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 techni