Radiation-induced Precipitation Research Articles

Identification of nano-sized precipitates in CLAM steel induced by 3.5 MeV Fe13+ ion irradiation

Nano-sized precipitate phases in normalized-and-tempered CLAM steel (HEAT-0912) before and after 3.5 MeV Fe13+ ion irradiation at 400 °C to 0.61, 1.29 and 2.77 dpa were studied using transmission electron microscope and high-resolution transmission electron microscope. Six types of nanoprecipitate phases based on Cr23C6, TaC, Ta3C2, VC, Cr7C3, and V2C were identified in unirradiated steel. During irradiation, a large number of nanoprecipitates uniformly distributed in the matrix occurred, and the number density of nanoprecipitates initially increased and subsequently decreased with increasing the irradiation dose. Nanoscale Cr-rich M23C6, V-rich MX and M2X phases but no nanoscale phases based on TaC, Ta3C2 and Cr7C3 were identified in the steel irradiated up to 2.77 dpa. After irradiation, five types of irradiation-induced nanoscale new phases were identified in the irradiated steel. These include Fe-rich M3X2 (Cr3C2 types I and II) carbonitrides with the same simple orthorhombic lattice and totally different lattice parameters (approximately a/b/c = 1.146/0.552/0.2821 nm and 0.285/0.925/0.696 nm), Fe-rich M2C, M3C and M5C2 carbides, but the occurrence of which under irradiation varied with the irradiation dose. Fe-rich M3X2 (Cr3C2 type I) and Fe-rich M5C2 phases were identified in the steel irradiated up to 1.29 dpa and only to 2.77 dpa, respectively. Fe-rich M2C and M3C in addition to Fe-rich M3X2 (Cr3C2 type II) phases were identified in the steel irradiated up to 2.77 dpa. The irradiation-induced Fe-rich M3X2 (Cr3C2 type I & II), M5C2 and M2C precipitates appear to be identified, for the first time, in reduced activation ferritic/martensitic steels including CLAM steel. The formation mechanism of these irradiation-induced precipitate phases is also discussed.

Prediction of radiation embrittlement of VVER-1000 RPV welded seam material at lifetime extension up to 60 years and more

The correctness of the model assumption about additivity of the contributions of hardening (formation of dislocation loops and radiation-induced precipitates) and non-hardening (formation and accumulation of grain boundary segregations) mechanisms to the final radiation embrittlement of VVER-1000 RPV weld materials has been verified. The paper presents the results of the experiment to obtain predicted values of the critical embrittlement temperature of weld metal with nickel content of 1.59–1.88% using accelerated irradiation in the IR-8 research reactor from states characterized by different starting level of grain boundary segregation. The procedure of accounting for the flux effect for the values of critical embrittlement temperature obtained using accelerated irradiation has been proposed. The conservativity of the weld metal radiation embrittlement dependence has been verified based on the modal assumption of additivity of the contributions of the strengthening and non-strengthening mechanisms up to the values of fast neutron fluence ~75–1022 neutrons/m2.

He bubbles inhibition and abnormal hardening in GH3535 alloy by niobium element addition

GH3535 and 2 wt.% Nb-modified GH3535 (Nb-GH3535) alloys were irradiated with 1.2 MeV He ion at fluences of 1 × 1016, 3 × 1016 and 1 × 1017 ions/cm2 at 750 °C. Microstructural changes, including the formation of He bubbles and superlattice phases, as well as the irradiation-induced hardening, were investigated by TEM and nanoindentation. Notably, Nb-GH3535 exhibited a smaller size and lower number density of He bubbles, especially at higher fluences, indicating the inhibitory effect of Nb on bubble evolution. First-principles calculations revealed that the presence of Nb inhibits the combination of He atoms and vacancies, reducing their diffusion rate and consequently impeding the nucleation and growth of He bubbles. Furthermore, the addition of Nb facilitates the irradiation-induced precipitation of the superlattice phase, leading to a higher degree of total hardening in Nb-GH3535 alloy.

Unveiling the characteristics of the residual point defects of collision cascade in Zr-xNb binary system: A molecular dynamics study

Zirconium-based alloys are commonly used as fuel cladding materials for light-water reactors. As an alloying element, niobium can further improve the performance of Zr alloys. However, the mechanism of the Nb effect on the irradiation damage (e.g., the irradiation-induced precipitation of Nb-rich clusters) of the alloy is still unclear and should resort to atomic-scale modeling. In this work, we unveiled the characteristic of the residual point defects produced at the end of the collision cascade caused by primary knock-on atoms (PKAs) in the Zr-xNb binary alloy through many molecular dynamics simulations. The effects of PKA directions (<0001> and <101¯0>), PKA energies (2 keV, 5 keV, 10 keV, 20 keV, 30 keV, and 50 keV), temperatures (300 K, 600 K, and 800 K), and the initial Nb contents in the alloy (cNb = 0 at.%, 0.5 at.%, 1 at.%, and 2 at.%) on the remaining Frenkel pairs, the displaced atoms, the composition of these point defects, the size and number of clusters, and the composition of the clusters with different-size have been studied. Several concepts were clarified from these simulation results. A special focus of this work is on the Nb proportion in the point defects and their clusters. It is worth emphasizing that the interstitial-Nb proportion in the interstitials and their clusters is far greater than the initial Nb content in the alloy, which can be the nucleate center of Nb-rich precipitation particles. This distinct feature may be solving the experimental puzzle of radiation-induced Nb precipitates/clusters of Nb-containing Zr alloys.

Phase Formation Features of Reactor Pressure Vessel Steels with Various Ni and Mn Content under Conditions of Neutron Irradiation at Increased Temperature

In this paper the phase formation and mechanical properties of VVER-type reactor pressure vessel (RPV) steels with various Ni (1.57–5.95 wt.%) and Mn (0.03–0.76 wt.%) content after neutron irradiation up to fluences in the range of (53–120) × 1022 n/m2 at 400 °C were studied. The possibility of carbonitride formation under these irradiation conditions is shown. In case of sufficient Ni (&gt;1.5 wt.%) and Mn (&gt;0.3 wt.%) content formation of Ni-Si-Mn precipitates is observed. Their chemical composition is close to G-phase and Γ2-phase and differs from that of radiation-induced precipitates in VVER-1000 RPV steels. This indicates the prerequisites for thermally conditioned mechanism of Ni-Si-Mn precipitates formation and growth at 400 °C enhanced by irradiation. It is also shown that the optimized steel manufacturing technology coupled with an ultralow Mn content (≤0.03 wt.%) in steel with increased up to 5.26 wt.% Ni content facilitates suppressing the Ni-Si-Mn precipitates and carbonitrides formation. This, in turn, reduces the contribution of the hardening embrittlement mechanism and, correspondingly, facilitates high radiation resistance of the steels with ultralow Mn content at the increased irradiation temperature (400 °C).

Radiation-induced structure of austenitic steels with different nickel content under neutron irradiation in SM-3 and BOR-60 reactors

Comparative studies of the radiation-induced structure of austenitic steels with a nickel content of 10, 20 and 25 wt.%, irradiated sequentially in the SM-3 and BOR-60 reactors, as well as to higher damaging doses in the BOR-60 reactor, have been carried out. The phase composition, dislocation structure, pores, and radiation-induced segregations at grain boundaries were studied by high-resolution analytical methods of transmission electron microscopy, scanning electron microscopy, and atomic probe tomography. The formation of radiation-induced phase precipitates based on nickel has been established, and its volume fraction correlates with the level of radiation-induced segregations, and increases, the higher the nickel content in the steel. The values of barrier strength factors for radiation-induced structural elements in the studied steels are adjusted by calculation and experiment, which makes it possible to determine their contribution to radiation hardening.It is shown that the largest contribution to radiation hardening as a result of neutron irradiation in BOR-60 at high irradiation temperature up to 29 dpa is made by large radiation-induced precipitates of (G + γ') phases. It is shown that with an increase in the damaging dose, the main factor limiting the performance of internal devices will be radiation swelling, since the contribution to the change in properties from radiation-induced phases and radiation defects will not increase due to their density reaching saturation. Steel with 25 wt.% Ni exhibits the lowest level of swelling at high radiation doses, which makes it possible to consider it as a material-candidate for internals for promising VVER reactors with higher temperatures and longer service life.

Radiation-induced precipitates in ferritic-martensitic steels and their radiation resistance effects

Designing high-performance radiation-tolerant materials and understanding the atomistic mechanism of radiation resistance are important topics in nuclear energy structural materials research. In this study, we investigated the atomistic mechanism of the formation of rod-like precipitates in helium-irradiated ferritic-martensitic steels (F-M steels) by positron annihilation spectroscopy (PAS) and transmission electron microscopy with energy dispersive X-ray spectroscopy (TEM/EDX). The results indicated that vacancies tended to accumulate near dislocation lines to form complexes in the reduced activation ferritic-martensitic (RAFM) steel. Moreover, Fe and Cr atoms diffused toward these complexes and eventually depleted therein, which was not conducive to the formation of rod-like precipitates. Conversely, impurity atoms such as C atoms were found to segregate near dislocations in the Y-bearing oxide dispersion strengthened (Y-ODS) steel. When Fe and Cr diffused toward dislocations, they exchanged positions with C atoms until Fe and Cr exceeded their saturation solubilities and precipitated on the glide plane. We also analyzed the distribution of vacancy defects and bubbles/voids in the four irradiated materials and discussed the radiation resistance mechanism of the precipitates. The results of this study are significant in demonstrating a microscopic mechanism of radiation-induced precipitates to swelling resistance in materials. The raw/processed data required to reproduce these findings cannot be shared online at this time due to technical limitations, but they are available upon contacting the corresponding author. Materials exposed to extreme environments have received great attention recently in the context of advanced energy systems, defense, and aerospace technologies. Specifically, radiation-damage-tolerant materials are in great demand for the safe operation and advancement of nuclear energy and accelerator systems. Designing high-performance radiation-tolerant materials and understanding the atomistic mechanism of radiation resistance are important materials science topics. In this study, we reported the contribution of a microscopic mechanism of radiation-induced precipitates to swelling resistance in helium-irradiated ferritic-martensitic (F-M) steels/alloys. Significant findings included the formation of large areas of rod-like precipitates, which exhibited swelling resistance effects in steels subjected to helium irradiation and a possible microscopic mechanism for their formation. • Dense rod-like precipitates were observed in samples irradiated with high-dose helium ions. • The Cr and C contents were important factors affecting the formation of rod-like precipitates. • We clarified the atomistic mechanism of the formation of rod-like precipitates in the steels. • High-density rod-like precipitates as "radiation-damage antibodies" reduced void swelling.

Irradiation-induced amorphization of UO2 films prepared by spraying-assisted combustion synthesis

Spraying-assisted combustion synthesis with uranyl nitrate − acetylacetone − 2-methoxyethanol solutions was used to prepare UO2 films on an aluminum alloy substrate. The tuning of the spraying parameters and annealing temperatures allowed the preparation of UO2 films with thicknesses varying from 10–300 nm and 5–10 nm UO2 grain size. High-resolution electron microscopy and X-ray photoelectron spectroscopy showed that increasing the annealing temperature promotes Mg diffusion from the substrate into the films. The incorporation of Mg reduced the overall crystallinity of the films. The irradiation with Ar2+ ions (1.7 MeV energy and a fluence of 2 × 1016 ions/cm2) did not degrade the quality of the films. However, the Mg content significantly influenced the irradiation-induced restructuring of the UO2 films. Irradiated films with low or no Mg content exhibit high crystallinity, and the UO2/Al interfacial layer becomes highly porous. Films with higher Mg content are mostly amorphized after irradiation. The origin of irradiation-induced amorphization was related to the formation of MgyU1-yO2±x solid solutions. Chemically complex, pore-free, and amorphous Mg-Al-O film/substrate interfacial layers enable continuous Mg diffusion during irradiation. Furthermore, the gradual increase in Mg amounts triggers irradiation-induced precipitation of a crystalline MgO-rich phase within the amorphous films.

A model experiment on nano-particle stability in 12Cr-ODS steel using high-resolution high voltage microscopy

In this work, the microstructure stability of 12Cr-ODS ferritic steel under irradiation is studied using a high-resolution high voltage electron microscope up to 1.5 dpa at 300 and 500 °C. Coherent YTiO3 nanoparticles (NPs) undergo coarsening during electron irradiation and the coarsening rate is positively correlated to irradiation temperature (the third power of NPs size is proportional to the irradiation dose). He pre-implantation can significantly reduce the coarsening rate. A model coupling Ostwald ripening and irradiation enhanced diffusion is used to explain this coarsening. The fitting results between NPs size and irradiation dose show that pre-implanted He causes a lower diffusion of Y in 12Cr-ODS steel compared with the non-implanted sample. In response to the irradiation, plate-like radiation-induced precipitates appear which result in diffraction streaks in fast fourier transform (FFT) pattern. Precipitations are paralleled to [100] or [110] direction of Fe matrix and located near the interface between NPs and matrix. The hypothesis that coherent Cr precipitate along NPs/matrix interface caused diffraction streaks is proposed. Besides, He pre-implantation can also delay this precipitation.

Influence of titanium on the clustering of vacancy, rhenium and osmium in tungsten-titanium alloys: First-principles study

The formation of point defects including vacancies and interstitials and transmutation elements rhenium (Re), osmium (Os), tantalum (Ta), etc., has been reported to seriously degrade the mechanical properties of tungsten in a fusion environment. In this work, the aggregation behavior of between vacancy and rhenium or osmium, and the effect of Ti as the interstitial atoms on the transmutation element (Re or Os) are studied using first-principles calculation. We demonstrate that the addition of Ti or Re/Os will promotes the clusters of vacancy in the W-Ti alloys. Here, the concentration balance of the vacancy and Ti plays an important role in the aggregation of Re/Os atoms in W-Ti alloys. Furthermore, we found that doping of Ti induced vacancy migration, this plays a positive role in the recovery of the interstitial with the vacancies. Interestingly, the presence of Re/Os similarly promote the migration of vacancy to Ti sites, but Ti makes the migration of vacancy to Re/Os harder. Besides, we found the most stable configuration with <11h> mixed dumbbell in W-Ti (h∼0.50), Ti-Os (h∼0.34) and Ti-Re (h∼0.50) interstitial, in which The Ti-Re/Os interstitial with most stable <11h> dumbbell has a strong attractive to Re/Os atoms. Consequently, the formation of the Ti-Re/Os interstitial greatly promotes the Os/Re to clustering, and further increase the probability of radiation-induced precipitation in W-Ti alloys.

Radiation-Induced Precipitation of Cr-Rich Clusters in Fe-Cr Alloys Containing C and N Solutes: First-Principles Modelling and Experimental Observations

Radiation-Induced Precipitation of Cr-Rich Clusters in Fe-Cr Alloys Containing C and N Solutes: First-Principles Modelling and Experimental Observations

Enhanced irradiation tolerance of Fe30Cr25Ni20Co15Mn10 high-entropy alloy via nanotwin boundaries

• There is no nanoscale precipitation and irradiation-induced hardening in Fe 30 Cr 25 Ni 20 Co 15 Mn 10 HEA at a high dose irradiation (∼ 70 dpa). • The HEA presents a very low swelling rate of ∼ 0.002%/dpa. • The RIS is observed for the first time at nanotwin boundaries in the HEA after irradiation. • The irradiation-induced segregation level is obviously suppressed in the HEA due to the sluggish atom diffusion. High-entropy alloys (HEAs) have been considered as a promising candidate for nuclear engineering due to their intriguing irradiation tolerance. The microstructure stabilities, including the radiation-induced precipitation (RIP), swelling, and segregation (RIS) in HEAs, are of great significance for applications. In this paper, the Fe 30 Cr 25 Ni 20 Co 15 Mn 10 HEA was irradiated by high-energy Au 2+ ions with a peak dose of ∼ 70 displacements per atom (dpa). The formation of voids, nanoscale precipitation, segregation, and mechanical properties before and after irradiation was characterized carefully, using scanning/transmission electron microscopy (S/TEM), atom probe tomography (APT), and nanoindentation. The results indicate that the HEA presents a very low swelling rate of ∼ 0.002%/dpa. After irradiation, no nanoscale precipitation is observed, confirming the excellent stability of the HEA microstructure. The RIS is observed for the first time at nanotwin boundaries in the HEA after irradiation. The elements, Co and Ni, are enriched slightly, while Fe, Cr, and Mn depleted on the nanotwin boundaries after irradiation. The irradiation-induced segregation level is obviously suppressed in the HEA due to the sluggish atom diffusion. The hardness remains almost unchanged after irradiation. The nanotwin boundaries act as defect sinks to absorb irradiation-induced vacancies and interstitials, reducing the swelling rate and relieving hardening. The low swelling, weak RIS level, and almost no hardening after a high dose irradiation indicate that the Fe 30 Cr 25 Ni 20 Co 15 Mn 10 HEA possesses an outstanding irradiation tolerance.

Impact of the local microstructure fluctuations on radiation-induced segregation in dilute Fe-Ni and Ni-Ti model alloys: A combined modeling and experimental analysis

From a systematic atom probe tomography (APT) characterization of the radiation-induced segregation (RIS) in dilute Fe-Ni and Ni-Ti model alloys, we highlight fluctuations of the solute local concentration up to the scale of the APT specimens. We deduce the RIS at dislocation loops from a solute diffusion equation, that is solved at steady state, within the Voronoi’s volume occupied by a single loop. From a statistical sampling of the Voronoi’s volume and the dislocation loop radius modeled after the characterization of the microstructure by transmission electron microscopy, we provide the full RIS distribution. The present statistical approach of RIS demonstrates that the fluctuation of local solute concentrations in Fe-Ni and Ni-Ti mainly results from the dispersion in size and density of the dislocation loop population. Besides, we highlight the impact of the post-treatment parameters used in the APT protocol on the extracted RIS profiles. In Ni-Ti alloys, the simulated Ti-depletion profiles are in very good agreement with the measured ones. Furthermore, the dispersion of the loop radius and density is shown to play a critical role on the fluctuations of the Ti local concentration. In Fe-Ni, the identification of discrepancies between the simulated Ni-enrichment profiles and the measured ones provides a signature of additional operating mechanisms of the solute redistribution, such as radiation-induced precipitation.

Radiation-induced swelling and precipitation in Fe++ ion-irradiated ferritic/martensitic steels

The radiation responses of two newly developed ferritic/martensitic steels, FC92-B and -N, were tested in comparison to reference alloys HT9 and Gr.92. Ion irradiations on the steels were performed up to 480 dpa at 475°C using 3.5-MeV Fe++ ions with a helium pre-implantation of 1 appm/dpa. Void swelling and M2X precipitation were characterized using FE-TEM and EDS. Swelling resistance was the greatest in FC92-N, which showed suppressed void nucleation and growth. The swelling rate in FC92-N was determined as 0.005 %/dpa, indicating that FC92-N did not reach the steady-state swelling regime with void nucleation behavior. The least swelling-resistant alloy was HT9 with a swelling rate of 0.048 %/dpa. Cr-rich carbide, M2X, was observed in only 9Cr-FC92 series; however, its formation did not depend on radiation damage. This exceptional M2X evolution in FC92 series may be attributed to B and N alloying, which resulted in suppressed M23C6 carbide formation during metallurgical production and sequentially high C contents in the alloy solution of FC92 series. A narrower range (800 nm) of M2X evolution compared to that of cavity formation (1,000 nm) indicates that radiation-induced precipitation (RIP) is more sensitive to the injected interstitial effect. Precipitation-induced Cr depletion and preferential interstitial outward sinking to the free surface synergistically modified local chemical composition before void evolution and led to double-peak swelling by locally forming a low-alloyed zone. This study provides the first experimental evidence that RIP modifies the swelling–depth profiles and in turn, determines double-peak swelling in ion-irradiated steels.

Development of a solute and defect concentration dependant Ising model for the study of transmutation induced segregation in neutron irradiated W–(Re, Os) systems

In this study, radiation-induced precipitation of transmutation products is addressed via the development of a new solute and vacancy concentration dependant Ising model for the W–Re–Os system. This new model includes interactions between both Os and Re atoms with vacancies, thus facilitating more representative simulations of transmutation in fusion reactor components. Local solute concentration dependencies are introduced for the W–Re, W–Os and Re–Os pair interactions. The model correctly accounts for the repulsion between small clusters of vacancies and the attraction between larger clusters/voids, via the introduction of local vacancy concentration dependant interaction coefficients between pairs of atoms and vacancies. To parameterise the pair interactions between atoms and/or vacancies, the enthalpy of mixing, ΔH mix, for various configurations and solute/defect concentrations, was calculated using density functional theory, within 6 binary systems: W–Re, W–Os, Re–Os, W–vacancy, Re–vacancy and Os–vacancy. The new energy model was implemented into the SPPARKS Monte Carlo code, and successfully used to predict the formation of voids decorated with Re and Os solute atoms. Analysis suggests that there is a strong thermodynamic tendency for Os to bind to these voids with a comparatively weaker binding from Re atoms. The binding energies of various solute/vacancy clusters were calculated and showed that Re and Os solute atoms tend to stabilise small clusters of vacancies, increasing the attractive binding energy between the constituents.

Microstructure-Dependent Rate Theory Model of Radiation-Induced Segregation in Binary Alloys

Conventional rate theory often uses the mean field concept to describe the effect of inhomogeneous microstructures on the evolution of radiation induced defect and solute/fission product segregation. However, the spatial and temporal evolution of defects and solutes determines the formation and spatial distribution of radiation-induced second phase such as precipitates and gas bubbles/voids, especially in materials with complicated microstructures and subject to high dose radiation. In this work, a microstructure-dependent model of radiation-induced segregation (RIS) has been developed to investigate the effect of inhomogeneous thermodynamic and kinetics properties of defects on diffusion and accumulations of solute A in AB binary alloys. Four independent concentrations: atom A, interstitial A, interstitial B, and vacancy on [A, B] sublattice are used as field variables to describe temporal and spatial distribution and evolution of defects and solute A. The independent concentrations of interstitial A and interstitial B allow to describe their different generation rates, thermodynamic and kinetic properties, and release the assumptions of interstitial generation and sink strength used in the conventional rate theory. Microstructure and concentration dependent chemical potentials of defects are used to calculate the driving forces of defect diffusions. With the model, the effects of defect chemical potentials and mobilities on the RIS in polycrystalline AB model alloys have been simulated. The results demonstrated the model capability in predicting defect evolution in materials with inhomogeneous thermodynamic and kinetic properties of defects. The model can be extended to materials with complicated microstructures such as a wide range of grain size distribution, coating structure and multiphases as well as radiation-induced precipitation subject to severe radiation damage.

Magnetic investigations on irradiation-induced nanoscale precipitation in reactor pressure vessel steels: A first-order reversal curve study

We have investigated magnetic properties for reactor pressure vessel model alloys with variable Cu contents, subjected to neutron irradiation up to a fluence of 9 × 10 19 n cm − 2 . Unlike a monotonic increase of microhardness with neutron fluence, the major-loop coercivity decreases at a higher fluence and the decrease becomes larger for the alloy containing a higher amount of Cu. The measurements of first-order reversal curves (FORCs) for the high-Cu alloy show that the position of the FORC distribution peak shifts toward a lower coercivity just after neutron irradiation, followed by a slight increase, associated with the broadening along both the coercivity and interaction field axes. The results can be explained by the enhancement of magnetic inhomogeneity in a matrix due to Cu precipitation and an increasing magnetostatic interaction between local magnetic regions with different coercivity. The magnetic method using FORCs can be a possible technique which provides in-depth information on microstructural changes due to neutron irradiation, which is not obtained by measurements of a conventional major hysteresis loop.

Nanocluster evolution and mechanical properties of ion irradiated T91 ferritic-martensitic steel

Ion irradiation has been used to investigate the radiation-induced precipitation of nanoclusters and changes in mechanical properties of commercial-grade T91 ferritic-martensitic steel irradiated with Fe4+ ions up to 4.10 dpa at 301 – 311 °C. Atom probe tomography was used to analyse the microstructure segregation and cluster formation, while nanoindentation was used to measure the change in mechanical properties. At 0.12 dpa, Si/P-rich clusters formation is observed. At 1.76 dpa, Mn, Ni and Si -rich precipitates (MNSP) were observed with a composition range that is distinctly different than the typically cited G-phase. The MNSP number density was similar to that from prior neutron irradiation studies at similar temperatures. Segregation of such species to dislocation loops and lines was also discussed. The hardness increased with radiation dose up to measured 1.83. Comparisons were made between the observed microstructural and mechanical property changes as a function of ion irradiated dose. The use of APT and nanoindentation, when applied in tandem, have shown to provide an insight into how radiation-induced microstructural effects can explain the observed changes in mechanical properties.

Microstructure of zirconium fuel claddings: TEM and EBSD studies of as-received and ne-utron-irradiated materials

Zirconium fuel claddings act as a first barrier against release of fission products during nuclear power plant operation and interim storage of the spent fuel. During the reactor operation, cladding tubes are exposed to different stress level at elevated temperatures and neutron irradiation in corrosive environment. It causes a material degradation by corrosion, cladding embrittlement by hydrides and radiation-induced damage or radi-ation growth and creep of the fuel rods. The irradiation damage effects mainly contribute to the loss of material ductility. In our study, microstructure of as-received (non-irradiated) Zr-alloys used in LWR (Zr1Nb, Zr-1Nb-1.2Sn-0.1Fe, Zr-1.5Sn-0.2Fe-0.1Cr) were examined by electron microscopy methods. Transmission electron microscope (TEM) was used to describe the microstructure of claddings used in different reactor conditions and identify the radiation-induced damage, which is presented on Zr1Nb irradiated to one standard campaign in the VVER-1000 active zone. Following Electron Backscatter Diffraction (EBSD) method on transparent foils complements the TEM results in larger area, i. e. by grain size and orientation or analysis of local misorienta-tion after irradiation. Radiation-induced damage was observed in Zr1Nb metallic matrix as type disloca-tion loops, presence of radiation-induced precipitates or partial amorphization of the secondary phase particles. EBSD method showed no changes in crystallographic orientation, but a local increase of dislocation density can be affected by neutron irradiation.

(S)TEM/EDS study of native precipitates and irradiation induced Nb-rich platelets in high-burnup M5®

We have investigated microstructure and microchemistry of precipitates and <c> dislocation loops in high-burnup M5® using (scanning) transmission electron microscopy ((S)TEM) equipped with energy dispersive X-ray spectroscopy (EDS). Two (S)TEM lamellae were made by cryo-FIB from the same cladding sample. The Nb-rich native precipitates were found in the metal, in the suboxide and in the oxide. Upon diffraction analysis, most of the Nb-rich native precipitates in the metal matrix remain as β-Nb phase, while no β-Nb native precipitates were found in the oxide. Nearby the oxide and metal (O/M) interface, the native precipitates in the oxide were already oxidized into t-NbO2 phase. At further distance away from the O/M interface, the Nb-rich native precipitates were gradually oxidized and became amorphous. Besides the native precipitates, Nb-rich irradiation induced precipitates (IIPs) were found in the metal matrix. Using g = <0002> vector for imaging, the length of the IIPs was aligned with <c> dislocation loops. However, no Nb segregation to the <c> dislocation loops themselves was observed. For the first time, we report that IIPs indeed exist in the oxide but only within about 1.5 µm away from O/M interface. However, the oxidation state of the IIPs in the oxide is still unclear. The presence of both native precipitates and IIPs in the oxide may indicate the Nb concentration in the oxide solid solution remain low nearby the O/M interface, which may explain the reduced corrosion kinetics of in-pile M5®. On the other hand, no IIPs were observed in the oxide at further distance and this may indicate that they have eventually dissolved back into the oxide. A few mechanisms related to IIPs stability in the oxide are presented.