expansion coefficients, specific heat capacity and thermal conductivity for CexTh1-xO2 solid solution all compare well with available experiments. The thermal expansion coefficient for ThO2 increases on substituting it with CeO2, and a complete substitution shows the highest expansion coefficient. On the other hand, the mixed CexTh1-xO2 (0.25le;xle;0.75) solid solution generally exhibits lower heat capacity and thermal conductivity than individual ThO2 and CeO2. Our calculations indicate a strong effect of Ce concentration on the thermodynamic properties of CexTh1-xO2 solid solution.

Uranium (U) exhibits a high temperature body-centered cubic (bcc) phase that is often stabilized by alloying with transition metals such as Zr, Mo, and Nb for technological applications. One such application involves U-Zr as nuclear fuel. Under fission reactor conditions, processes based on radiation damage are of critical importance. Such phenomena include, but are not limited to, radiation induced segregation, void formation and growth, constituent redistribution and swelling. Also, mechanical properties are largely linked with damage on a displacements per atom basis. This often serves to provide lifetime limits of operation for structural materials and can serve as a limiting factor in the operational lifetime of metallic fuel. Characterizing and understanding the displacement energy is a first step to understand the rate and accumulation of radiation damage. In this work, the displacement energy is calculated in bcc U as a function of angle for eleven different directions, including the three high symmetry directions: <100>, <110> and <111>. The probability of stable Frenkel pair formation is analyzed as a function of energy. Finally, the displacement energy in the <135> direction is analyzed. The <135> direction is an average direction that also minimizes channeling in the bcc crystal structure.

To understand the complex fuel cladding chemical interaction between U-Zr metallic fuel with stainless steel, in particular to document the effects of Zr, Cr and Ni alloying elements, systematic multicomponent diffusion studies were carried out using pure U, U-10wt.%Zr, pure Fe, Fe-15wt.%Cr, and Fe-15wt.%Cr-15wt.%Ni alloys. Solid-to-solid diffusion couples of U vs. Fe, U vs. Fe-15wt.%Cr, U vs. Fe-15wt.%Cr-15wt.%Ni, U-10wt.%Zr vs. Fe, U-10wt.%Zr vs. Fe-15wt.%Cr and U-10wt.%Zr vs. Fe-15wt.%Cr-15wt.%Ni were assembled and annealed in the temperature ranging from 580 to 700 °C. Microstructures, phase constituents, and composition developed during thermal anneal were examined by scanning electron microscopy, electron probe microanalysis and X-ray energy dispersive spectroscopy. In pure U diffusion couples, only two intermetallics, U6Fe and UFe2, developed in all diffusion couples, and U6Fe was observed to grow faster than UFe2. The solubility of Cr or Ni in U6Fe and UFe2, and their influence on diffusion behavior in each phase were assessed. Addition of Zr in U made the interdiffusion microstructures extremely complex, and multiple, multiphase layers with various intermetallic precipitates were observed. U diffused faster into Fe-alloy side than Zr, while, into the U-alloy, Ni diffused faster than Fe, and Cr diffused the slowest. A phenomenological description for diffusional interaction is proposed to describe the discrepancies reported among previous studies on interaction between U-Zr and Fe alloys.

Heavy liquid metals (HLMs), such as Pb or Pb-Bi eutectic, are under wide investigation for use as coolant materials in advanced nuclear systems due to their attractive thermal and neutronic properties. Nevertheless, a major bottleneck for the deployment of these systems is the ability of the foreseen structural steels to withstand erosion degradation and corrosion phenomena at high temperatures [1]. Although several protection techniques are already being developed [1-3], a breakthrough in corrosion protection of materials is still required to allow further improvements in the field. In this work, a novel technique, namely pulsed laser deposition (PLD), was employed for synthesizing advanced Al2O3 barrier layers on 9Cr1Mo type martensitic steel. The characterization of the environmental barriers was performed using scanning electron microscopy (SEM), X-ray diffraction (XRD) and high-resolution transmission electron microscopy (HR-TEM). The tribological and mechanical performances were investigated through nano-scratch tests and a combination of nanoindentation, ellipsometry and Brillouin spectroscopy (BS). Metal-like mechanical properties and enhanced hardness were attained, owing to the nanostructure of the coatings. The latter consists of an embedding of ultra-fine nano-domains (2-5 nm) in an amorphous alumina matrix. Our results show that these ductile ceramic coatings are moderately stiff (E=200 GPa, poisson ratio = 0,29, H = 10 GPa), and that they possess a unique combination of compactness, superior plastic and tribological behaviour, and strong interfacial bonding. The corrosion resistance of the barrier layers was tested at a preliminary stage by exposing samples to HLM environment at 600°C for 500 hours. No sign of corrosion was found. The proposed technology is therefore a suitable and promising candidate for protection of steels from HLM corrosion. Nevertheless, further work is required, especially concerning long-term behaviour in stagnant and flowing HLM, creep resistance and irradiation toughness, as well as performance under the combined effect of these phenomena. [1] V. Engelko et al., J. Nucl. Mater. 415, 270 (2011). [2] A. Weisenburger et al., J. Nucl. Mater 376, 274 (2008). [3] P. Dou, R. Casada, J. Nucl. Mater. 409, 177 (2011).

Creep rupture life and microstructural degradation have been studied in two heats of Gr. 91 steels. The coarsening of subgrains and precipitates, mainly M23C6 and MX, has been evaluated during static aging and creep. Hardness of head (static aging) and gauge (creep) portions of crept sample were measured to know its relation with microstructural degradation during long-term exposure. There are two creep regions with different creep characteristics: short-term creep region “H”, where precipitates and subgrains are thermally stable, and long-term creep region “L”, where thermal coarsening of precipitates and subgrains appear. In region “H”, degradation of microstructure is mainly due to recovery of subgrains controlled by creep plastic deformation, and precipitates do not have a major role. The appearance of three recovery mechanisms was found during long-term creep, namely: strain-induced recovery, pure static recovery and strain-assisted static recovery. By equated correlations between subgrain size, precipitates and hardness, the contribution of three recovery mechanisms to subgrain coarsening and hardness drop were calculated for two creep conditions at 700oC in long-term creep region, where breakdown of creep strength has happen. The calculated data accord well with experimental data obtained from aging and creep samples. The contribution of static recovery to the subgrain coarsening and consequent hardness drop during long-term creep increases with increasing creep time. The significant contribution of both static recovery mechanisms can result in the breakdown of creep strength in long-term creep region.