Nuclear materials under extreme conditions

Resume : First Mirrors (FMs) will play a crucial role in optical diagnostic systems of future fusion devices like the International Thermonuclear Experimental Reactor (ITER). Unlike to- day’s tokamaks, forthcoming reactors are expected to produce a high level of radiations and neutrons, preventing the use of common optical components (windows, fibers). Instead, an alternative solution based on an optical labyrinth embedded in the neutron shielding and employing metallic mirrors was proposed. Being the first element of the optical path which allows light to cross the neutron shielding, FMs will be placed close to the thermonuclear plasma and will, therefore, be subject to intense thermal and radiations loads, bombardment by plasma particles (mainly charge-exchange neutrals, CXNs) and deposition of material eroded from the plasma-facing components. The choice of the FM materials and the physical structure will be developed in details including several tests under plasma exposure and tokamaks. Net deposition of particles eroded from the First Wall (FW), i.e. mainly beryllium (Be) and tungsten (W), can degrade the reflectivity of FMs severely compromising the reliability of the optical diagnostics. Although passive mitigation techniques are predicted to reduce the amount of Be and W on FMs, optical degradation cannot be fully suppressed; in situ mirror cleaning techniques are indispensable and proposed techniques will be presented.

Resume : Because of their high surface/volume ratio which may allowed self-healing processes, mesoporous materials could be tolerant materials to radiation damage defects [1]. To investigate this topic, mesoporous and nonporous sol-gel silica thin films (e~100 nm) deposited on Si substrates [2] were irradiated with gold ions of medium energies from 0.5 MeV to 12 MeV. X-ray reflectivity measurement and SEM observations were used to monitor the porous network of the thin films. Infrared measurement was used to characterize the silica network. These characterizations show an evolution of the porous network of the materials depending on the irradiation conditions (fluence, energy of ions) and the initial parameters of the porous network (pore size, porous volume). A total compaction is achieved for a fluence of about 2×1014 cm-2 (~5×1021 keV.cm-3) (figure 1)[3]. The process of mesoporosity collapse seems different according to the irradiation regime (nuclear versus electronic). The sol-gel mesoporous and nonporous samples exhibit a delayed radiation damage compared to material elaborate by classical route, which indicates that sol-gel materials are more radiation tolerant from this point of view. The presentation aims to discuss these different observations, and clarified the role of the interfacial surface on the healing of the defects created by irradiation. From another point of view, the sensitivity of these mesoporous structures to the radiation damage opens interesting prospects for obtaining self-conditioning materials.

Resume : Helium (He) and hydrogen (H) are two most important transmutation products in structural materials applied in fusion and accelerator driven systems. In a fusion reactor, the first wall blankets are exposed to 14 MeV neutron irradiation, which produces He and H at about 11 appm He/dpa and 40 appm H/dpa. In accelerator driven systems (ADS) the high energy protons produce He and H at much higher rates, up to about 200 appm He/dpa and 1000 appm H/dpa in components like proton beam windows. Since it is known that He or H may induce strong embrittlement effect in many structural materials, the He and H effects are of great concern for the safe operation of fusion and ADS facilities. In the last two decades, various structural materials were irradiated in the targets of the Swiss spallation source (SINQ) within the STIP irradiation program. Of the particular interest to developing high power spallation sources for both neutron science and ADS applications, austenitic steels, tempered martensitic steels, Ni- and Al-based alloys have been extensively investigated. Post-irradiation examinations (PIE) include mechanical testing using tensile, bend, impact etc techniques and microstructural analysis employing transmission electron microscopy (TEM), positron annihilation spectroscopy (PAS) and atom probe tomography (ATP). The specimens used for the PIE are with irradiation dose up to about 20 dpa and helium concentration of about 1800 appm, irradiated in a temperature range of 60 – 500°C. The hydrogen content is hardly to determine, due to its high mobility in materials, especially at relatively higher temperatures. The results of mechanical tests demonstrate pronounced embrittlement effect when the He concentration is above about 500 appm. Meanwhile, additional hardening effect is also observed, which can be attributed to high-density He bubbles of 1-2 nm in diameter. In this presentation, the STIP irradiation program will be briefly introduced. The results of austenitic steels, tempered martensitic steels and ODS steels irradiated in SINQ targets will be shown and compared with those obtained from irradiations in fission reactors. The He and H effects on these materials will be discussed.

Resume : Mo mirrors, pristine and coated with W films, were irradiated with Mo and He ions to simulate the effects of plasma exposure on diagnostic mirrors to be applied in D-T fusion devices. The aim was to determine the impact of irradiations on mirrors’ reflectivity. Scanning transmission electron microscope investigations were carried out to determine microstructural changes in the surface region. Both surfaces and lamellae cut as cross-sections through the implanted regions were studied. Irradiation of uncoated Mo mirrors with Mo and He ions of 1.5x1015/cm2 and 8x1016/cm2, respectively leads to the formation of He nano-bubbles (d < 1nm) uniformly distributed beneath the mirror surface down to the depth of 15 nm and blisters of the average diameter of 50 nm. The presence of W layers of 4 and 10 nm in thickness had a significant impact on reflectivity as well as on the size and distribution of bubbles and blisters. With the increase of W thickness bubble formation is restricted to within the W layer and the size of blisters increases considerably to even 4 m. This is attributed to He accumulation at the interface between W and Mo especially in pristine mirror surface defect such as scratches coated by the W layer free from defects and grain boundaries that alters the diffusion process. In such places the gas pressure can built high enough to cause the blister onset and growth resulting in non-uniform distribution of blisters in thicker W layer coated Mo mirrors.

Resume : Ceramic Compounds such as nitrides and carbides of transition metals (ZrN, ZrC, TiC, or TiN) are of increasing interest nowadays for various technological applications, aiming to their high hardness, high wear resistance, high melting temperature, and high electrical and thermal conductivities. The latest potential applications include space technology, nuclear barrier coatings in nuclear reactors, and materials for plasmonic devices. In this talk, we present measurements of broadband optical reflectance (20 meV to 6 eV) on argon and gold-ion irradiated ceramic thin films, obtained by Pulsed Laser Deposition. We show that at fluences in the range of 1E14 - 2E15 cmE-2, both Ar and Au-ion irradiations have a significant effect on far and mid-infrared reflectance, which is dominated by the free (Drude) carriers response. Low fluence irradiation enhances the Drude conductivity by reducing the scattering rate. Additionally, Au-ion irradiation increases the carrier concentration, therefore acting as a charge dopant. We further show that the effect of irradiation is to enhance the plasmonic performances of these ceramics, especially in the visible and near-infrared ranges.

Resume : The phase separation observed at low temperature (below \textit{circa} 600~K) in the U_{1-y}Ce_{y}O_{2-x} system and for values of y between roughly 0.34 an 0.5, purportedly involves fluorite structures only. However, for y values above 0.5, an oxygen deficient C-type bixbyite is also reported. In this work, the phase separation in U_{0.54}Ce_{0.46}O_{2-x} has been reexamined using x-ray and neutron diffraction. Below a critical temperature, the existence of two fluorite related structures in the miscibility gap is confirmed: a stoichiometric U_{0.54}Ce_{0.46}O_{2} phase and an oxygen-deficient U_{0.54}Ce_{0.46}O_{2-x} phase. Although the former is indeed a fluorite, we show that the other end-member phase has a C-type bixbyite structure. This would suggest that the oxygen deficient phase can be described as a bixbyite over the entire cerium composition range.

Resume : Measurement of structural properties of UO2 fuels is a key component in understanding and for the modeling of the performance and safety of nuclear fuels in the reactor. The steep temperature distribution and the concomitant intense radiation field during fuel burnup affect and alter the crystallographic UO2 lattice structure in the irradiated matrix. This study examines the effect of neutron irradiation on the distortion of UO2 crystal lattice structure in high burnup spent fuel material having an average burnup of 62 MWd/kgU in the pellet. Experiments have been carried out using synchrotron radiation microbeam X-ray diffraction (XRD) technique at the Swiss Light Source, Paul Scherrer Institute, Switzerland. Microbeam XRD data of Bragg peaks originating from different (hkl) planes have been analyzed. Structural analysis with XRD indicates that the lattice constant and unit cell volume of UO2 in spent fuel have increased, and the measured lattice strain in the irradiated matrix is considerably higher when compared to the predictive critical strain value of UO2 ploygonization. These results can reasonably be ascribed to the fission products dissolution and accumulated radiation-induced defects, which modify the UO2 crystallites in the irradiated matrix. The experimental method adopted here also allows investigation of UO2 crystal properties and possible dimensional changes within a single grain in the irradiated matrix, necessary to understand better the initiation of UO2 grains (sub)division, origin of high burnup structure and fuel restructuring at extended burnup. All these results will be presented and discussed.

Resume : Grain boundary energies in austenitic twinning-induced plasticity (TWIP) steel, computed using the meta-atom embedded atom method (EAM) potential, are compared with experimental boundary energies in a high-manganese TWIP steel [H. Beladi, N. T. Nuhfer, and G. S. Rohrer, Acta Materialia (2014) 70: 281–289]. We found that there is a strong correlation between the experimental and simulated boundary energies for the most frequently observed boundaries, while the experimental energies of less observed boundaries are not correlated with the simulated energies [S. Ratanaphan, R. Sarochawikasit, N. Kumanuvong, S. Hayakawa, H. Beladi, G. S. Rohrer, and Taira Okita, J Mater Sci (2019) 54:5570–5583]. We also observed that the simulated grain boundary energies and the measured grain boundary population in the high-manganese TWIP steel are inversely correlated when groups of boundaries with fixed misorientations are compared. Because, these inverse relationships are consistent with our recent study in the high purity polycrystalline copper [S. Ratanaphan, D. Raabe, R. Sarochawikasit, D. L. Olmsted, G. S. Rohrer, and K. N. Tu, J Mater Sci (2017) 52:4070–4085], it might be possible to obtain a larger set of grain boundary energies in the TWIP steel by using the Bulatov-Reed-Kumar (BRK) grain boundary energy function for face-centered cubic (fcc) metals [V.V. Bulatov et al. Acta Materialia (2014) 65:161–175]. It should be noted that these interpolated grain boundary energies were consistent with the simulated energies and also inversely correlated to the measured grain boundary populations. Regardless of compositional complexity in the TWIP steel composed of multiple elements, this work shows that the shapes of the grain boundary energy distributions in the austenitic alloy and pure fcc metals are similar. The results in the current study could impact to the field of Grain Boundary Engineering (GBE) in the complex alloys and the application of GBE to the high entropy alloys.

Resume : Understanding the phase stability of alloys with multiple principal elements under irradiation is one of the great challenges in the developing of materials for engineering components such as the plasma-facing materials for fusion reactors. Recently, it was shown that high entropy alloys (HEAs) based on Fe-Cr-Ni-Mn have better radiation resistance in a comparison with austenitic steels. The investigation of phase stability and magnetic properties of quaternary Fe-Cr-Mn-Ni alloys using a combination of density functional theory (DFT), cluster expansion method and Monte Carlo (MC) simulations has shown that the addition of Mn to Fe-Cr-Ni alloys decreases the effect of magnetic moment changing with the volume. This fact supports the experimental results of reduced swelling. The point defect properties of Fe-Cr-Mn-Ni were studied as a function of local environment of a defect and chemical short-range ordering of alloy. The representative structures of HEA have been obtained using DFT-based MC simulations. Vacancy formation energies and migration barriers have been calculated for each element, including varying next-nearest neighborhood in the disordered alloys. The dependence of migration properties on the chemical composition and short-range ordering is analyzed and presented.

Resume : Recently, multi-component alloys containing four or more elements in equal or near-equal atomic compositions or so-called High-Entropy Alloys (HEAs) have attracted great interest in high-temperature materials applications due to the presence of significantly larger configurational entropy contributions to alloy free energies. Understanding the phase stability under irradiation of alloys with multiple principal elements is one of the great challenges in developing materials components for nuclear reactors. A body-centered cubic (bcc) quaternary W-Ta-Cr-V refractory high entropy alloy with outstanding radiation resistance has been developed as a thin film using magnetron sputtering deposition and no sign of irradiation-created dislocation loops, even after 8 dpa, was observed. The ab-initio based Cluster Expansion (CE) model developed for quinary Cr-Ta-Ti-V-W system in combination with Monte Carlo (MC) simulations has been used to explain the formation of Cr- and V-rich second-phase particles observed using TEM, APT and XRD in the irradiated W38Ta36Cr15V11 alloy. Moreover, the Density-Functional Theory (DFT)-based MC simulations enabled to investigate systematically the dependence of phase stability and short-range order of HEAs as a function of composition. Experimental characterization using light microscopy, SEM, EDS and XRD on arc-melted and annealed samples has been performed to validate the modelling results from the DFT-based MC simulations.

Resume : Laser induced breakdown spectroscopy (LIBS) technique has got a lot of attention in the last years as a useful tool for elemental analysis of various compounds. The flexibility of the technique makes it perfect for the analysis of multi-element compounds like high entropy alloys (HEAs). HEAs are defined as metallic alloys containing 4 or 5 elements which lead to physical properties superior to classical alloys, such as higher degree of fracture resistance, tensile strength, as well as corrosion and oxidation resistance, which are essential for applications in nuclear industry. In this study we investigated the elemental composition of a series of 6 HEAs using advanced surface investigations techniques like SEM, EDS and LIBS in order to understand the elemental distribution in both surface and volume of targets and deposited thin films. The surface mapping was performed by implementing LIBS technique (10Hz, 10ns, 5J/cm2, 1024 nm) on a relative wide area (20 mm x 20 mm) with the subsequent EDS and SEM analysis performed in the ablated crater area for an accurate comparison and a better elemental identification. Quantitative investigations were performed regarding the plasma excitation temperature and electron density. The local heterogeneity of the irradiated spot was corelated with the values of these plasma parameters. Finally, thin films were deposited using pulsed laser deposition technique on a wide range of experimental conditions (various fluences, background pressures, repetition rates). Correct stoichiometry transfer was attempted as well as tailored changes in stoichiometry for the improvement of the thin film physical properties.

Resume : Reduced activation ferritic/martensitic (RAFM) steels are key materials studied in the fusion and Generation-IV reactor materials programs. In the last two decades, extensive mechanical testing and electron microscopy investigation were carried out to reveal the changes in mechanical properties and microstructure of these steels. In order to get a better understanding of the irradiation-induced microstructural and chemical evolutions, a comprehensive study has been performed on F82H, Eurofer 97 irradiated at doses up to 20 dpa by combining Atom Probe Tomography (APT) and Transmission Electron Microscopy (TEM) techniques. The main subjects include: 1) radiation-induced segregation (RIS) at grain-boundaries (GBs) and interfaces of large carbides, 2) radiation-induced precipitate formation and evolution. The segregation behaviors of all the detectable solute elements of the steels (Cr, W, Mn, C, V, Si, Ni, Ta, P, B) and the transmutants of spallation reaction (Ti, Sc, Ca, K) at both GBs and precipitate-matrix interfaces have been studied. It was observed that some elements (e.g. Cr, Si) enriched, while some other elements (e.g. V, W) depleted at GBs after irradiation. Some differences between low-angle and high-angle GBs could be figured out. For most of solute elements, RIS is more evident at high-angle GBs. Irradiation induced precipitate evolution includes the formation of new precipitates and the dissolution of pre-exist precipitates. It was found that the irradiation changed the composition of the outer edge of pre-exist carbides (M23C6 and MX), e.g. C, Cr, W and V diffused from the M23C6 precipitate into matrix, W, V and Ta depleted in the phase boundaries. Although ’ phase was not detected, Cr-rich clusters, together with enrichment of some other elements (e.g. Si, Mn), were observed after irradiation.

Resume : The passage of ionising radiation through structural materials, depositing energies in the keV to MeV regime, has been under scrutiny for decades [1, 2]. Specifically, the characterisation by means of multiscale modelling techniques of the Nuclear and Electronic Stopping Powers in response to irradiation of advanced nuclear materials [3] traditionally described via Monte Carlo [4] methods, not designed to deal with electron dynamics, has become a hot research topic. A multiscale modelling methodology based on state-of-the art DFT simulations (hybrid formalisms, perturbation theory, time-dependent DFT) in combination with Ab Initio MD and Ehrenfest dynamics techniques to tackle intense electronic excitation effects in nuclear materials is presented. An operational interconnection among these modelling strategies relies undisputedly on the development of transferable interatomic potentials. Machine learning approaches as GAP [5], SNAP [6] or Atomicrex [7], among others, may be used for that purpose. The main advantages lie on DFT-like accuracy of MD results at a low computational cost. The present approach has been checked against the behavior of light H and He impurities following the passage of an ionising particle through the archetypical plasma facing material W metallic matrix. [1] P. Sigmund, Particle penetration and radiation effects, Springer, Volumes I (2006) and II (2014). [2] K. Nordlund et al., J. Nucl. Mater. 512 (2018) 450. [3] A. Correa, Comp. Mater. Sci. 150 (2018) 291. [4] J. F. Ziegler, M. D. Ziegler, J. P. Biersack, Nucl. Instr. Meth. Phys. Res. B 268 (2010) 1818. [5] W. J. Szlachta, A. P. Bartók, G. Csányi, Phys. Rev. B 90 (2014) 104108. [6] A. P. Thompson, L. P. Swiler, C. R. Trott, S. M. Foiles, G. J. Tucker, J. Comp. Phys. 285 (2015) 316. [7] A. Stukowski, E. Fransson, M. Mock, P. Erhart, Modelling Simul. Mater. Sci. Eng. 25 (2017) 055003.

Resume : Spinel-structured magnesium aluminate (MgAl2O4) possesses high transparency from visible to infrared wavelength range, enhanced strength and melting temperature, as well as excellent chemical and radiation resistance. Combination of these properties makes magnesium aluminate suitable for a number of technological applications, including involvement in construction of fusion and fission reactors inert matrices for nuclear fuels, radiofrequency windows for fusion reactors, where spinel exhibits a very high tolerance to irradiation with fast neutrons due to efficient recombination of primary Frenkel defects – vacancies and interstitials (i–V pairs). Here we analyze by means of ab initio calculations stability and properties of all types vacancies and antisite defects in various charge states for better understanding of annealing processes. Using LCAO approach and hybrid B3LYP functional as implemented in the CRYSTAL17 [1] computer code structural and optical properties are evaluated for each defect, suggesting possible defect-specific "fingerprints" for EPR and vibrational spectroscopy. For example, we calculated the Al or Mg vacancy and Mg antisite with simultaneous hole defect localization on oxygen in accordance with the EPR measurements [2]. Analysis of structural changes and calculated density of states revealed a correspondence between our calculation results and EPR results. In addition, we were able to compare model Hamiltonian parameters found in present ab initio calculations and experiments model. [1] http://www.crystal.unito.it/index.php [2] Lushchik, A., S. Dolgov, E. Feldbach, R. Pareja, A. I. Popov, E. Shablonin, and V. Seeman. Nucl. Instrum. Methods Phys. Res 435 (2018) 31-37

Resume : Fe-Cr based alloys are promising structural materials for components in fusion power plants. One of the main issues of using these alloys is related to the embrittlement by He produced under neutron irradiation. He atoms have low solubility in metals and tend to segregate in defects such as vacancies or grain boundaries (GBs) where He can form bubbles. Cr atoms may also segregate to GBs. Moreover, GBs have considerable influence on mechanical properties. Therefore, the segregation of Cr and He to GBs is an important effect for understanding properly the performance of ferritic steels based on the Fe-Cr alloys. Currently we investigated the segregation of Cr and He atoms at Fe-Cr tilt GBs, namely Σ3 and Σ5. Density functional theory (DFT) calculations of systems with GBs were performed to study Cr and He segregation energies and their impact onto GBs cohesion. Moreover, the formation energies of point defects were studied as a function of distance from the GB’s plane. The NEB method was applied to determine the migration barriers of Fe, Cr, He. The established database was used to parameterize the Atomistic Kinetic Monte Carlo (AKMC) model that allowed to obtain Langmuir-McLean segregation isotherms. Those simulations enabled to analyse the relations between global chemical composition and temperature with segregation processes (depletion or enrichment of element at GBs) for different types of GBs. DFT calculations of bulk systems combined with the Cluster Expansion method were applied to study the interactions between atoms and vacancies. The obtained effective cluster interaction will be used in the future in the AKMC model to investigate the segregation to different GBs in Fe-Cr-He alloys as a function of temperature and Cr, He concentrations

Resume : Research and development in nuclear energy requires analysis of the investigated nuclear materials for assessing their performances. After virtual or real sampling the analysis is performed in-, on-, off- line in a passive or interactive way. Passive techniques imply detection of self-emitted particle or radiation such as phonons, photons, leptons, neutrons or ions… Interactive techniques deal with interaction methods with excitation using phonons, leptons, neutrons or ions… The paper presents examples of analytical techniques utilised to characterize fresh fuel, spent fuel, reactor structural material as well as specific issues associated with their operational status.

Resume : The color F centers (electron trapped in the halide vacancy) are very common defects in alkali halides, identified by means of the ESR and optical absorption/luminescence. In particular, their properties in fluorites (CaF2, SrF2 and BaF2) are very well studied and understood in detail. However, the manifestation of similar defects in binary oxides with fluorite structure (CeO2 or UO2) is debated for a long time. Recently, we successfully applied the Mollwo-Ivey rule (correlation between lattice constant and optical absorption energy) to explain similarity of optical properties of the F-type centers in alkali halides, alkaline earth oxides and sulfides with NaCl lattice structure [1]. In this study, we have performed a similar comparative analysis of the F-type centers in alkaline earth halides and oxides with fluorite structure supported by detailed analysis of the literature. Special attention is paid to oxygen vacancies in partly covalent CeO2. Obtained results and conclusions are supported by the first principles calculations of the atomic and electronic structure of defective ceria [2]. [1]. A.I. Popov, E.A. Kotomin, J. Maier, Nucl.Inst.Meth. B 268 (2010) 3084-3089. [2] R.A. Evarestov, E.A. Kotomin, J.Maier et al, Phys Chem Chem Phys 19 (2017) 8340

Resume : In the past decades, zirconium alloys have been widely used as fuel cladding materials because of their low thermal neutron absorption cross-section, excellent mechanical properties, and good corrosion resistance under normal pressure water reactor (PWR) operating condition. However, zirconium alloys have weak oxidation resistance especially at elevated temperature, because they react with water, releasing a large amount of hydrogen gas and heat above 1473 K. After the Fukushima Daiichi nuclear accident, the major issue of nuclear researchers has been an improvement in the oxidation resistance under beyond-design accident conditions. Therefore, accident tolerant fuel (ATF) has been widely studied, which has high-temperature oxidation resistance. Coating technology is an effective way to protect fuel claddings from steam oxidation. Several surface coating technologies have been applied to deposit coatings on zirconium alloys, including physical vapor deposition, chemical vapor deposition, and spray coatings. Among coating methods, PVD is one of the promising coating methods, and it is widely used in the fields of oxidation protection. Moreover, Preparation of chromium alloy coatings, known for their high hardness, excellent corrosion, and wear resistance, by PVD is well established for corrosion resistance. In this work, we deposited chromium alloy coatings on Zircaloy-4 tubes by cathodic arc ion plating, and r.f. sputtering to study the effect of PVD deposition technologies on corrosion, adhesion, and residual stress.

Resume : Research reactor fuels are fabricated as rod-type or plate-type dispersed fuels by mixing the uranium powder into the metallic matrix powder which has high thermal conductivity for the efficient discharge of heat by fission. The dispersion fuel fabrication processes are carried out in the order of powder preparation, powder mixing, compaction, extrusion or rolling. Due to the nature of powder mixing process, partial heterogeneous mixing such as segregation is likely to occur. As a result, the homogeneity of uranium in the dispersion fuel is lowered and the irradiation stability of the fuel can be impaired. In this study, aluminum matrix was coated on the each fuel powder instead of powder mixing to improve the uranium homogeneity in the fuel. The aluminum coating was deposited to uniform thickness throughout the powder surface, and the coating thickness was controlled according to the PVD process parameters. Because the uranium loading in the fuel is determined by the thickness of the matrix coating and the size distribution of the fuel powder, correlation equation was derived.

Resume : Development of erosion-resistant functional materials applicable as plasma facing first wall components is crucial to guarantee a reasonable lifetime for future fusion reactors. In the frame of the qualification activities on the ITER like divertor plasma-facing components [1], X-ray imaging (microtomography / laminography – μXCT/L) and microbeam fluorescence (μXRF) were developed for microstructural analysis of tungsten coated carbon substrate materials and of smart marker lamellae that are fabricated in multilayer configuration of high Z materials. Appling μXCL, planar objects were visualized by increasing the quality of the grey level histogram based on the X-ray interaction with the W coating [2]. Thus, high resolution (~2 μm in lateral and ~5 μm in depth) stack of images of W coated CFC samples were collected. As expected, the W layers spatially correlate with the 3D morphology of the carbon-based composite substrate, which is isotropically resolved by μXCT at maximum space resolution of the W coated layer. The μXCL analysis is reinforced with additional information on W layer thickness as determined via an ad-hoc K-line XRF method intended to overcome the thickness saturation threshold. Smart marker lamellae that are fabricated in multilayer configuration of high Z materials (W and Mo layers on bulk W) are good functional simulator of the current ITER full-tungsten divertor design. Our solution for erosion analysis consists of high-resolution thickness and composition mapping by μXRF. Customized μXRF calibration protocols were based on a combination of standard samples, Monte Carlo simulations and Fundamental Parameter theoretical approach. X-ray imaging methods of plasma exposed samples could provide a useful reference overview of the coating structural integrity including those with opaque deposits or contaminants. [1] Rubel, Marek et al. Journal of Nuclear Materials, vol. 438, pp. S1204-S1207, 2013; [2] Misawa M. and Tiseanu I., 2006 Oblique-view cone-beam CT, US Patent 7139363.

Resume : Behavior of materials in radiation harsh environment is an important issue and damage build-up within the material in such environment can exhibit complex behavior. Therefore, as one of the most prominent candidate materials that can be used in radiation harsh environments, response of GaN to ion irradiation should be studied in detail. Dense electronic excitation in the wake of the swift heavy ion passage through a material (mass > 20 amu, kinetic energy > 1 MeV/amu) can lead to nanoscale material damage along ion trajectory called ion track [1], [2]. Irradiation with heavy ions in the keV energy range can also produce defects but via different energy dissipation channel. In this case, it is well known that direct ion-atom collision (i.e. nuclear energy loss) can produce defects in GaN [3]. In our previous study [1], we found no evidence of ion track formation in the bulk after swift heavy ion irradiation using 23 MeV I and 90 MeV Xe beams. Recently, synergistic effects of nuclear and electronic energy loss came into research focus. After introducing additional disorder into moderately damaged GaN crystals via pre-irradiating them either with 2 MeV Au or with 900 MeV Xe ion beams, we were able to introduce additional damage using the same (as in our previous study [1]) 23 MeV I and 90 MeV Xe ion beams. Here, we report the results of sequential ion irradiation of GaN based on the RBS/c, TEM and AFM measurements. [1] M. Karlušić et al., Response of GaN to energetic ion irradiation - conditions for ion track formation, J. Phys. D: Appl. Phys. 48 (2015) 325304 [2] M. Sall et al., Track formation in III-N semiconductors irradiated by swift heavy ions and fullerene and re-evaluation of the inelastic thermal spike model, J. Mater. Sci. (2015) 50:5214–5227 [3] S. O. Kucheyev et al., Ion implantation into GaN, Mater. Sci. Eng., R , 33 (2001) 51-107 *Corresponding author: marko.karlusic@irb.hr

Resume : HLW(High Level Nuclear Waste) is usually stored as spent nuclear fuel or vitrified glass. In order to provide a solid scientific basis of waste disposal we try to understand the properties of various relevant materials, including waste and waste forms, and their long-term performance. In this contribution, we report the results of joint atomistic modeling and experimental efforts of spent nuclear fuel/HLW related materials such as nuclear glass and UO2-based model systems. Vitrification into borosilicate glass is a widely used process for long-term storage of the spent nuclear fuel. In order to understand the physical and chemical properties of glass waste forms and their changes under irradiation conditions we have performed a series of measurements and simulations of elastic moduli, hardness and internal energies of a series of irradiated borosilicate glasses. Both experimental and simulation results show significant change (decrease) of Young’s modulus and hardness of borosilicate glasses and amorphization at ~0.1 dpa (displacements per atom). The molecular dynamics simulations of the irradiation process indicate significant changes in volume and stored energy, which are B/Si content dependent. We also applied atomistic simulations to investigate the behavior of Cr (fission product) in the UO2 matrix and the properties of uranium-based secondary phases. We will discuss the simulated doping geometry and oxidation state of Cr, the limit of the Cr solubility in the host matrix, electronic states of the mixed U-oxides as well as the properties of the O-depleted MUO4-x (M=Ca,Sr) phases.

Resume : Y3Al5O12 (YAG), single crystals are known for their interesting properties such as high radiation-resistance, high melting point and high thermal conductivity. Consequently, YAG is among candidates to several technological applications such as fusion energy devices and nuclear applications. Irradiation of single crystal Y3Al5O12 in the reactor (i.e. fast and thermal neutrons and gamma radiation) produces many color centers (such as F-type center, interstitials [1-5] and many others) in the material. Similar defects are also formed by heavy ion irradiation or by fast electrons, while only electronic processes are important in the case of UV, X-ray and low energetic electron irradiation.. In this presentation we report the results of the thermostimulated luminescence measurements, performed between 300 and 720°K of the stored energy in Y3Al5O12 single crystals, irradiated by fast neutrons with fluences of 2.1 x 1017 or 2.18 x 1019 n/cm2, or also by 1.8 MeV electrons, or thermochemically reduced. A clear pronounced dose effect was found and analyzed. In particular, four TSL peaks were observed in Y3Al5O12 samples subjected 2.18 x 1019 n/cm2, while in sample subjected 2.16 x 1017 n/cm2, only three TSL peaks were detected. A comprehensive kinetic analysis of the glow peaks in Y3Al5O12 is performed. As usual, each TSL peak is characterized by the appropriate activation energies, which both crystals are 0.8 – 1.3 eV. The obtained values are compared with the appropriate activation energies for F-type center annealing in neutron- and heavy-ion irradiated Y3Al5O12 as well as also with similar TSL data for Al2O3 and Y3Al5O12.

Resume : Thin films used for nuclear fuel encapsulation applications in advanced nuclear reactors should exhibit many excellent characteristics such as high temperature stability, a good radiation tolerance and retention of fission products. ZrC is a potential material for such applications since it possesses a high melting temperature, a single crystalline phase up to the melting point and good thermochemical stability. Many investigations showed that ZrC has a high irradiation tolerance when irradiated with high energy ions. We investigated Pd diffusion, one of the most radioactive fission products, through thin films of ZrC. First, a thin film of Pd was deposited using magnetron rf sputtering at room temperature on a Si wafer. Then, without breaking the vacuum, a thin ZrC film was then deposited using pulsed laser deposition on top of the Pd film. The structure, surface, interface morphology and density of the deposited films were investigated using X-ray reflectivity and grazing incidence X-ray diffraction. After deposition, the sample was transferred to a Time of Flight Secondary Ion Mass Spectroscopy (ToF SIMS) system. The sample was mounted on the heating stage of the ToF SIMS. A depth profile was obtained at room temperature, where after the sample was heated to 400 °C at a rate of 1 K/s. The temperature was maintained at 400 °C for 10, 20, 40, 80 and 160 minutes, respectively. In between each heating cycle, a depth profile was obtained. From the depth profiles, it was clear that the ZrC formed a diffusion barrier that prevented the Pd to diffuse towards surface. The Si diffused up to the ZrC diffusion barrier and was blocked by the ZrC layer.

Resume : Thin films and devices used in advanced nuclear reactors, fusion installations or space exploration are exposed to high level of various type of radiation. It has been recently observed that nanostructured thin films, which possess very small crystal grains exhibited better radiation resistance than polycrystalline or single crystal films. Therefore, the structure, composition and properties of these films will be less affected by exposure to radiation. Dislocations cannot form in such small crystalline grains and the increase of mechanical hardness after irradiation, observed for single crystals or large grain films is not observed in these nanostructured films. To investigate the radiation effects on properties, chemical composition and structure we used the Pulsed Laser Deposition (PLD) technique to grow ZrC and ZrN nanocrystalline thin films. The effects of 800 keV Ar and 1.0 MeV Au ions on the properties of nanocrystalline ZrC and ZrN thin films were investigated using high resolution transmission electron microscopy, nanoindentation, X-ray specular and diffuse reflectivity, and X-ray diffraction investigations. The results confirmed that nanocrystalline films could withstand high irradiation fluences without degrading their crystalline structure, while the Si substrate was completely amorphized. Grazing incidence XRD investigations found that there is grain growth in ZrC films as an effect of ion irradiation, while a decrease of the grain size was observed under similar irradiation conditions for ZrN.

Resume : Amorphous Indium Zinc Oxide (IZO) is a transparent conductive oxide (TCO), which could be synthesized at low temperatures for applications on optoelectronic devices fabricated on plastic or paper substrates. Since IZO films are amorphous, devices incorporating them are rather insensitive to radiation and could be used in outer space. IZO electrical and optical properties strongly depend on both In/(In+Zn) value and oxygen content. In this study, thermal conductivity of high and low electrical conductivity IZO thin films (In/(In+Zn) = 0.40 and 0.70) before and after gamma irradiation was investigated using the laser flash method (TF-LFA). IZO films were grown on Si substrate at room temperature using pulsed laser deposition technique (PLD) with an ArF laser under various oxygen pressures. The IZO films electrical properties were measured using a 4-point probe system, the stoichiometry was investigated using energy dispersive X-rax spectroscopy and X-ray photoelectron spectroscopy, while the thickness was obtained from cross-section scanning electron microscopy images. For thermal conductivity measurements, a ~200 nm thick film of Au was deposited using thermal evaporation. The thermal conductivity of IZO films ranged from 0.7 W/m/K to 1.95 W/m/K. The phonon contribution to the total thermal conductivity was constant. The electric conductance of the IZO films varied inversely to the O2 pressure during deposition. No major change after gamma irradiation was observed in these IZO films.

Resume : In nuclear power plants, stainless steels are the predominant structural material. Ferritic and ferritic-martensitic steels are known to offer better dimensional stability than austenitic steels under neutron irradiation. The operating temperatures of these steels are however limited to about 600◦C. In order to achieve higher operating temperatures, oxide dispersion strengthened (ODS) steels have been developed. These modern steels are multicomponent alloys (with about 20 components: Fe,C,Mn,P,Si,Ni,Cr,Mo,V, Ti, Co, Cu, Al, B, W, Zr, N, O, Y) made up of Fe-Cr ferrite matrix in which a homogeneous distribution of small precipitate particles is created by ball milling and subsequent consolidation process involving either hot extrusion or hot isostatic pressing. These ODS steels have superior creep resistance and stability under irradiation, attributed to the finest oxide particles densely dispersed in the ferrite matrix with high dislocation density. The dissolution energy of the oxide particles in an iron matrix and bulk modulus of the oxide embedded iron are important thermophysical parameters of the ODS iron. At the high-temperature environment, there will be a thermal expansion so high bulk modulus is required to resist it. And the knowledge of the dissolution energy of oxides in iron is used to stop the movement of dislocation in the alloy. Several types of oxide particles are found in the ODS steel such as Y2Ti2O7, Y2TiO5, and a defective NaCl-type TiO. We, in this work, compute the dissolution energy and bulk modulus of NaCl-type oxide particles AO embedded iron, where A is transition metal, using density functional theory calculation. Our model of oxide-embedded iron is made by removing 4 Fe atoms at the centre of 128-atoms 4x4x4 bcc supercell and placing a NaCl-type A4O4 cluster such that the A’s substitute for the removed Fe atoms and O’s placed at the appropriate octahedral interstitial sites. The physical origin behind the endothermic/exothermic dissolution energy will be analyzed through the projected density of state. Analysis of the dissolution energies together with bulk moduli would improve our understanding of these oxides as precipitate particles in ODS steel.

Resume : The processing of plutonium and minor actinides (MA) waste, which have accumulated over the years, is an important problem in nuclear power. Creating inert matrix fuels (IMFs) for post-burning Pu and transmutation MA would be the best solution to this problem. Potential IMF are CeO2, PuO2, LnPO4, Y3Al5O12, etc. One of the main disadvantages of such materials is their low thermal conductivity, which can lead to uneven heating of the fuel pellet and, as a result, to its mechanical damage and decrease in the degree of the fuel burn time-up The solution to this problem can be the introduction of additional components, for example, metals, their oxides, nitrides, carbides, etc. In the present work, studies aimed at increasing the thermal conductivity of ceramics are performed. The object of this study are composites CeO2–SiC (0, 10, 20 vol.%). They were prepared using the SPS method (T=1025 °С, sintering time did not exceed 10 min). The microstructure of ceramics, their microhardness, fracture toughness and thermal conductivity (in the temperature range 25–1100 °C) were studied. It was established that an increase in the concentration of SiC leads to a significant increase in the thermal conductivity of composites at room temperature: by 49% (10% SiC) and 79% (20% SiC). In the high-temperature region, a significant increase in the thermal conductivity is not observed. This study has been carried out with the support of the Russian Science Foundation (Grant No. 16-13-10464).

Resume : Tungsten is chosen as a primary candidate to serve as a plasma facing component (PFC) in future fusion reactors. Demanding fusion environment features frequent thermal shocks, neutron bombardment and exposure to hydrogen isotopes. PFCneed to maintain mechanical properties during the lifetime as well as they should not store radioactive hydrogen isotopes both during reactor operation and after its decommissioning. Laboratory equipment is constantly developing in order to faithfully reproduce fusion reactor environment and study plasma facing materials behavior during their service. This study focuses on comparison of material damaged in two-step procedure of firstly ion bombardment and then deuterium exposure and single-step damaging by simultaneous W ion irradiation and D exposure. The latter seems to be closer to real fusion environment that the former, since in real conditions PFC are subjected to mutual actions of multiple particles. TEM investigation revealed that dislocation structures in the studied samples vary. Differences include both morphology of defects as well as their density. This result indicates that future experiments concerning D retention in irradiated tungsten targets need to be designed with more caution.

Resume : We demonstrate that the Swift-Hohenberg functional as used to describe patterning observed in out of equilibrium systems such as diblock copolymers, Rayleigh-Benard convection and thin film magnetic garnets, can be applied to radiation-induced patterns that occur in non-miscible alloys. By comparing ground states obtained from the minimization of this functional and a 2D numerical simulation performed on an irradiated AgCu material which is the archetype of a non-miscible alloy we show that the Swift-Hohenberg functional provides all possible patterns generated under irradiation and the solubility limits of radiation-induced precipitates in these patterns. To rationalize the formation of these radiation-induced patterns, we propose a generic "pseudo phase diagram" that relies not only on the irradiation flux and temperature but also the overall composition of the alloy. Tuning this overall composition offers the opportunity to tailor new materials with various micro-structures overcoming the limitation of the equilibrium phase diagram.

Resume : The characterization and testing of neutron irradiated materials for understanding the changes in microstructure and mechanical properties is time consuming and hazardous. Ion irradiation is an attractive alternative as it is relatively quick (~ hours/days) and does not render the sample radioactive in general. However, ion irradiation forms an extremely thin damaged layer (~ 100s of nm to a few m). Secondly, the damaged layer has a large gradient in the dose through the thickness, which makes the microstructural and mechanical changes vary appreciably through depth, making it difficult to characterize these aspects of the material after irradiation. Recent advances in technology have made the study of the mechanical properties of these thin layers easier. Nanoindentation has been used successfully to study the hardness change in the irradiated layers for some years. However, the three-dimensional stress state in nanoindented samples and the interaction of such a stress field with irradiated layers having a strong gradient makes it difficult to interpret the results. Finite element modelling at ANSTO has shown a pathway to analyse and interpret these results. As a further step towards an easier and improved methodology to study the mechanical properties of ion irradiated layers, in situ micromechanical testing including tension and compression tests have become more widespread in recent years. This progress has been facilitated by the advances in focused ion beam (FIB) technology allowing the fabrication of micro-sized samples. Here, the authors present some of the recent experimental and modelling results of such tests performed on ion irradiated materials, and discuss their pros and cons vis-à-vis nanoindentation methods.

Resume : Low-density materials are of great interest both for fundamental material science and high-tech applications. They belong to the class of nano-porous materials characterized by unconventional physical properties: nanometer-size pores, densities lower than water, within the 100 mg/cm3 regime, high specific surface areas, and high porosity. They have been used for microsensor pre-concentrators, bone implantation, scaffold for tissue regeneration, electrochemical supercapacitors, and lightweight microwave shielding structures. Foams show unique properties, as for example an almost perfect black-body-like absorption, an anomalous ferromagnetic behavior, additionally an increased gas and liquid adsorption and storage capability. The absorption of high-power fs-laser pulses by matter is observed to significantly increase in such density range. We report here on the generation of low density carbon foams by pulsed laser deposition through the irradiation of graphite targets in controlled environment. Different deposition geometries were attempted in order to take into account the flip-over effect and the strong asymmetries from the carbon generated plasma. The deposition process was calibrated by adjusting the laser beam energy, frequency, and deposition time on a wide range of values. Various background gasses (Ar, CH4) and pressures were used to find the best media that could aid the generation of complex carbon structures and the subsequent formation of foam-like films. The obtained films were investigated by means of complimentary surface analysis methods like ATR-FTIR spectroscopy, XRD, XRR, XPS, SEM, and EDS in order to accurately describe their structure, chemical bonds between the elements, and morphology.

Resume : Amorphous indium zinc oxide (IZO) films have good electronic and optical properties, which are used in transparent thin film transistors and display devices. These films maintain their properties when grown at room temperature, which allows for the use of inexpensive substrates such as paper or plastic. We used the pulsed laser deposition technique to grow thin films of IZO with In/(In+Zn) values from 0.1 to 0.9 on Si and glass substrates at room temperature. The films were irradiated by Co gamma radiation up to a dose of 30 kGy to investigate the effects of radiation on their structure and properties. The films surface morphology was smooth, with rms values below 1 nm. Spectroscopic ellipsometry was used to observe the effect of gamma irradiation on the optical properties. The results showed that after gamma irradiation the films density, extracted from simulations of the acquired X-ray reflectivity curves, decreased by a 1-2 %, which resulted in a small thickness increase, while spectroscopic ellipsometry measurements indicated that the refractive index values also decreased by 1-2 percent, consistent with the observed changes in density. Similar small changes were also observed in films electrical conductivity. These results showed that amorphous IZO is a material that could withstand a high level of gamma radiation without adverse effects on its structure or properties.

Resume : Tungsten is a promising candidate for a plasma-facing material in fusion reactors, given its excellent response in such extreme environment. However, the unavoidable presence of H and He impurities that tend to aggregate in preexisting material defects may lead to detrimental performance. Degradation may be deferred by nanostructure manufacturing, whereby a large density of interfaces are created. The influence of grain boundaries (GBs) on the separate behaviour of H and He in nanostructured W has been recently analysed [1, 2], albeit the simultaneous presence of both species at semicoherent W/W interfaces has not yet been simulated via ab initio methods. These may help discerning a long time running experimental controversy, namely, if a strong attraction between He and H, pointing to preferential retention of H nearby He clusters, is dominant [3]; or if on the contrary He addition leads to reduced/supressed blistering [4], whereas reduced H retention derives from nanobubble formation in the near surface layer [5]. Here, DFT simulations have been performed following the procedure presented in [2] for the construction of a W/W GB. The aim is to distinguish between the behaviour common to the presence of either H or He and the synergistic detrimental effects due to the coexistence of both light impurities and other intrinsic point defects. The stability and self-healing properties of nanoscale interfaces and their potential use in advanced nuclear reactors and other extreme engineering scenarios is assessed. [1] C. González et al, Nucl. Fusion 55 (2015) 113009 [2] C. González, R. Iglesias, J. Nucl. Mater. 514 (2019) 171 [3] S. Markelj et al, Nucl. Fusion 57 (2017) 064002 [4] Y. Ueda et al, J. Nucl. Mater. 386 (2009) 725 [5] M. Miyamoto et al, J. Nucl. Mater. 463 (2015) 333

Resume : Ultra-high power fs-laser opened new areas of physics and technology of laser-matter interaction under extreme conditions. There are already several operational 1 to 10 PW-level laser installations in Europe and worldwide, while even higher power installations, up to 100 PW are being planned. Experiments with these high power fs-laser beams require high-quality mirrors, which are using either dielectric layers/multilayers or metallic layers that have single pulse laser-induced damage thresholds from 0.1 up to 2 J/cm2 for 25 fs long laser pulses. However, multiple pulse laser-induced damage thresholds are getting lower with the increase of the number of pulses due to a process known as laser incubation. The actual physics of this incubation process is still a matter of investigation. We used the EPR technique to investigate the changes induced by fs-laser irradiation on HfO2 thin films at fluences below the laser-induced damage threshold. The HfO2 films were deposited by the pulsed laser deposition technique (PLD) using an ArF excimer laser. The fs-laser irradiations were performed with the aid of two laser installations. One installation emits 400 fs pulses at kHz repetition rates, while the other one emits 25 fs pulses at a repetition rate of 100 Hz. The EPR investigations showed that after fs-laser irradiation an increase of the signal with respect to the as-deposited value was observed, indicating a higher concentration of defects in the substrate. The increase of the signal was correlated with a similar increase in the leakage current values measured through these films.

Resume : The radiation-resistant oxide insulators (Al2O3, Y3Al5O12, MgAl2O4 etc) are important materials for applications in fusion reactors, e.g. in optical windows. It is very important to predict/simulate a long-time defect structure evolution including thermal defect annealing after irradiation. For further prediction of the radiation stability of materials is necessary extract main kinetic parameters - interstitial migration energy Ea and diffusion pre-exponent D0. Primary radiation defects in ionic solids consist of Frenkel defects—pairs of anion vacancies with trapped electrons (F-type centers) and interstitial ions. The thermal annealing is controlled by the interstitial ion migration, whose mobility is much higher than that of the F centers. The theory (how to extract from experimental data the migration energy of interstitials and pre-exponential factor of diffusion) was developed and applied to irradiated insulators in our recent study [1,2]. It was showed that the correlation of these two parameters satisfies the so-called Meyer–Neldel rule (MNR) [2] observed more than once earlier in glasses, liquids, and disordered materials, but not in the irradiated materials. We analyzed here the available experimental kinetics of the F-type center annealing for two different ionic solids: neutron/ion-irradiated Al2O3 (sapphire) [1,2] and ion-irradiated Y3Al5O12 (YAG) [3]-- both wide gap insulating materials but with different crystalline structures. We demonstrated that in sapphire upon an increase of radiation fluence, both the migration energy and pre-exponent are decreasing, irrespective of the type of irradiation. This is MNR with normal dose dependence. For YAG we have confirmed MNR, but the dose dependence is inverse. We discuss the cause of this phenomenon. Thus, in this study, we demonstrated that the dependence of defect migration parameters on the radiation fluence plays an important role in the quantitative analysis of the radiation damage of real materials and cannot be neglected. [1] V.Kuzovkov, E.Kotomin, A.Popov, R.Villa, Nucl. Inst. Meth. B 374, 107 (2016). [2] E.Kotomin, V.Kuzovkov, A.Popov, J.Maier, R.Vila, J.Phys.Chem.A,22,28 (2018). [3] M. Izerrouken, A. Meftah, M. Nekkab, Nucl. Instr. Meth. B 258, 395 (2007).

Resume : In this work, advanced characterization techniques such as APT and TEM were combined to investigate how ion irradiation conditions and prior heat treatment influence the evolution of the microstructure of a new heat of the DIN 1.4970 alloy (a 15-15Ti austenitic stainless steel). First, different ageing heat treatments were explored, with a particular focus on precipitation and minor element partitioning. Heat treatment conditions had only a modest influence on the TiC precipitate population; the most important considerations were avoiding recrystallization at high temperature and minimizing intergranular M23C6 formation at low temperature. Second, Fe ion irradiations were performed on heat treated and non-heat-treated material up to 40 dpa for 3 different temperatures. The resulting materials were studied by various TEM techniques, APT and nanoindentation, with a focus on Frank loop populations, precipitates and radiation induced segregation (RIS). Small TiC clusters were stable down to at least 450 °C under ion irradiation, regardless of prior heat treatment. Heat treated material showed a lower TiC number density and coarser particles, but an increased volume fraction of RIS induced phases compared to non-heat treated material. This may be correlated with reduced swelling resistance. Overall, an improved understanding of the role of Ti and MC precipitates in the microstructure during irradiation was established, which may lead to more targeted neutron irradiation experiments as well as general alloy improvement strategies in the 15-15Ti alloys.

Resume : Magnesium aluminate spinel single crystals and optical ceramics are important optical materials for the use in harsh radiation environment and many other applications. The kinetics of thermal annealing of the basic radiation-induced electron centers (F and F+ - two or one electron in the field of an oxygen vacancy) as well as complementary hole-containing ones (a set of the so-called paramagnetic V-centers) in Mg-Al spinel with different ratio of stoichiometry irradiated by ~1-MeV fast fission neutrons, 100-keV protons, and 0.23 GeV Xe ions is analyzed in terms of diffusion-controlled bimolecular reactions. Properties of single crystals and optical ceramics with different grain sizes exposed to irradiation of different type/fluence are compared. The EPR spectra recorded by an X-band spectrometer Bruker ELEXYS-II E500 and the optical absorption bands were taken as the measure of the radiation-induced V- and F-type centers, respectively [1]. The annealing of radiation damage was performed in a stepwise regime. It is demonstrated that both ceramics and single crystals, as well as heavy ion irradiation show qualitatively similar kinetics but the major impact on the effective migration energy Ea and pre-exponent D0 is caused by the radiation fluencies. These two parameters are strongly correlated and their values increase considerably with radiation fluence. This effect is discussed in terms of the Meyer-Neldel rule known in chemical kinetics of condensed matter [2]. Recommendations on the enhancement of spinel radiation tolerance are suggested. The results for irradiated spinel samples are compared with those for sapphire, MgO and other radiation-resistant materials. [1] A. Lushchik et al., Nucl. Instrum. Meth. B 435 (2018) 31-37. [2] E.A. Kotomin, V.N. Kuzovkov, A.I. Popov, R.Villa, Nucl. Instum. Meth. B 374 (2016) 107-110.

Resume : The most used nuclear fuel in the world is a UO2 based ceramic. During operation, fissions generate fission products creating heat that is removed through the pellet surface towards the primary circuit water. This heat flow creates a thermal gradient inside the pellet. Because UO2 is a non-stoichiometric compound, this thermal gradient induces the redistribution of oxygen along the pellet radius. We will explain how this oxygen redistribution can induce the formation of a hypo-stoichiometric area in the pellet centre and a hyper-stoichiometric area outside this centre. We will then discuss our current knowledge about the structural transformations that are induced in UO2 crystalline structure by both hypo and hyper-stoichiometry. In hyper-stoichiometric UO2, the formation of clusters made of interstitial oxygen accommodates the excess in oxygen. These clusters named cuboctahedra are associated to new cationic crystalline sites, which could be preferential incorporation sites for some fission products. In hypo-stoichiometric UO2, stacking faults were evidenced but their exact crystallographic nature is not known. The incorporation of Lanthanide fission products in hypo-stoichiometric UO2 can lead to formation of a so called “miscibility gap”, which will also be discussed. These crystallographic statements are a driving force for more investigations on the behaviour of non-stoichiometric

Resume : Zirconium alloys have been used as fuel cladding materials in light water reactors (LWR) since the 1960s due to their low neutron absorption cross section. However, it has been reported that zirconium alloys have some disadvantages in accident conditions such as loss of coolant accident (LOCA). As a LOCA occurs, cladding temperature increases rapidly, which causes abrupt increase of oxidation rate. One of the most critical oxidation behaviors that can occur is breakaway oxidation. It is known that zirconium alloys are vulnerable to breakaway oxidation during oxidation at 800 and 1000 ºC. So far, many researches have been carried out to analyze the breakaway oxidation that occurs at 1000 ºC. However, a few studies have been performed to characterize the breakaway oxidation that occurs at 800 ºC. Hence, in this study, aspect of breakaway oxidation that occurs on a zirconium alloy cladding tube was analyzed after high-temperature steam oxidation tests at 800 ºC. It was found that breakaway oxidation at 800 ºC proceeds as radial cracks are formed in the oxides, and these radial cracks caused local excessive oxidation. Comparing to the breakaway oxidation that occurs at 1000 ºC, at which breakaway oxidation proceeds as circumferential cracks are formed in the oxides, the microstructural aspect during breakaway oxidation at 800 ºC is clearly different. Analyzing the reason of this phenomenon will enhance our understanding of breakaway oxidation, and a new method improving accident resistance of zirconium alloy cladding can be developed in the future.

Resume : One of the promising technologies for recycling spent nuclear fuel is dissolution it in chloride melts, which are resistant to extreme conditions, with including of radwaste into insoluble mineral-like compounds. For example in molybdates and tungstates with the structure of mineral scheelite (CaWO4), which are studied as forms of binding such fission products as alkaline earth and rare earth elements. In the present work was developed technology concentration of Sr, Ba and Nd from the molten LiCl-KCl in mineral form; stability of such phases in the melt and kinetics of dissolving the formed precipitates at T = 450, 500, 600 °C were studied. Samples were obtained by us using the precipitating processes in aqueous systems or in melts and characterized by X-ray diffraction (XRD). The alkaline earth and rare earth element transition into the solid phase from both of liquid mediums is ~ 98 - 99 %. Mineral phase remain stable in contact with melt LiCl-KCl to T = 600 °C, t = 10 h. An increase in temperature increases the solubility of obtained compounds in 1.5 - 2 times. This study has been carried out with the support of the Russian Science Foundation (Grant No. 16-13-10464).

Resume : The increasing safety of radwaste is a key element in the development of atomic energy. The basic factors: a scientifically-based choice of the structure of compounds and the technology of forming of ceramics. We obtained and characterized ceramics with the structures: fluorite, garnet, pollucite, monazite, kosnarite, langbeinite, scheelite. The achieved under extreme conditions: do not decompose at T 1300-1400 °C, to thermal stress, to radiation (γ-radiation, accelerated xenon ions and alpha particles), in water systems: minimum leaching rates 10^(-6) g/(cm^2·Day). To improve these service characteristics, we used the method of sintering ceramics SPS. The SPS has important advantages for nuclear materials compared to traditional methods: production of ceramics with a relative density of more than 99.9 % in very short times (3-10 min). They contribute to increasing chemical stability and lowering release of radionuclides into the gas phase. Such tests of above mentioned ones confirmed the correctness and feasibility of such concept. It can be concluded that this new method will be used in technology centers in the world for the production of ceramic materials for nuclear energy. The work was financially supported by RSF. №16-13-10464.

Resume : A newly developed clamping system for ceramic specimen is described in the present report. By using this system, the uniaxial tensile loading was first successfully applied up to fracture (950 MPa) on a SiCf/SiC minicomposite. A minicomposite is a 1D model composite. It consists of a SiC-based fiber tow embedded in a SiC matrix, which has a diameter of about 0.5 mm. Afterwards, the irradiation creep properties of SiCf/SiC minicomposite were studied. In-situ creep was performed in an in-beam creep device under uniaxial tensile stresses from 50 to 400 MPa during homogeneous helium irradiation. Homogenous irradiation was carried out by helium implantation with energies varying from 0 to 45 MeV. The displacement dose rate was 1.25e-6 dpa/s. The average temperature was controlled to 700 and 900°C within ±10°C. Irradiation creep compliance of minicomposite was measured to be 2.6e-5 and 4.32e-5 1/(dpa*MPa) at 700°C and 900°C, respectively.　

Resume : A grid-to-rod fretting (GTRF) is well known as one of the most frequent failure mechanisms of nuclear fuel rod in operating PWRs. GTRF damages could be accelerated when the contact force is loosened during operations. In this case, both fuel rod and grid are gradually oxidized in reactor condition, and Zr oxides are formed on their contact surfaces. In this study, effect of the surface oxide of both rod and grid on the fretting wear behaviors was evaluated. Zr-based rod and grid samples were exposed to high-temperature water (360°C) up to 360 days for preparing pre-oxidized samples with different oxide thickness. Both oxide thickness of the rod and grid samples gradually increased with the exposed time. It was found that the oxidation rate of the grid was faster than that of the rod due to the difference in the microstructure and manufacturing process. In the results of fretting wear tests, the wear volume and maximum wear depth of the pre-oxidized rod were dramatically decreased with the increase of the oxide thickness. It was concluded that the oxide thickness is one of the important factors for determining the fretting wear resistance of fuel rod.