The third accelerator of the multi-ion irradiation platform JANNUS (Joint Accelerators for Nanosciences and NUclear Simulation), a 6SDH-2 Pelletron from National Electrostatic Corporation, Middleton was installed at Saclay in October 2009. The first triple beam irradiation combining Fe, He and H ion beams has been performed in March 2010 on a Fe-12Cr model ODS alloy. In the first part of this paper, we give a technical description of the triple beam facility. Then, we present its performances and experimental capabilities. Typically, damage dose up to 100 dpa can be reached in 10 hours irradiation with heavy ion beams, with or without simultaneous bombardment by protons, helium-4 ions or any other heavy ion beam. In the second part of this paper, we illustrate some recent experiments relative to advanced nuclear ceramics and composite materials developed in the frame of the Generation IV Forum research program or for fusion applications.

We consider a tight binding model for magnetic systems, in which we allow atoms to become charged and to interact via the long ranged Coulomb interaction to a published tight binding model; this is then applied to the study defects in ferromagnetic iron. We encounter several problems with achieving self consistency with existing schemes. To address the issue of instability and slow convergency we developed a robust, efficient scheme for charge and spin self consistency. This is based on minimizing an extended form of the Harris-Foulkes functional which includes spin, leading to a Newton-Raphson iterative procedure. We then apply this to both bulk and defect calculations for iron.

Mo3Ru5MPd (M = Ru, Rh, Pd) as the simulated materials for the undissolved residue in the nuclear fuel reprocessing were prepared by arc melting method. The physical properties and oxidation behavior of the alloys were evaluated from viewpoint of the safety and economy in the reprocessing. The electrical resistivity, ρ, of Mo3Ru5RhPd was shown to be 0.8 μΩm at room temperature. On the other hand, the ρ values of samples without Rh were marked at 0.4 μΩm. The thermal properties of the each sample had the different thermal transfer characteristics. In particular, although the thermal conductivities of Mo3Ru5RhPd and Mo3Ru5Pd2 samples show almost the same value, the lattice thermal conductivities of both samples showed different values. Oxidation behavior was analyzed using the thermogravity(TG) and differential thermal analyses(DTA). The TG curve of each sample by oxidation showed different results. These results indicate that the simulated materials of the alloys without Rh: Mo-Ru-Pd were not appropriate to simulate the thermophysical characteristics of the typical simulated materials with undissolved residue Mo-Ru-Rh-Pd alloys. Therefore, in the spent nuclear fuel reprocessing, the mock test of reprocessing without to use Rh is difficult to carry out.

The study of the irradiation effects on titanium surfaces in oxidizing environment using multi-charged Argon ions in the MeV range shed into light the following points:-Significant oxide film thickening for the film grown at 500°C under irradiation at 4 and 9 MeV, by comparison with the TiO2 rutile film grown under same environmental conditions without irradiation;-Formation of large round –shaped craters, of diameter approaching 200 nanometers, at the titanium surface under irradiation at 500°C provided that the environment is enough oxidizing or provided that the metal surface is covered by a sufficiently thick oxide film.

Practically, and for the present system, the superficial craterization is observed if the thickness of the superficial oxide is equal to twice that of the native oxide (~3 nm).

Crevice-corrosion tests were performed in gamma-ray irradiated high-temperature water of 288 °C on Type 316L stainless steel. The gamma-ray dose rate was about 30 kGy h−1. Tested specimen surfaces were analyzed using SEM, laser Raman spectroscopy and TEM/EDX. Experimental data were presented in order to show the differences made by the irradiation and crevice-shape simulated structure. Both gamma-ray irradiation and crevice-shape simulated structure changed the corrosion phenomena. On the gamma-ray irradiated crevice-shape simulated surface, α-Fe2O3 particles more than 5 μm in diameter were observed. It suggested that corrosion environment on the crevice-shape simulated surface became severer by gamma-ray irradiation.

Processes of severe plastic deformation have been investigated for a wide range of ductile alloys over the past decade, generally with an objective of refining the microstructural scale, for example the grain size, but have hardly been considered for intermetallics. This presentation discusses processing of a boride-containing Fe3Al alloy using a multidirectional, high-strain and high-temperature forging technique. Iron aluminides with relatively low Al contents can be regarded as Al-rich ferritic steels with outstanding oxidation-corrosion properties. However, as for many ferritic steels, they show poor creep resistance at temperatures above about 600ºC. The deformation processing leads to a material with large grain size and refined dispersion of thermally-stable boride particles. The particles produce a large increase in creep strength under conditions of moderate stresses and low strain rates at temperatures near 700ºC. This high-strain forging technique can be seen as an intermediate processing method between conventional wrought metallurgy and mechanical-alloying powder metallurgy, whereby an initially coarse and inhomogeneous dispersion of second phase is refined and made more homogeneous, and can be considered as a useful processing technique for a wide range of particle-containing materials.

Millimeter-wave thermal analysis instrumentation is being developed for characterization of high temperature materials required for diverse fuel and structural needs in extreme high temperature reactor environments. A two-receiver 137 GHz system with orthogonal polarizations for anisotropic properties resolution has been implemented at MIT and is being tested with graphite and silicon carbide specimens at temperatures up to 1300ºC. Real time measurement sensitivity to submillimeter surface displacement and simulated anisotropic surface emissivity is demonstrated.

We present here new information on the effect of irradiation temperature on the strength and mechanical anisotropy of Zr-2.5%Nb CANDU pressure tube material. Polished samples aligned normal to the transverse (TN), axial (AN) and radial (RN) directions of the pressure tube were irradiated at 300°C with 8.5 MeV Zr+ ions to assess the effect of concurrent thermal annealing of the irradiation damage. Constant-load micro-indentation creep tests were performed at 25°C at indentation depths from 0.1 to 2.0 μm on the ion irradiated samples.

The increase in the initial indentation stress with increasing levels of Zr+ ion irradiation at 300°C was lower than that reported earlier for similar samples exposed to Zr+ irradiation at 25°C. While the anisotropy of the indentation stress decreased significantly with Zr+ ion irradiation, the level of the decrease was reduced when the irradiation was performed at 300oC compared to 25oC. The apparent activation energy ΔG0 of the obstacles that limit the rate of dislocation glide during indentation creep did not change with indentation direction but did increase with increasing levels of Zr+ ion damage. The values of ΔG0 were, again, lower for samples that were irradiated at 300°C than for those irradiated at 25oC.

The observed differences in the magnitude of, and the anisotropy of, the initial indentation stress and also the decrease in the apparent activation energy of the indentation creep process of Zr-2.5%Nb samples irradiated with Zr+ ions at 300oC compared to those irradiated at 25oC indicate the effect that concurrent thermal annealing has on the accumulation of irradiation damage. The effect of irradiation temperature on reducing the degree of, and the strength of, irradiation induced crystallographic damage must therefore be considered when predicting the strength and thermal creep behaviour of irradiated nuclear materials.

In the fusion irradiation environment, helium created by transmutation will play an important role in the response of structural materials to neutron radiation damage. Recently we have developed a new 3-body potential to describe the Fe–He interaction in an Fe matrix. We have used this potential to investigate the equilibrium state of He bubbles embedded into the bcc Fe matrix. We have investigated bubble size, He content and temperature effects. It was found that the equilibrium He content is rather low and at a room temperature it is ~0.38 to 0.5 He per vacancy for bubble diameters from 1 to 6 nm. At constant bubble size, the equilibrium He/vacancy ratio decreases with temperature increase. For bubbles of 6 nm diameter it goes down as low as ~0.25 at 900K. The results are compared with the capillarity model often used for estimating the equilibrium pressure of He bubbles.

An oxide dispersion strengthened steel is produced which contains Y-Al-Ti-O nanoparticles with an average diameter of 21 nm. HRTEM analysis shows that the chemical composition of the Y2O3 oxide is modified with perovskite YAlO3 (YAP), Y2Al5O12 garnet (YAG) and Y4Al2O9 monoclinic (YAM) particles. Irradiation of these alloys was performed with a dual ion beam system operating simultaneously with 2.5 MeV Fe+ to 31 dpa and 350 keV He+ to 18 appm/dpa. Ion bombardment causes atomic displacements resulting in vacancy and self-interstitial lattice defects and dislocation loops. TRIM calculations for ODS steel indicate a clear spacial separation between vacancies and self-interstitials at which the vacancy distribution is close to the surface and the interstitials are deposited at a deeper position. The helium atoms mainly accumulate in the vacancies. Fine He cavities with diameters of a few nanometers were identified in HRTEM images. Additionally to structural changes, irradiation generated defects also affect the mechanical properties of the ODS steel. These were investigated by nanoindentation, which is a suitable measuring method as the irradiation damage is created within a thin surface layer. A clear hardness increase in the irradiated depth region was observed, which reaches a maximum close to the surface. This indicates the He condensation in the vacancy dominated region predicted by the simulations.

We have performed a detailed analysis of the magnetic (collinear and noncollinear) order and atomic and electron structures of UO2, PuO2 and UN on the basis of density functional theory with the Hubbard electron correlation correction (DFT+U). We have shown that the 3-k magnetic structure of UO2 is stabilized for the Hubbard parameter value of U=4.6 eV (while J=0.5 eV) when Dudarev’s formalism is used. UO2 keeps cubic shape in this structure. Two O atoms nearest to each U atom in direction of its magnetic moment move toward this U atom. Neither UN nor PuO2 shows the energetical preference for the rhombohedral distortion, in contrast to UO2, and, thus, no complex 3-k magnetic structure in these materials. Both materials have the AFM tetragonal <001> structure at reasonable choice of parameters U and J.

The reported work focuses on developing antidiffusion barriers capable to increase the thermal stability of metal contacts above 700 C. In the chosen approach, such an antidiffusion barrier consists of several bilayers of materials with different crystalline structures. It has been demonstrated that an interface between such materials effectively blocks the atomic interdiffusion. In this work the following groups of materials were used as the bilayers: ZrB2 and ZrN and TaSiN and TiN. The materials were deposited by means of room temperature sputtering from elemental and compound targets in inert Ar and reactive Ar+N2 atmospheres. The structures were characterised using secondary ion mass spectroscopy depth profiling and scanning electron microscopy cross sectional imaging directly after deposition and after degradation. I-V characteristics were measured and contact resistivities were determined from the circular transmission line method.

In this paper, we present technique to fabricate nanopatterns on Cu thin films via an electrochemical nanomachining (ECN) using an atomic force microscope (AFM). A conductive AFM cantilever tip (Pt/Ir5 coated) was used to form an electric field between tip and Cu substrate with applying a voltage pulse, resulting in the generation of an etched nanopattern. In order to precisely construct the nanopatterns, an ultra-short pulse was applied onto the Cu film through the AFM cantilever tip. The line width of the nanopatterns (the lateral dimension) increased with increased pulse amplitude, on-time, and frequency. The tip velocity effect on the nanopattern line width was also investigated that the line width is decreased with increasing tip velocity. Experimental results were compared with an equivalent electrochemical circuit model representing an ECN technique. The study described here provides important insight for fabricating nanopatterns precisely using electrochemical methods with an AFM cantilever tip.

A tensile test procedure that accommodates specimens with gage section 25 μm thick, 70 μm wide and 360 μm long was developed and demonstrated. The instrumentation and technique were adapted from those previously developed and used to test thin films, by increasing both the force capacity of the load cell and the stiffness of the pull rod. Specimens with bow-tie geometry were fabricated by photolithography from nominally 25 μm thick full hard stainless steel 302 foil. A silicon test frame fabricated by bulk micromachining techniques included tapered grips in the form of recesses in its top surface that accepted and retained the specimen grip sections. One grip was on the fixed outer portion of the frame. The other grip was on a plate suspended in the center of the frame by long slender silicon beams. Force was imposed on this plate by pin loading. The force was measured by use of a custom load cell. The displacement was measured by sub-pixel digital image correlation to surface features on the two ends of the gage section, applied to images with a resolution of approximately 0.8 μm per pixel. Yield and ultimate strengths and elongation values consistent with vendor-provided information were obtained. The values of Young’s modulus were scattered but within the range of expected behavior for the specimen material.

Oxide dispersion strengthened austenitic stainless steel (ODS316), which is based on advanced SUS316 steel, has been developed by mechanically alloying and hot extrusion. Hafnium and titanium were added to make a fine distribution of oxide particles. The stability of oxide particles dispersed in ODS316 under irradiation was evaluated after 250 keV Fe+ irradiation up to high doses at 500 °C. TEM observation and EDS analysis indicated that fine complex oxide particles with Y, Hf and Ti were mainly dispersed in the matrix. There are no significant changes in the distribution and the size of oxide particles after irradiation. It was also revealed that the constitution ratio of Ti in complex oxide appeared to be decreased after irradiation. This diffuse-out of Ti during irradiation could be explained by the difference in oxide formation energy among alloying elements.

Atom probe tomography is used to investigate the clustering of Y-Ti-O in a 14%Cr-2%W-0.3%Ti & 0.3% Y2O3 ODS steel. The clusters in the consolidated material are compared to clusters observed in the powder prior to consolidation. A higher density of smaller clusters is observed in the powder, and the clusters are found to contain more O and less Y.

The mechanism of damage production in solids during irradiation is of great practical interest in nuclear technology. The need to increase the life time of current nuclear plants as well as extreme conditions (high temperature and high neutron flux) in new generations of nuclear plants leads to have a deep insight into radiation damage in solids. In fact, the slowing down of particles in solids leads to a non homogeneous distribution of defects in solids, giving rise to complex microstructures with unusual properties. Numerous experiments, Molecular Dynamics (MD) and Monte Carlo (MC) simulations have clearly shown that highly damaged areas called displacement cascades are produced by neutron or impinging ions in solids. It is now clearly established that the number and the distribution of these subcascades dictate the long term evolution of the microstructure under irradiation. In this work, we present a new model to calculate the mean number of displacement cascades produced in a mono-atomic solid by an incident particle within the Binary Collision Approximation framework (BCA) taking into account all information extracted from MD simulations. To reach such a goal, the notion of subcascade threshold energy is introduced and discussed on some examples. Within this formalism, we exhibit a new way to estimate the number of defects created in a displacement cascade integrating results of MD simulations of cascades.

Surface modification effects on patterned surface with gas cluster ion beam (GCIB) were studied by observation with a cross-sectional transmission electron microscope in order to use it for planarization of patterned media such as discrete track media (DTM) or bit-patterned media (BPM) for future hard disk drives. As a model structure of patterned media, line-and-space or bit patterns were fabricated on Si substrates, and subsequently amorphous carbon films were deposited on them. After Ar-GCIB irradiations on amorphous carbon, it was shown that GCIB preferentially removed bumps or crest on the surface of amorphous carbon at normal incidence. The required thickness for planarization was close to the initial peak-to-valley. At an incident angle of 57°, line-and-space patterns became sharp-pointed shape. On the contrary, line-and-space patterns were planarized without tiny asperity formation at 77°. These results indicate that quite effective planarization of patterned surface is possible using GCIB at normal or glancing angle irradiation.

Low-pressure MOCVD, with tris(2,4-pentanedionato)aluminum(III) as the precursor, was used in the present investigation to coat alumina on to cemented carbide cutting tools. To evaluate the MOCVD process, the efficiency in cutting operations of MOCVD-coated tools was compared with that of tools coated using the industry-standard CVD process.

Three multilayer cemented carbide cutting tool inserts, viz., TiN/TiC/WC, CVD-coated Al2O3 on TiN/TiC/WC, and MOCVD-coated Al2O3 on TiN/TiC/WC, were compared in the dry turning of mild steel. Turning tests were conducted for cutting speeds ranging from 14 to 47 m/min, for a depth of cut from 0.25 to 1 mm, at the constant feed rate of 0.2 mm/min. The axial, tangential, and radial forces were measured using a lathe tool dynamometer for different cutting parameters, and the machined work pieces were tested for surface roughness. The results indicate that, in most of the cases examined, the MOCVD-coated inserts produced a smoother surface finish, while requiring lower cutting forces, indicating that MOCVD produces the best-performing insert, followed by the CVD-coated one. The superior performance of MOCVD-alumina is attributed to the co-deposition of carbon with the oxide, due to the very nature of the precursor used, leading to enhanced mechanical properties for cutting applications in harsh environment.

The radiation response of nano-sized tantalate pyrochlores, KxLnyTa2O7-v (Ln = Gd, Y, and Lu) with average grain sizes of ~ 10 nm was investigated using 1 MeV Kr2+ ion beam irradiations. EDS measurements and XRD refinement reveal that the Y3+ and Lu3+-doped samples consist of two pyrochlore phases as K0.8YTa2O6.9/K0.4Y0.8Ta2O6.4 and KLuTa2O7/K0.4Lu0.8Ta2O6.4 respectively; whereas a single phase of K0.8GdTa2O7 only exists in the Gd3+-doped tantalate pyrochlore. In situ TEM observation confirms ion beam-induced amorphization occurring in all of the nano-sized KxLnyTa2O7-v. At elevated temperatures, both K0.8GdTa2O7 and K0.8YTa2O6.9/K0.4Y0.8Ta2O6.4 exhibit higher radiation tolerance than KLuTa2O7/K0.4Lu0.8Ta2O6.4, and the critical temperatures of K0.8GdTa2O7 and K0.8YTa2O6.9/K0.4Y0.8Ta2O6.4 are estimated to be 1167 ± 41 K and 1165 ± 34 K, respectively, lower than that of KLuTa2O7/K0.4Lu0.8Ta2O6.4 (~ 1291 K). The K0.8GdTa2O7, K0.8YTa2O6.9 and KLuTa2O7 phases have less structural deviation from the parent fluorite structure and thus may be responsible for the overall radiation tolerance. The high K+ occupancy at pyrochlore A sites in KLuTa2O7 is believed to contribute to the decrease of radiation tolerance, consistent with the large ionic radius ratio of K+/Ta5+. These results highlight that the radiation tolerance of nanostructured materials is highly compositional dependent, and nano-sized tantalate pyrochlores are sensitive to radiation damage.