Materials for current and future nuclear applications: processing, characterization, performance

Resume : The migration, segregation and retention behavior of fission products, among which iodine is one, are important for performance and safety throughout the irradiation cycle of nuclear fuel and waste disposal. This study examines the speciation as well as chemical interaction of iodine with zirconia (zirconium oxide), which may form by the contact of UO2 fuel with zircaloy cladding at its inner surface, and acts as the first migration barrier towards the fission products release pathway. Cubic zirconia (CZ) samples have been prepared by iodine ions implantation. Grazing incidence extended X-ray absorption fine structure (EXAFS) spectroscopy measurements at the Zr and I K-edges are applied to analyze the first-shell local structural environment of Zr and I atoms in the samples before and after thermal annealing treatments. The results show that the iodine irradiation introduces structural vacancies both in the zirconium and oxygen sub-lattices of CZ. The coordination number for the first nearest oxygen neighbors of Zr is decreased after implantation which appears under the impacts of the implanted ions, and the first-shell iodine speciation in CZ is modified in the annealed sample compared to that in the as-implanted state. Atomistic simulations using density functional theory (DFT) based models have been performed to get a deeper insight into the dopant iodine distribution and defect-impurity interactions, and the chemical affinity of iodine in CZ microstructure. The retention behavior of iodine as a fission product in zirconia is discussed, and potential of EXAFS measurements as a local probe of atomic-scale structural modifications induced by iodine implantation in CZ is addressed.

Resume : A possible solution to the nuclear waste problem is a long-term geological disposal of the waste. Ceramics such as monazites (LnPO4) and pyrochlores (A2B2O7) are considered as potential immobilization matrices for selected radionuclides including Pu. The assessment of the long-term behavior of ceramic waste forms requires good understanding of the thermodynamic properties of these materials, which is often challenging for experimental methods and could be provided by atomistic simulations. We thus perform systematic ab initio investigations of structural, thermodynamic, elastic and energetic properties of these ceramics and the corresponding solid solutions involving actinides. We calculated the Margules interaction parameters that describe the excess properties of monazite-type solid solutions and correlated them with the elastic strain induced by the incorporation of different cations. This indicates that the measured or computed elastic moduli together with the structural information could be sufficient for estimation of the thermodynamic excess properties of these solid solutions. The common feature of certain pyrochlores having high radiation damage resistance is the radiation induced disordering of the cationic and anionic sublattices and transformation to the defect fluorite-type phase. In order to better understand the defects accumulation process in these ceramics we calculated the formation energies of cation anti-site and anion Frenkel pair defects and the combination of them as well as the barriers for oxygen diffusion in series of pyrochlores. Our results indicate that the ease of the formation of anion Frenkel pair through oxygen migration is one of the driving forces that lead to the order-disorder transition of selected pyrochlores.

Resume : Alloys based on α-Fe are proposed as structural materials in fusion applications. Iron atoms after neutron irradiation produced He from (n, α) transmutation reactions. The solubility of helium in all iron based alloys is very low, which is the reason why He atoms have tendency to be trapped at defects such as vacancies, dislocations and grain boundaries (GB). Helium atoms influence strongly the cohesive energy of GB. In these studies, the segregation energy of helium at Σ3 (111) and Σ5 (210) tilt GB is investigated both in Fe dilute alloys and Fe-Cr alloys. Different concentrations and positions of Cr and He are considered. The results show that the presence of helium atoms significantly influences the magnetic properties of the system in the relatively distant neighbourhood. It is observed that the segregation energy of helium atom is influenced by Cr position at the GB. The migration energies of He atom in these systems are also calculated. Density functional theory calculations are performed by using VASP code, with generalized gradient approximation (GGA) of Perdew-Burke-Ernzerhof (PBE) for exchange-correlation. The modelling on the atomistic level presented here will enable to better understand the phenomena of helium embrittlement and formation of helium bubbles in Fe-based alloys.

Resume : This work shows that realistic irradiation-induced phase separation and the resulting microstructures can be obtained via an adapted Phase Field (PF) modelling combined with atomistic Monte Carlo simulations in the pseudo-grand canonical ensemble. The last allow for calculating the equilibrium phase diagram of the silver-copper alloy, chosen as a model of binary systems with large miscibility gap and, for extracting the parameters of the excess free-energy PF functional. Relying on this methodology the equilibrium phase diagram of the alloy is predicted in excellent agreement with its experimental counterpart whereas, under irradiation, the predicted microstructures are functions of the irradiation parameters. Different irradiation conditions trigger the formation of various microstructures consistently presented as a non-equilibrium “phase diagram” aiming at facilitating the comparison with experimental diffraction patterns.

Resume : Raman scattering is well-known to be a precious tool for investigating irradiation damage in materials. A review will be given of the efficiency of the method on actinide oxides : UO2, PuO2 and ThO2. These compounds exhibit a three-line structure of Raman peaks induced by irradiation, whatever the irradiation type : heavy or light ions, or electrons. This set of lines is investigated by an original Raman device allowing in situ characterization during irradiation (for U and Th compounds) . The Th is complicated by photoluminescence processes whatever the excitation line. Nevertheless the three-line structure has been identified by time-resolved Raman experiments which can discriminate Raman from luminescence. Annealing treatments, electron irradiations and positon annihilation spectroscopy experiments give some insight to the origin of these defect lines. Besides this, these studies on model materials help to interpret the Raman spectrum of highly-irradiated materials such as spent nuclear fuels. In a last part, a recent new device for in situ Raman investigations under irradiation will be detailed : a combined T-P-He2+ irradiation-Raman cell, conceived for characterizing the structural behavior of nuclear graphites (used as moderators in UNGG power plants) in working conditions of these plants. The goal is to quantify the release of specific long-lived radioisotopes (14C and 36Cl) simulated by implanted ions (13C and 37 Cl).

Resume : Silicon carbide (SiC) satisfies the key criteria for cladding materials for the Generation IV nuclear reactors (LFR, VHTR and GFR reactors) in terms of corrosion resistance, fission product retention ability, mechanical properties and low neutron cross section absorption (0.10 relative to Zr) [1-5]. However, a major drawback of such materials is their brittleness, thus composites have been fabricated to overcome this behavior. In addition to this, understanding the various interfaces in SiC composites will give the crucial information for TRISO fuel design and simulation. In this research, TiC/SiC composites have been generated by TiC (2 µm) and β-SiC power (1 µm) with different molar ratio (50 and 30 % mole) using spark plasma sintering (SPS) from 1600-2100 ˚C, 80 MPa and 15 minutes for fabrication. The density of both composites reach to 99 % of theoretical density at 2100 ˚C without sintering aids. The Vickers hardness of the composites were significantly enhanced to ~ 31GPa at 50 % mole of SiC. Their fracture toughness calculated from indentation method are 7.94 and 7.86 MPa m1/2, respectively which are significantly higher than bulk TiC and SiC. Additionally, their thermal diffusivities lie between bulk SiC and TiC and are inversely proportional to temperatures (50 – 500 ˚C). Scanning Electron Microscopy (SEM) micrographs show different morphologies of TiC/SiC interfaces indicating potential interesting interfaces between two compounds. Thus, further investigation will focus on examining these interfaces using TEM (Transmission Electron microscopy) with more detailed examination of the mechanical and thermal properties. Keyword: Silicon carbide, carbide composites, interfaces Reference 1. Yvon, P. and F. Carré, Structural materials challenges for advanced reactor systems. Journal of Nuclear Materials, 2009. 385(2): p. 217-222. 2. Bloom, E.E., The challenge of developing structural materials for fusion power systems. Journal of Nuclear Materials, 1998. 258–263, Part 1(0): p. 7-17. 3. Heinisch, H.L., et al., Displacement damage in silicon carbide irradiated in fission reactors. Journal of Nuclear Materials, 2004. 327(2–3): p. 175-181. 4. Lee, Y. and M.S. Kazimi, A structural model for multi-layered ceramic cylinders and its application to silicon carbide cladding of light water reactor fuel. Journal of Nuclear Materials, 2015. 458(0): p. 87-105. 5. Zinkle, S.J., Advanced materials for fusion technology. Fusion Engineering and Design, 2005. 74(1–4): p. 31-40. 6. Ferraris, M., et al., Joining of machined SiC/SiC composites for thermonuclear fusion reactors. Journal of Nuclear Materials, 2008. 375(3): p. 410-415. 7. Henager Jr, C.H., et al., Coatings and joining for SiC and SiC-composites for nuclear energy systems. Journal of Nuclear Materials, 2007. 367–370, Part B (0): p. 1139-1143. 8. Katoh, Y., et al., Radiation effects in SiC for nuclear structural applications. Current Opinion in Solid State and Materials Science, 2012. 16(3): p. 143-152.

Resume : Detailed studies investigating the effects of 800 keV Ar and 1 MeV Au ion irradiation on the microstructure and properties of nanostructured ZrC and ZrN thin films were performed. The films were grown on Si substrates using the pulsed laser deposition technique. A low residual vacuum prior to deposition and a high repetition rate pulsed laser allowed us to obtain films having small size crystalline grains and a low level of oxygen contamination. The films were grown on Si substrates from room temperature up to 500 ºC. Atomic force microscopy images showed that films were very smooth, with rms values below 1 nm. The smooth and flat surface allowed for X-ray reflectivity (XRR), X-ray diffuse reflectivity and grazing incidence X-ray diffraction (GIXRD) investigations to be performed at low incidence angles. Also, nanoindentation studies could be performed using very small loads, which will results in penetration depths smaller than 100 nm, confining the investigations to the area of the films affected by the ion irradiation, without any contribution from the substrate. The XRR and GIXRD results indicated that after irradiation there is a small density decrease, accompanied by an increase in the lattice parameters. More intriguing, a strong change in the texture of the films and a significant increase in the grain size were also observed. While the underlying Si substrate was completely amorphized for fluences around 1015 ions/cm2, the ZrC and ZrN films retained their crystalline structure without the appearance of any large scale defects.

Resume : Radiation tolerance of complex oxide like spinels and pyrochlores and more generally fluorite structures still remains an open question. By analogy with works performed on metals and alloys, numerous authors discussed the stability of these materials in term of point defects kinetics and agglomeration. Under these assumption validated by MD calculations, they considered the motion of cations and anion irrespectively to their charge valence, i.e their local environment. The application of green chemistry offers the unique opportunity to mimics radiation damage in term of grain size on a simple pyroclhore La2Zr2O7. Based on this analogy, we showed that the defect fluorite structures resulting from radiation damage in these materials can be understood as resulting of a simple reconstruction process without any order/disorder phase transition. This work may be extended to explain the radiation tolerance of other complex oxides like spinels and may be applied to discuss amorphization in these materials.

Resume : Fe-Cr-based alloys are intensely studied to understand the fundamental mechanisms whereby irradiation modifies the nanostructure and the chemical element distribution in ferritic/martensitic steels, that are planned to be used to build in-core components in future GenIV fission energy reactors and tritium breeder blankets in fusion energy systems. Well-known microstructural effects of irradiation in these alloys are the appearance of high densities of dislocation loops and, at sufficiently high dose, voids. In the thermodynamic range of stability (>~9-10%Cr, depending on temperature), a/a' demixing occurs, while recently the formation of solute clusters that do not correspond to thermodynamic phases and probably segregate at small dislocation loops have been identified. Based on the existing experimental evidence and on recent theoretical advances, physical models are being developed that intend to describe, and possibly predict, both the defect production and evolution under irradiation and the microchemical changes (precipitation, segregation, ...) that occur in Fe-Cr-based alloys during irradiation. In this talk, results of post-irradiation examination of irradiated Fe-Cr alloys are briefly reviewed to identify the issues that models should address and recent advances in the direction of developing such models, mainly based on the object kinetic Monte Carlo technique, are reported, showing the results hitherto obtained and discussing the perspectives. The work reported is largely based on the work done in the FP7 projects GETMAT and MatISSE, both contributing to the EERA Joint Programme on Nuclear Materials.

Resume : An efficient dispersion of nano-oxides in ODS steels was achieved by employing high efficient attrition milling. A combined wet and dry milling process of fine ceramic and steel particles is proposed. Spark Plasma Sintering (SPS) was applied to realize nanostructured steel compacts. Grains with 100 nm mean size have been observed by SEM in sintered austenitic ODS. In comparison the sintered martensitic dry milled and maretnsitic dry and combined milled ODS microstructure consisted of grain size with 100-300 nm in each case. A brittle behavior is shown in all of the cases. The martensitic ODS is two times harder than the austenitic ODS. The bending strength high as 1806.7 MPa was found for the martensitic ODS, whereas 1210.8 MPa was determined for the austenitic ODS. The combined milling assured higher strength and hardness compared to dry milling.

Resume : Molten salt reactors are Generation IV nuclear reactors where the fuel is dissolved in an alkali chloride or fluoride salt at temperatures in excess of 500°C [1]. The structural materials used must possess high strength at such temperatures and resist chemical attack in high neutron and gamma fluxes [2]. The MAX phases, a class of layered ternary carbide and nitride materials with a mixture of ceramic and metallic properties, are strong at high temperatures and resistant to chemical attack and radiation damage [3]. However, to date their resistance to reaction in alkali halides has undergone limited research. To this end, bulk Ti3AlC2 has been synthesised and exposed to LiCl-KCl eutectic (LKE) in order to investigate its resistance to chemical attack. Fully dense samples with greater than 90% Ti3AlC2 and less than 10% TiC have been consistently produced using mechanical alloying followed by spark plasma sintering. Samples have undergone long-term exposure to LKE under an argon atmosphere followed by characterisation via SEM-EDX, Raman spectroscopy and glancing-angle XRD. In addition, a variety of electrochemical experiments have been performed in-situ: the open-circuit potential has been measured and cyclic voltammetry has been used to investigate the corrosion mechanism. 1 J. Serp, M. Allibert, O. O. Beneš, S. Delpech, O. Feynberg, V. Ghetta, D. Heuer, D. Holcomb, V. Ignatiev, J. L. Kloosterman, L. Luzzi, E. Merle-Lucotte, J. Uhlíř, R. Yoshioka and D. Zhimin, Prog. Nucl. Energ., 2014, 77, 308–319. 2 V. Ignatiev and A. Surenkov, in Comprehensive Nuclear Materials, Elsevier Inc., Karlsruhe, Germany, 2012, pp. 221–250. 3 K. R. Whittle, M. G. Blackford, R. D. Aughterson, S. Moricca, G. R. Lumpkin, D. P. Riley and N. J. Zaluzec, Acta Mater., 2010, 58, 4362–4368.

Resume : Tungsten is being considered for fusion applications as one of the major materials in plasma facing components. Indeed its high melting point, good thermal conductivity and low sputtering yield, which minimizes the impurity generation, are highly desired properties. However, there is one main disadvantage in the tungsten use, which is very hard to overcome: the ductile-to-brittle transition temperature (DBTT) is too high. A possible solution for this weakness is the production of W-Ta composites since tantalum has a high ductility and toughness relative to W. Therefore, dispersions of ductile Ta fibers in a W matrix have been proposed as a novel approach for the development of suitable plasma facing materials [1,2]. In this study, W-Ta composites were produced from W powder and 10 at.% Ta fibers and consolidated by pulse plasma compaction. Implantation was carried out at room temperature with He+ or D+ or sequentially with He+ and D+ using ion beams with fluences in the 1020-1021 at/m2 range. Microstructural changes and deuterium retention in the W-Ta composites after He+ and/or D+ implantation were investigated by scanning and transmission electron microscopy coupled with energy dispersive X-ray spectroscopy and focused ion beam, Rutherford backscattering spectrometry, nuclear reaction analysis and X-ray diffraction. The composite materials consisted of a nanostructured W matrix with dispersed Ta fibers presenting Ta2O5 at the interfacial regions. Blistering was observed after He+ implantation with a stronger effect in Ta2O5 than in Ta or W and was worsened with prior D+ implantation. The blister cavities in Ta2O5 exhibited a nanometer-sized fuzz structure. Nuclear reaction analysis showed that D retention increased with He+ pre-implantation, which justifies the shifts of the tungsten X-ray diffraction peaks after He+ implantation and reflect the strain associated with the presence of interstitial atoms in the W lattice. Transmission electron microscopy observations revealed the presence of dislocations in the W and Ta metallic phases after the sequential implantation with He+ and D+ ions, while a lower density of defects was detected in Ta2O5 due to an apparent shielding effect of the blisters. [1] V. Livramento, D. Nunes, J.B. Correia, P.A. Carvalho, R. Mateus, K. Hanada, N.Shohoji, H. Fernandes, C. Silva, E. Alves, Tungsten–tantalum composites for plasma facing components, in: Materials for Energy 2010, ENMAT-2010, 4–8July 2010, Karlsruhe, Germany. [2] M. Dias, R. Mateus, N. Catarino, N. Franco, D. Nunes, J.B. Correia, P.A. Carvalho, K. Hanada, C. Sârbu, E. Alves, Journal of Nuclear materials 442, Issue 1-3, (2013), 69-74.

Resume : The effects of grain boundary engineering (GBE), utilizing bulk thermomechanical processing with iterative cycles of 10% cold work and strain annealing, as well as a novel surface GBE method (SGBE) involving ultrasonic nanocrystal surface modification plus strain annealing to modify the near-surface microstructure (~250 m), on corrosion and stress corrosion cracking (SCC) behavior of alloy 600 (Ni-15Cr-9Fe) was studied. The associated microstructural and cracking mechanisms were elucidated using transmission (TEM) and scanning electron microscopy (SEM), coupled with precession electron diffraction (PED) and electron back scatter diffraction (EBSD) mapping. Such bulk and surface processing resulted in increased fraction of special grain boundaries and triple junctions whilst decreasing the connectivity of random high angle grain boundaries. A disrupted random grain boundary network and large fraction of coincident site lattice (CSL) boundaries (Σ3-Σ27) reduced the propensity to sensitization, i.e. carbide precipitation and depletion of Cr at grain boundaries. This GBE Alloy 600 also showed higher intergranular corrosion resistance. Slow strain rate tests in tetrathionate solution at room temperature show GBE lowered susceptibility to intergranular SCC. The improvements with SGBE were comparable with those from bulk GBE. To better understand the improvements in corrosion and SCC resistance, EBSD of regions around cracks was used to analyze the interactions between cracks and various types of grain boundaries and triple junctions. Detailed analysis showed that cracks were arrested at J1(1-CSL) and J2 (2-CSL) type of triple junctions. The probability for crack arrest at special boundaries and triple junctions, calculated using percolation models, was found to have increased after GBE, which also explains the increase in resistance to corrosion and SCC in grain boundary engineered Alloy 600. A clear correlation and mechanistic understanding relating grain boundary character, sensitization, carbide precipitation and susceptibility to corrosion and stress corrosion cracking was established. The implications of GBE on radiation induced segregation behavior and hence on potential effects on IASCC will also be discussed.

Resume : In this study, CeO2, a surrogate for nuclear fuel ceramics, was irradiated with 200 MeV Au ions at oblique incidence. Observation of as-irradiated samples by transmission electron microscope (TEM) shows that hillocks are created not only at the wide surfaces, but also at the side edge of the thin samples. Since the hillocks created at the side edge can be imaged without overlapping of matrix image, their shape and crystallographic features can be revealed. From the images of hillocks created at the side edge, many of the hillocks are found to be spherical. We present an experimental evidence that hillocks created for CeO2 irradiated with swift heavy ions have a crystal structure whose lattice spacing and orientation coincide with those of the matrix. The present method of observing hillocks can be a complementary technique to AFM(Atomic Force Microscopy), which has been mainly used so far for characterizing hillocks. The present method is also applied for observing hillocks created for CaF2, which has the same crystal structure with CeO2. Although hillock-like shape has been reported so far for hillocks of CaF2, the present study revealed that not only hillock-like shaped but also polyhedron shaped hillocks are created by irradiation with swift heavy ions. The polyhedron shaped hillocks have faceted surfaces reflecting the crystal lattice structure.

Resume : Slow cooling of phosphate melt used for liquid nuclear waste solidification in glass yields glass-crystalline material. Partial crystallization at melt solidification results in elemental partitioning among crystalline phase and glass: Al, Cr, Fe are concentrated in crystalline constituent while Na, P, Ca, Ni, La, U enters predominantly residual glass. Glass dissolution rate and leach rate of La and U as RE and actinide surrogates exceed those from glass with specified composition by three orders of magnitude. The work was granted from Russian Science Foundation (14-13-00615).

Resume : A fundamental understanding of the actinides incorporation process in envisioned nuclear waste forms, such as monazite ceramics, is required for a reliable prediction of the long-term stability of ceramic materials for safe nuclear disposal. The present study provides structural insights into the formation of monazite solid solutions by incorporation of Cm3+ or Pu3+ and verifies previous results on surrogate materials, where Eu and Gd served as inactive analogues for trivalent actinides. The incorporation of Cm3+ (50 ppm) in La1-xGdxPO4 (x = 0, 0.2, 0.5, 0.8, 1) with monazite structure has been investigated using TRLFS and XAS methods. The spectroscopic results confirm Cm3+ substitution for the host cation sites for all compositions. The Cm3+ excitation spectra of the end-members obtained with the TRLFS method show four well-resolved peaks (4-fold splitting of the Cm3+ ground state) indicating a well-defined crystalline environment for the incorporated Cm3+. The loss of the splitting fine-structure and the broadening of the excitation peaks for the solid solutions are due to a decrease of the short-range order. The Debye-Waller factor for the O coordination shell of Cm, obtained by EXAFS, is higher for La0.5Gd0.5PO4 than for the end members indicating increasing structural disordering for the solid solution. Pure-phase La1-xPuxPO4-monazites with x=0.01-0.15 were synthesized by a solid state method. A linear dependence of the lattice parameters on composition confirms solid solution formation. XAS measurements verified the +III valence state of plutonium in the monazite solid solutions. XRD-patterns of the compounds with x=0.25, 0.50 revealed the formation of two phases: (La,Pu)PO4-monazite and a cubic phase (PuO2/Pu2O3).

Resume : During the extraction processing of spent nuclear fuel significance is to identify the rate of oxidative processes, radiation and radiation-thermal effects influencing on it. Depending on the oxidative processes conditions in extraction systems may flow with relatively constant speeds of liquid-phase and gaseous products formation or (in autocatalytic mode) with self-heating of the mixture and sharply accelerated gassing can occur. We carried out series of studies on thermal interaction of non-irradiated and irradiated double -phase systems. The organic phase - 30% tri-n-butyl phosphate (TBP) in ISOPAR-M was saturated by 4, 8 and 12 mole/l HNO3 at temperatures ranging from 70оС to 110оС. All samples were irradiated with an electron accelerator UELV-10-10-C-70 at a dose rate of 10 kGy/h to integral doses of 0,5; 1 and 2 MGy. In unirradiated double -phase systems «TBP in ISOPAR-M - aqueous HNO3 solution», significant outgassing having been observed even at concentrations of nitric acid in the aqueous phase of less than 8 mole/l and temperatures below 90°C. The maximum gassing rate unsignificantly increases with increasing nitric acid concentration, but (at times) – less than with increasing temperature. The volume of the released gases are significantly increased with increasing temperature rising and increase of the nitric acid concentration, reaching at СНNO3 = 12.0 mole/l value of 36 lg/ll, which is by 4 times higher than the gassing of the system with a diluent C13 under similar conditions. This study was financially supported by the Ministry of Education and Science of the Russian Federation in the framework of the Agreement with IPCE RAS, the unique identifier RFMEFI60414X0153.

Resume : Nagelschmidtite, Ca7P2Si2O16, is an end-member of continuous solid solution Ca2SiO4 – Ca3(PO4)2х2Ca2SiO4 within the pseudo-double system Ca3(PO4)2 – Ca2SiO4 (whitlockite – larnite). This phase is capable to wide isomorphic exchanges in Ca, P and Si sites including An, RE, Na, Al in Ca site. It was found earlier in metallurgical slags and geological formations. We revealed negelschmidtite-type phase in vitrified phosphorus-bearing radioactive incinerator slags. The materials were glass-crystalline and composed of nano-sized nagelschmidtite crystals distributed in vitreous matrix phase. Average chemical compositions of the largest (few microns) crystals were recalculated to formula Na1.21K1.05Ca2.22Al2.02Fe0.46Si2.69P1.26U0.08O15.76. Significant oxygen misbalance suggests higher than U(IV) oxidation state for uranium – U(V) or U(VI). Capability of nagelschmidtite to be crystallized from melt makes it promising phase at using a melting route to nuclear waste forms including cold crucible induction melting and self-propagating high-temperature synthesis. The work was granted by Program of Presidium of the RAS No. IV.5.8.

Resume : An important task to develop the fractionation technology of high-active waste of UNEX process is to ensure explosion-safety - reducing the risk of uncontrolled chemical reactions occurring during the interaction of the organic extractant with concentrated nitric acid. To assess the ability of uncontrolled reactions during the technological processes of fractionation, we estimated the kinetics of the interaction of the extractant main component (heavy diluent trimethylphenylsulfone, FS-13) with 14 mole/l nitric acid. Therefore, we carried out experiments with heating the double-phase system with a ratio of phase’s diluent: nitric acid as 1:2 in an autoclave at temperatures of air thermostat 170 and 2000С. Using special software TSS (CJSC "Himinform"), by the methods of mathematical optimization the parameters have been estimated for kinetic equation, allowing a satisfactory description of the experimental data. TMR (time maximum rate) was estimated based on the kinetics evaluation of the interaction in the most conservative assumption, namely the absence of heat exchange of the reaction mixture with the environment. The results show that TMR is reduced exponentially with the increase of the process temperature and reaches the lowest value for about 5 hours at the boiling temperature of concentrated nitric acid (~1150С). Registered temperatures are significantly higher than regulating temperatures of the extraction processes, based on which we can make a preliminary conclusion, that the risk of thermal explosion is low, because it is determined as such simultaneous violations of the process as: heating for more than 500С and equipment hermetization. This study was financially supported by the Russian Science Fund (project 16-19-00191).

Resume : Helium embrittlement poses a major challenge for the application of ferritic/martensitic steels in future nuclear devices. The strategy for an effective suppression of He embrittlement is to reduce the mean free diffusion path of He in steel. This can be achieved by introducing nanoparticles into the steel and thus increasing the specific interface area available for He trapping. The study is aimed at clarifying the role of oxide nanoparticles on the formation of He bubbles and irradiation-induced hardening for the case of He-ion irradiations performed in the He-ion microscope (HIM). The materials are the reduced-activation ferritic/ martensitic 9%Cr steel Eurofer97 and an oxide-dispersion strengthened (ODS) variant produced by means of hot isostatic pressing of Eurofer steel powder after mechanically alloying with 0.3 mass% yttria powder. In a first run, the samples were irradiated with He ions up to a very high exposure of 2e18 ions/cm2. Movies taken in the HIM indicate blistering starting at exposures between 1e17 ions/cm2 and 1e18 ions/cm2. In a second run, regularly arranged square areas were irradiated up to different He-ion exposures in the range from 1e14 ions/cm2 to 1e17 ions/cm2. Nanoindentation was applied to study irradiation-induced hardening as a function of exposure. Eurofer97 and ODS-Eurofer will be critically compared with respect to both hardening and blistering.

Resume : Because of high radiation resistance and wide band gap, Al2O3 (corundum) is widely used as an effective detector of ionizing radiation. Its potential applications include also components of breeder blanket and diagnostic windows. Radiation-induced changes in the structural and optical properties of corundum are mainly associated with primary Frenkel defects: neutral and charged interstitial oxygen atoms Oi, as well as oxygen vacancies Vo (F-type color centers). Unlike the latter, the former are not well studied yet. In this study, we present results of periodic ab initio simulations on basic properties and mobility of the charged oxygen interstitials using the CRYSTAL14 computer code. The defect geometries and migration energies, Mulliken atomic charges and the electron density distributions for neutral and charged impurities are compared. It has been shown that the single-charged interstitial ion forms a dumbbell with a regular oxygen ion shifted from one of the nearest lattice sites (which distance is 1.87 Å) preserving site symmetry. Charged oxygen interstitials as well as neutral form the bonds with regular Al ions in corundum lattice rather than occupy centers of octahedron consisting of six nearest oxygen ions as one could intuitively expect. The calculated migration energies are compared with available experimental data.

Resume : Due to their excellent thermo-mechanical properties carbon-based materials are promising candidates for application in extreme radiation environments, like nuclear reactors, beam protection elements and targets in accelerator facilities. Structural changes and radiation damage-induced thermal conductivity degradation of isotropic polycrystalline graphite and flexible graphite were tested by irradiation with 4.8 and 5.9 MeV/u C, Ca, Au and U ions at the UNILAC, GSI, Darmstadt. Significant decrease of the thermal diffusivity, as measured by laser flash analysis, occurs at high fluence, when the entire sample is covered by ion tracks. For Au and U ions at fluences higher than 1e13 ions/cm2, the thermal diffusvity value of polycrystalline graphite appoaches the value for glassy carbon. The different trends for light (C and Ca) and heavy (Au, U) ion irradiations indicate the existence of a critical energy loss threshold of significant disordering in graphite. Additionally, Raman spectra of the irradiated samples were analyzed to estimate structural changes.

Resume : The stability of heterophase interfaces between metal systems, their kinetic, structural, and thermomechanical properties are a matter of concern for high demanding applications involved in the development of technological coatings for the first wall materials in fusion reactors and their prototypes (ITER). We would like to discuss preliminary results of numerical simulations and X-ray experiments on model coatings made of tungsten, in particular the problems related to strain and adhesion of thin or thick metal films on heterophase substrates. This research is supported by a research grant of Investissements d’Avenir of LabEx PALM (ANR-10-LABX-0039-PALM).

Resume : The European Spallation Source (ESS) is a large scale neutron facility build in Lund, Sweden, and it will be the world’s most powerful neutron source. The neutrons are produced in tungsten through a spallation process, which due to its high atomic number has a high neutron production rate. Tungsten is the most critical non-structural material and its integrity during operation is essential and must operate reliably and predictably for the planned lifetime of the target. In order to estimate the target life, reliable data is needed on the mechanical properties of tungsten, in both unirradiated and irradiated conditions. In this study, we evaluate the properties of unirradiated and irradiated tungsten. In order to see the effect of irradiation in tungsten, we compare the microstructure, dislocations and defects in both types. Transmission electron microscopy (TEM) samples were prepared by focused ion beam (FIB) and then the lamellas were thinned by flash electrochemical polishing. In order to obtain quantitative information of the dislocations and defect clusters, they are examined on several low and high magnification images in different areas of the sample.

Resume : Uranium dioxide (UO2) is employed as a fuel material in most nuclear reactors world-wide. Its doping with small amounts of chromium sesquioxide (Cr2O3) is technically applied to obtain larger average grain size after the fuel sintering process which leads to a better retention of fission gases. However, the accurate mechanism of Cr cation incorporation into the UO2 matrix is still not fully identified. In this study, the atomic structure of chromium doped uranium dioxide is investigated using the density functional theory (DFT) approach as well as classical pair potentials as it is implemented in the CP2K code. The obtained theoretical results are compared with those from the experimental x-ray absorption spectroscopy study [1]. The U L3-edge extended x-ray absorption fine structure spectrum is interpreted within the multiple-scattering (MS) theory using the results of the classical and ab initio molecular dynamics simulations, allowing us to validate the accuracy of theoretical models. The Cr K-edge x-ray absorption near edge structure is simulated within the full-multiple-scattering formalism considering substitutional models (Cr at different sites in the host lattice). More complicated environments for the embedding of Cr atoms into the UO2 matrix, e.g. dislocations, will be discussed. [1] D. Bocharov, et al., J. Phys.: Conf. Ser. 712 (2016) 012091:1-4.

Resume : Thin films of indium zinc oxide with various In/(In+Zn) values were deposited by the pulsed laser deposition technique on Si and glass substrates at room temperature. Grazing incidence X-ray diffraction investigations showed that films were amorphous, regardless of their composition. The grown films were irradiated by 1 MeV Au ions, gamma, X-ray and UV radiation to investigate the effects on the films structure and properties. The surface morphology of the deposited films, investigated by atomic force microscopy, was very smooth, with rms values below 1 nm, allowing for the use of surface sensitive characterization techniques such as X-ray reflectivity, X-ray diffuse scattering, grazing incidence X-ray diffraction, X-ray photoelectron spectroscopy, or nanoindentation that all possess depth resolutions of the order of few nm. In addition we used optical reflectometry and photoluminescence to characterize the optical properties of the irradiated films. The results showed that these amorphous films could tolerate a high level of radiation without adverse effects upon their structure, stoichiometry, optical and electrical properties.

Resume : The objective of this talk is to demonstrate the validity of transmission electron microscopic (TEM) in situ micropillar compression tests for quantitative assessment of yield stress and elastic modulus in irradiated oxide dispersion strengthened (ODS) alloys. Materials testing for future advanced nuclear reactor concepts is increasingly calling upon charged particle irradiations as a surrogate for neutron irradiation. Ion irradiations can deliver a damage rate up to three orders of magnitude higher than that of neutron irradiations, with little or no residual radioactivity, and at fractions of the cost. However, evaluating the effects of ion irradiation on mechanical performance remains a challenge, due to the shallow damage layer of a few micrometers, at the surface. Thus, small-scale mechanical testing techniques have gained prominence for evaluating the mechanical performance of thin irradiation damage layers. Transmission electron microscopic (TEM) in situ mechanical testing holds even greater promise, however, as it is able to provide quantitative load-displacement data alongside TEM-resolution video of fundamental deformation processes in the material. This study focuses on a model Fe-9%Cr ODS alloy irradiated with 5.0 MeV Fe++ ions to 3 and 100 dpa (displacements per atom) at 500°C. The unirradiated condition is also studied as a control. We use a Hysitron PI95 TEM in situ mechanical testing holder fitted with a diamond flat punch tip. Compression tests are conducted on micropillars fabricated from the ODS alloy using focused ion beam milling. We observe a size effect only for pillars having minimum dimension <100 nm. For pillars above this size threshold, TEM in situ compression tests accurately measure yield stress values for both the irradiated and unirradiated specimens. The measured changes in yield stress with irradiation are consistent with microstructure-based predictions using the dispersed barrier hardening model. Elastic modulus values, when corrected for the amount of deflection and deformation occurring in the base material below each pillar, fall in agreement with expected values.

Resume : Here we look at radiation damage in nuclear materials from the perspectives provided by atomistic simulations, experimental studies, and natural systems (e.g., actinide minerals). These approaches provide valuable information on radiation effects over short (e.g., picosecond), intermediate (years), and very long (106-109 years) periods of time. Atomistic simulations provide valuable input to the understanding of processes that are difficult to follow by experiment due to the short time scales involved. Molecular Dynamics (MD) approaches can give insight to the formation and recovery of thermal spikes and collision cascades as a function of temperature, including the types of defects produced. Furthermore, Density Functional Theory (DFT) methods provide estimations of the energetics of defect formation and migration in complex systems. These studies allow an immediate assessment of the near-instantaneous damage mechanisms and relative resistance to amorphization in materials. Based on ion irradiation experiments (relatively high dose rate) wherein temperature is a variable, materials are identified that either 1) remain crystalline to very high ion fluence at very low temperature, 2) can be rendered amorphous at very low temperature but exhibit a critical temperature for complete recovery at moderate temperature, or 3) become amorphous at low dose and have a very high critical temperature for recovery of damage.

Resume : The modelling of the behavior of uranium oxides under oxidation conditions is important for applications and for the protection of the environment. In order to assess the safety of nuclear fuels, it is desirable to predict and to establish the characteristic elementary mechanisms that are likely to take place, and it is essential to know the relative stabilities of the compounds that may form under the relevant conditions. Also, those systems represent a large class of compounds with high radiation tolerance, somewhat related to their peculiar ability to accommodate a variety of defects and to form nonstoichiometric compounds with a large homogeneity range. For those reasons, the study of binary U oxides is of fundamental interest: they provide insight into the U element’s basic chemical properties, including oxidation state stabilities, coordination preferences, and bonding. Those analyses are particularly rich and interesting for the U compounds that exhibits an intriguing structural variety, with numerous phases and solid solutions having U/O ratios between 1:2 and 1:3.

Resume : Americium will contribute to a large extend to the radiotoxicity of spent fuel during the first centuries of storage/disposal. In the scenario of spent fuel reprocessing and minor actinide separation, the americium could be transmuted in fast reactors in so-called minor actinide bearing blankets (MABB). It is also envisaged to study americium dioxide for its potential use in radioisotopic thermal generators (RTG) to be integrated in batteries for space probes. A first transmutation experiment of such type of fuel took place in the mid-eighties in the SUPERFACT experiment; UO2 based-fuels with different content of americium (and other actinides) were irradiated in the Phénix fast reactor. Irradiated Superfact fuels have been recently observed by transmission electron microscopy (TEM) showing a rather remarkable behaviour against the formation of radiation damage. Archive materials from the Superfact fuel with different americium content (2 at% and 20 at%) have also been investigated, as well as material with higher amount of americium (up to 50 at%). Electron energy loss spectroscopy (EELS) and EXAFS studies were performed with the aim of determining the evolution of the damage in these materials as a function of cumulated alpha-dose and composition (U:Am ratio). Some conclusions on the use of such materials will be drawn based on their resistance to damage formation.

Resume : Phenomena such as diffusion, accumulation and fission product release have a strong impact on the UO2 fuel properties and influence the heat and mass transport. Strontium is a reactive fission product. Its stability in uranium dioxide depends on the fuel oxygen potential and the chemical affinity with other fission products in irradiated UO2 matrix. In this work, we report ligand field density functional theory (LFDFT) approach dedicated to the simulations of X-ray absorption near edge structure (XANES) profiles of Sr in UO2 microstructure. The chemical properties and the next neighbor Sr atomic environment have been analyzed considering the limited Sr solubility in UO2 matrix in addition to the formation of strontium oxide phase as well as other fission-products compounds normally found in spent UO2 nuclear fuel. The free energy of fission products oxides and incorporation energy of Sr in UO2 microstructure have been calculated by means of the DFT+U formalism. The results reveal the nature as well as configuration of XANES profiles proper to 1s electrons excitation in Sr and to the respective electric dipole allowed transitions at the Sr K-edge. Multiple resonances in the Sr K-edge XANES spectrum, predominantly due to the simultaneous excitations such as of 1s4s and 1s4d electrons etc., have been also identified. The theoretical LFDFT approach presented in this work serves as the basis for modeling experimental Sr K-edge XANES of irradiated uranium dioxide samples.

Resume : Large area electronics provides revolutionary means to integrate electronic functionality in formerly passive surfaces. Transfer of this technology to space applications, medical X-ray detectors or areas with radioactive contamination requires the materials and devices to resist to the continuous exposure to ionizing radiation. Here we demonstrate the radiation hardness of oxide semiconductors such as Gallium Indium Zinc Oxide (GIZO) which combine unique advantages in processing with robust electrical performance in thin-film transistors (TFTs) [1]. In contrast we find that a material platform based on organic electronics shows degradation under ionizing conditions. In the experiments we subjected oxide as well as organic transistors to X-ray radiation and monitored the transistor performance metrics as a function of total ionizing dose. Flexible oxide TFTs maintained a constant mobility of 10 cm2/Vs even after exposure to doses of 410 krad(SiO2) whereas OTFT lost 50% of their transport performance. Relevant temporary damage originates in oxide TFTs only due to the trapping of ionization charges in dielectric layers. However, these effects are only observed at high dose-rates and recover on a time-scale of hours. In addition they scale with the thickness of the dielectric, thus allowing to fabricate radiation transparent flexible transistors by employing thin dielectrics. We attribute these exceptional resistance of oxide semiconductors to ionization damage to their intrinsic properties such as independence of transport on long-range order and large ionic lattice energy. [1] T.Cramer et al. Adv. Elec. Mat. 2016, DOI: 10.1002/aelm.201500489

Resume : FeCr alloys are model alloys of high-Cr Ferritic-Martensitic steels that are candidates for structural alloys for the future GEN IV and fusion reactors. Their in-service behaviour is thus a key issue. Nevertheless, neutron irradiations are expensive and access to facilities is restricted. The community thus focuses on model irradiation experiments using alternative irradiation sources. However, transferability issues arise. One major difference between the irradiation response of ion and neutron Fe-Cr irradiated alloys concerns the α/α' decomposition: α’-phase particles were formed in Fe-Cr alloys with 9 at% Cr or more under neutron irradiation [1, 2] at 300°C at low dose but not under ion irradiation [3]. On the other hand, in low purity Fe-Cr model alloys, formation of NiSiPCr-enriched clusters formed under both ion and neutron irradiations was revealed with Atom Probe Tomography (APT). Given the widespread use of ion-irradiation as a surrogate for neutron irradiation, understanding of the origin of these differences and commonalities is of high importance. In order to improve our understanding, Fe-Cr alloys were irradiated with ions and electrons at 300°C with similar dose rate and characterized with APT. Comparison with neutron data enabled to conclude on the role of cascade size, dose rate and injected ions on the α/α' decomposition and to explain the absence of α' precipitation under ion irradiation reported in previous studies. The influence of irradiation conditions on the formation of the NiSiPCr-enriched clusters has also been investigated. [1] Kuksenko et al. , J. Nucl. Mater. 432 (2013) 160. [2] Bachhav et al. Scr. Mater. 74 (2014) 48 [3] Pareige,et al. J. Nucl. Mater. 456 (2015) 471

Resume : The objective of this talk is to demonstrate the validity of transmission electron microscopic (TEM) in situ micropillar compression tests for quantitative assessment of yield stress and elastic modulus in irradiated oxide dispersion strengthened (ODS) alloys. Materials testing for future advanced nuclear reactor concepts is increasingly calling upon charged particle irradiations as a surrogate for neutron irradiation. Ion irradiations can deliver a damage rate up to three orders of magnitude higher than that of neutron irradiations, with little or no residual radioactivity, and at fractions of the cost. However, evaluating the effects of ion irradiation on mechanical performance remains a challenge, due to the shallow damage layer of a few micrometers, at the surface. Thus, small-scale mechanical testing techniques have gained prominence for evaluating the mechanical performance of thin irradiation damage layers. Transmission electron microscopic (TEM)in situ mechanical testing holds even greaterpromise, however, as it is able to provide quantitative load-displacement data alongside TEM-resolution video of fundamental deformation processes in the material. This study focuses on a model Fe-9%Cr ODS alloy irradiated with 5.0 MeV Fe++ ions to 3 and 100 dpa (displacements per atom) at 500°C. The unirradiated condition is also studied as a control. We use a Hysitron PI95 TEM in situ mechanical testing holder fitted with a diamond flat punch tip.Compression tests are conducted on micropillars fabricated from the ODS alloy using focused ion beam milling. We observe a size effect only for pillars having minimum dimension <100 nm. For pillars above this size threshold, TEM in situ compression tests accurately measure yield stress values for both the irradiated and unirradiated specimens. The measured changes in yield stress with irradiation are consistent with microstructure-based predictions using the dispersed barrier hardening model. Elastic modulus values, when corrected for the amount of deflection and deformation occurring in the base material below each pillar, fall in agreement with expected values.

Resume : To study and understand the evolution of the microstructure of tungsten under irradiation in an environment close to those of future fusion reactors such as ITER and DEMO, the irradiations with 1.2 MeV W ions were carried out in tungsten samples. The irradiations have been performed at different conditions were investigated in the microscope in situ at JANNuS Orsay. Various fluences (with corresponding damage doses in dpa, calculated by SRIM between 0-70 nm) were investigated between 0.01 and 0.06 dpa at irradiation temperature between -182 and 700 °C. These irradiated samples were then characterized by using Transmission Electronic Microscopy. In the case of low-temperature irradiation, defects like small vacancy clusters were detected in the TEM pictures. Their size and/or concentration increase with irradiation dose. Annealing of irradiated samples between RT and 1100 °C shows that the size and concentration of defects change. The vacancy cluster size increases with the increase of the temperature between 500 and 900 °C probably due to migration - agglomeration of defects. In order to understand the migration-agglomeration process during annealing some samples irradiated by 20 MeV W ions in JANNuS Saclay were characterized by Positon Annihilation Spectroscopy (PAS). Vacancy clusters were detected. Their size and/or concentration increase with annealing temperature. It is important to notice that PAS technique is more sensitive to small vacancy clusters and monovacancy can be detected by PAS [1]. The PAS and TEM observations shows that the vacancy cluster size increases but because of their different sensitivity the agglomeration stage is detected at higher temperature in TEM compared to PAS. [1]: Lhuillier, P.E., et al., Positron annihilation studies on the nature and thermal behaviour of irradiation induced defects in tungsten. physica status solidi (c), 2009. 6(11): p. 2329-2332.

Resume : Due to their high strength and high-temperature properties, tungsten-based alloys are being considered as plasma-facing materials in fusion devices. Under neutron irradiation, rhenium produced by nuclear transmutation, has been found to precipitate in elongated precipitates forming thermodynamic intermetallic phases at concentrations well below the solubility limit. This can lead to substantial hardening and affects the fracture toughness of W alloys. Here, using first-principles calculations based on density functional theory, we study the energetics of mixed interstitial defects in W-Re, W-V, and W-Ti alloys. We find that mixed interstitials in all systems are strongly attracted to each other with large binding energies and form interstitial pairs that are aligned along parallel first-neighbor <111> strings. Low barriers for defect translation and rotation enable defect agglomeration and alignment even at moderate temperatures. We propose that these elongated agglomerates of mixed-interstitials may act as precursors for the formation of needle-shaped intermetallic precipitates. Our study also reveals a negative heat of formation on the W-rich side of the W-Ti system. This is at odds with currently available thermodynamic assessments, which are limited by the sparsity of experimental data in the region below 1700 K. Here, using our first-principles data we present a revision of the solubility limit at low temperatures, which is relevant under fusion reactor conditions.

Resume : When a material is subjected to a flux of high-energy particles, its constituent atoms can be knocked from their equilibrium positions with a wide range of energies, depending on the exact nature of the collision. The spectrum of damage energy, derived from the exact knowledge of the recoil spectra for each nuclear reaction occurring in the solid, constitutes a vital data set required for understanding how materials evolve under irradiation. The knowledge of such damage energy is relevant to compare the impact of different facilities on the structural behavior and relevant properties of materials. The DART code was developed to give a comparison of radiation damage produced in materials by different nuclear plants and particle accelerators.