Irradiation-induced Microstructural Evolution Research Articles

Measurement of atomic dispersion on buried Ag nanoparticles decomposed by Ne[formula omitted] ion irradiation

The effects of room temperature 200keV Ne+ ion irradiation on Ag nanoparticles (NPs) embedded in a silicon nitride film have been studied via Medium Energy Ion Scattering (MEIS), Scanning Transmission Electron Microscopy (STEM), and the results are discussed in comparison with Monte Carlo TRI3DYN simulations. The experiment was designed to test the mechanisms controlling the room temperature irradiation-induced microstructure evolution of individual NPs and the overall system. We present an innovative MEIS approach capable of quantitatively determining Ag’s content and spatial distribution in solution. The STEM observations demonstrate the nucleation of smaller Ag NPs distributed within the solute field. The TRI3DYN simulations suggest that a fraction of the solute content is reincorporated within the NPs. These results are discussed considering that room-temperature thermal diffusion processes are significantly retarded in silicon nitride substrates, which leads to an interpretation primarily based on the irradiation-induced atomic displacements that render a microstructure evolution driven by the minimization of chemical potential as an elemental thermodynamic force. Our findings are of applied interest concerning the application of ion beams to manipulate NP ensembles.

Experimental emulation of 10B(n, α)7Li reaction-induced microstructural evolution of Al-B4C neutron absorber used in the dry storage of spent nuclear fuel

Al-B4C neutron absorber in spent fuel dry storage could suffer significant radiation damage at elevated temperature (∼150–300 °C) mainly via 10B(n, α)7Li reactions. Several recent studies reported significant bubble formation in the absorber, however only from wet storage environment at near ambient temperature (< ∼90 °C). In this study, we irradiated a commercially widely used Al-B4C absorber of interest utilizing 120-keV helium ion-beam accelerator with twelve different conditions; three different doses (0.01, 0.1, and 1 dpa) combined with four different temperatures (room temperature, 150, 270, and 400 °C). This was to investigate the irradiation-induced microstructural evolution of the absorber in dry storage environment in expedited manner instead of time-consuming neutron irradiation test. The effects of irradiation dose and temperature on helium bubble nucleation and growth were investigated by image analysis on BF-TEM images, which showed a typical tendency of increasing bubble size with increasing temperature at the expense of decreasing bubble number density. Helium bubble morphology at grain boundaries was quite similar with the ones reported from the previous studies on the surveillance coupons used in the wet environment, elliptic helium bubbles around B4C particles. Notable difference was the prominent formation of numerous helium bubbles even from the lowest dose (0.01 dpa) at slightly elevated temperature (150 °C), perhaps due to higher irradiation temperature than that of spent fuel pool. This study may experimentally confirm that significant bubble formation and growth in the absorber could be the case from the early stage of the dry storage, even without the participation of hydrogen produced from the absorber corrosion. Particularly in the case of non-clad Al-B4C absorber, possible B4C particle detachment may need to be concerned since the bubble coalescence was concentrated at the interface between Al alloy matrix and B4C particle which would be microcracked under neutron irradiation.

Modeling of irradiation-induced microstructure evolution in Fe: Impact of Frenkel pair distribution

This study investigated the irradiation-induced microstructure evolution in Fe, employing the Creation-Relaxation Algorithm and different interatomic potentials. The influence of self-interstitial atoms (SIAs), which were either locally or uniformly being distributed during the creation of the Frenkel pairs, was investigated on the evolving microstructure. The spatially localized distribution of SIAs, mimicking the low-energy transfer irradiation conditions, moderated the microstructure development, compared to uniform distribution of SIAs, delaying the nucleation of dislocation for higher irradiation doses. Introducing multiple Frenkel pairs facilitated a cumulative irradiation dose of 5 dpa in large supercells. In small supercells, the accumulation of SIAs led to the formation of an artificially stabilized self-interacting planar interstitial cluster, suggesting a minimum cell dimension of 10 nm for an accurate modeling of microstructure evolution when the development of the dislocation network is of interest. The formation and evolution of the C15 Laves phase structure were explored. The evolving C15 structure developed larger clusters with uniformly distributed SIAs, and their sizes depended on the interatomic potential employed. Finally, a comparison with experimental measurements demonstrated that the density and the average size of interstitial dislocation loops aligned well with those observed in experimentally irradiated ultra-high purity Fe at low and room temperatures.

Microstructural evolution of silicon carbide fiber under argon ion irradiation at room temperature

In current work, irradiation-induced microstructural evolution of the Cansas 3303 SiC fiber was investigated. This fiber was irradiated by 1.8 MeV argon ions at room temperature in the Shanghai Institute of Applied Physics, Chinese Academy of Sciences. The ion fluences were 5 × 1014, 1 × 1015, 2.5 × 1015 and 5 × 1015 ions/cm2 corresponding to the maximum doses of 0.27, 0.54, 1.35 and 2.70 displacements per atom (dpa), respectively. Irradiation effects were characterized by using micro-Raman spectroscopy and transmission electron microscopy. The results showed that the Cansas 3303 fiber consists of nanocrystalline SiC grains, carbon packets and carbon ribbons. Irradiation-induced amorphization accompanied with the decrease of SiC grain size was observed in the fiber with the increase of ion fluences. However, SiC grains still remained in this fiber up to the dose of 2.70 dpa. Besides, the reduction of the quantity and size in carbon packets and the annihilation of carbon ribbons in this fiber were also detected.

Irradiation-Assisted Microstructure Evolution and Mechanical Properties Loss of 310S Welded Joints

In order to reveal the effect of irradiation damage caused by high-level liquid radioactive wastes on the welded joint of the container, the irradiation-induced microstructure evolution and mechanical properties degradation of the 310S stainless steel welded joints were investigated in this study. For this purpose, the 1.3 MeV 60Co and 2 MeV accelerators were used to simulate irradiation experiments on 310S welded joints. The uniaxial tensile tests characterized the specimens' mechanical properties and fracture morphology. The results revealed that elongation was reduced by about 5% of irradiation damage by 60Co, and the fracture morphology shows a large number of secondary cracks. In contrast, the elongation was recovered irradiated by the accelerator, and the fracture morphology showed a large number of dimples. Following the interrupted creep deformation, creep fracture tests were conducted for irradiation specimens. The 60Co irradiation damage significantly decreases the creep resistance, leading to deformation of creep, which is increased to 1.5 times that of those unirradiated specimens. At the same time, the ductility is seriously degraded for the irradiated creep fracture specimens. As a result, the creep fracture strain of 60Co specimens is reduced to 70% of that of unirradiated specimens. Further, ductility reduction was related to the irradiated hardening by 60Co, while Nano-indenter hardness was 5.9 GPa, higher by 44% than the unirradiated specimens. The results are shown in an enrichment of Cr, C and P elements at phase boundaries for 60Co irradiation specimens, while the magnitude of element segregation increased by the accelerator combination irradiation. Finally, the creep cracking analysis results show intergranular cracking was observed on the surfaces of the irradiated specimens, while the M23C6 has a primary relationship with the intergranular cracks. The synergic effect of irradiation promoted damage, and element segregation was the primary cause of the intergranular cracking of the 310S welded joints.

In-situ TEM investigations on the microstructural evolution of SiC fibers under ion irradiation: Amorphization and grain growth

SiC/SiC composites are attractive candidates for many nuclear systems. As reinforcements, SiC fibers are critical to the in-service performance of composites. In this work, the temperature effects on the irradiation-induced microstructural evolution of Cansas-III SiC fibers were investigated using in-situ transmission electron microscopy (TEM). With in-situ 800 keV Kr ion irradiation, at room temperature (RT) the SiC fiber experienced heterogeneous amorphization and became completely amorphous at ∼2.6 dpa. Above the critical temperature of crystalline-to-amorphous (Tc), SiC fibers underwent a simultaneous process of carbon packet disappearance and nano-grain growth at 300 °C and 800 °C. Possible mechanisms were discussed.

TEM characterization of irradiation-induced dislocation loops and voids in ion-irradiated pure chromium

Chromium (Cr) based coatings have been proposed for Zircaloy fuel rods to enhance the accident tolerance in water-cooled reactors. However, the response of Cr to irradiation has received limited attention in prior studies. To get a better understanding of radiation damage in Cr for future nuclear applications, in the present study, the irradiation-induced microstructure evolution of dislocation loops and voids was studied in pure Cr subjected to 2.8 MeV iron ions irradiation at 550 °C (belonging to stage III, intermediate-temperature). TEM observation results show that the distribution of dislocation loops extends the SRIM simulated irradiation depth to 3.5 µm, while voids were only formed in the irradiated region up to a depth of 1.3 µm. A rigorous method with combined consideration of edge-on view and inside-outside analysis was proposed to determine the habit plane. With this method, what has normally considered as an unsafe zone becomes a favored zone for determining the nature of dislocation loops. The nature of dislocation loops was determined including both interstitial and vacancy types with ½<111> Burgers vector in all regions. Within the irradiation region, the loops are approaching half interstitial and half vacancy, while beyond the irradiated region interstitial loops became dominant. The irradiation-induced swelling was quantified by the void size and density. The swellings are mostly corresponding well with the irradiation dose distribution which is simulated by SRIM. The swelling rate close to the peak region is around 0.3%/dpa. Results demonstrate that Cr tends to experience some similarities in irradiation loop evolution to W-based, Mo-based materials, as they are in the same group in the periodic table, despite the discrepancy in the characteristic of loops such as the irradiation temperature-dependent dislocation loop formation was observed as well. The present study is significant not only for supplementing the fundamental knowledge of irradiation damage in Cr but also for the application of Cr coating to the accident tolerant fuel claddings or other nuclear structure materials.

Small-scale mechanical testing and characterization of fuel cladding chemical interaction between HT9 cladding and advanced U-based metallic fuel alloy

Fuel cladding chemical interaction (FCCI) occurred on the interface between the nuclear metal fuel and cladding is the primary cause of cladding wastage, weakening cladding mechanical integrity, and placing fuel and cladding at risk. Although the microstructural and phase information of FCCI has been fairly understood, mechanical properties remain less studied due to limited reaction volume. Through a combining of advanced electron microscopy characterizations and small-scale mechanical testing techniques, including indentation and micro-tensile testing, this study investigated the microscale mechanical properties of FCCI between the high-Cr tempered martensitic HT9 cladding and an advanced Uranium (U)-based metallic fuel irradiated at the Advanced Test Reactor to 2.2% FIMA with peak inner cladding temperature (PICT) reached to 650 ℃. Mechanical testing results show significant hardening and embrittlement in the FCCI region. The brittle fracture of FCCI specimen is mainly attributed to the formation of nano-crystallized intermetallic σ-FeCr phase. Whereas mechanical softening was revealed in the unreacted HT9 matrix due to irradiation-induced microstructural and microchemical evolution, specifically, the disappearance of martensitic lath structure and the formation of Fe2Mo Laves phase precipitation which consumed the solid solution strengthening Mo from the HT9 matrix. Due to the achieved high cladding temperature, this fuel pin is of particular significance for revealing the high-temperature irradiation effect on the mechanical properties of HT9 cladding. Therefore, the outcomes of this study are expected to contribute to the development of multi-scale mechanical behavior modeling of HT9 cladding for Generation IV reactors which requires cladding to run at higher temperature (above 600 ℃).

He2+ irradiation induced microstructure evolution in sub-surface layer of the coarse-grained tungsten accessed by synchrotron GIXRD and GISAXS

In this study the microstructure evolution in sub-surface layer of the coarse-grained W after irradiation of 400 keV He2+ at different fluences and temperatures was assessed by synchrotron grazing incident X-ray diffraction (GIXRD) and grazing-incidence small-angle X-ray scattering (GISAXS). The splitting of GIXRD peaks obtained at different incident angles indicates that the un-irradiated coarse-grained W exhibits a compression strain state in the near surface and the He ions irradiation results in a lattice swelling which is more severe at higher irradiation temperature and fluence. The GISAXS results combined with the TEM analysis indicate that after irradiation at room temperature, uniformly distributed high density dislocation loops with an average radius of about 3.7 nm were formed due to the extrusion of He-V clusters. After irradiation at 400 °C, high density He bubbles with radii between 1 and 3 nm were formed in the sample interior, while large blisters appeared near the surface. This work provides a reference for the study of irradiation damage in the sub-surface layer of W alloys via synchrotron GIXRD and GISAXS.

Statistical distribution of spontaneous recombination radii of Frenkel pairs in FCC and BCC metals

The recombination radius of Frenkel pairs induced by incident high-energy particles represents a critical determinant of irradiation-induced microstructure evolution. A better understanding of the atomistic processes controlling recombination can help us determine more accurately the average recombination radius, an essential input parameter of mesoscale models of radiation damage evolution. This work used the kinetic activation relaxation technique—an off-lattice, self-learning kinetic Monte Carlo algorithm in conjunction with molecular dynamics simulations—to study the statistical distribution of the recombination radii in six FCC (nickel, copper, silver, gold, palladium, and platinum) and one BCC (iron) metals. We found that recombination can occur from the 2nd nearest neighbor site to as far as the 15th neighbor site. Long-range recombination is realized via crowdion formation along <110> directions. Vacancies were found to diminish the recombination energy barrier, yet no correlation was found between energy barriers and the average recombination radius. Instead, the average recombination radius increases linearly with elastic constants. Although the atomic displacement fields near self-interstitials are identical in different metals, metals with higher elastic constants exhibit larger local stresses and broader stress distribution, facilitating long-range recombination. Similarly, hydrostatic pressure enhances the average recombination radius by rising elastic constants of the compressed structure. Thus, it is the elastic constants that determine the average recombination radius, not energy barriers.

Dynamic observation of dual-beam irradiated Fe and Fe-10Cr alloys at 435 °C

The attractive mechanical properties and superior resistance to void-swelling make ferritic/martensitic alloys a promising structural material for advanced nuclear reactors. However, one anomaly that has intrigued researchers for more than 50 years is the proportion of two types of dislocation loops in Fe and Fe-Cr alloys with Burger vectors b=½<111> and b=<100>. Although the possible mechanisms responsible for the presence of 〈100〉 loops continue to be the subject of intense modeling studies, there remains incomplete experimental understanding of fundamental irradiation processes in Fe(Cr) alloys. Here, the dose dependence of the irradiation-induced microstructural evolution was examined from 0 to 20 displacement per atom (dpa) in high purity Fe and Fe-10Cr during simultaneous dual-beam (1 MeV Kr + 10 appm He/dpa) irradiation at 435 °C. We experimentally revealed that the mechanism for the formation of 〈100〉 loops may not follow the conventional simple dislocation reaction between two ½<111> loops. Real-time dynamic formation and evolution of defects including black dot loops, loop coarsening, loop decoration, network dislocations, and cavities were demonstrated. Several results indicated that the addition of Cr and He could impede dislocation loop motion. The evolution of the defect size/density and relative fraction of ½<111> vs 〈100〉 loops were quantitatively summarized. With increasing dose, ½<111> loops became the dominant type of loop in both materials. Notably, 〈100〉 loops were predominantly observed near grain boundaries only for pure Fe, while arrays of nanoscale black dot defects composing the 〈100〉 loop strings were observed in plenty in Fe-10Cr.

Formation and behaviors of helium bubbles in Li4SiO4: A molecular dynamics simulation

Molecular dynamics simulations were conducted to investigate the formation and behaviors of helium bubbles in Li4SiO4. We found that a single helium atom easily moved to a lithium vacancy with minimal energy, and helium dimers were also observed in Li4SiO4. The temperature directly influenced the formation and diffusion of helium, and helium atoms aggregated and formed helium bubbles at a temperature of 700 K. The release of helium bubbles at several temperatures was simulated to analyze the effects of helium bubble release on the Li4SiO4 surface. The temperature was also important factor influencing the recovery of defects. These results will be useful for establishing a physical model of irradiation-induced microstructure evolution inside tritium breeding material and provide a good reference for understanding the initial stage of helium bubble formation and growth in Li4SiO4.

Inferring radiation-induced microstructural evolution in single-crystal niobium through changes in thermal transport

Ion beams enable accelerated radiation exposure experiments, but traditional destructive post-irradiation examination techniques remain time-consuming. Here, measurements of thermal diffusivity are used to monitor microstructural changes in single crystal niobium irradiated with Si3+ ions up to 1.9 dpa (5.1 × 1015 ions/cm2). Changes in thermal transport are correlated to defect generation and clustering through dose-microstructure relationships. These measurements are made using transient grating spectroscopy (TGS), a multi-modal, non-destructive, rapid characterization technique. This work demonstrates that in this low-dose point defect clustering regime, direct measurements of thermal properties are a powerful tool for understanding irradiation-induced microstructure evolution.

Ion Irradiation-Induced Microstructural Evolution of Ni⁻Mo⁻Cr Low Alloy Steels.

As leading candidates of sheet steels for advanced nuclear reactors, three types of Ni–Mo–Cr high-strength low alloy (HSLA) steels named as CNST1, CNST2 and CNSS3 were irradiated by 400 keV Fe+ with peak fluence to 1.4 × 1014, 3.5 × 1014 and 7.0 × 1014 ions/cm2, respectively. The distribution and morphology of the defects induced by the sample preparation method and Fe+ irradiation dose were investigated by transmission electron microscopy (TEM) and positron-annihilation spectroscopy (PAS). TEM samples were prepared with two methods, i.e., a focused ion beam (FIB) technique and the electroplating and twin-jet electropolishing (ETE) method. Point defects and dislocation loops were observed in CNST1, CNST2 and CNSS3 samples prepared via FIB. On the other hand, samples prepared via the ETE method revealed that a smaller number of defects was observed in CNST1, CNST2 and almost no defects were observed in CNST3. It is indicated that artifact defects could be introduced by FIB preparation. The PAS S-W plots showed that the existence of two types of defects after ion implantation included small-scale defects such as vacancies, vacancy clusters, dislocation loops and large-sized defects. The S parameter of irradiated steels showed a clear saturation in PAS response with increasing Fe+ dose. At the same irradiation dose, higher values of the S-parameter were achieved in CNST1 and CNST2 samples when compared to that in CNSS3 samples. The mechanism and evolution behavior of irradiation-induced defects were analyzed and discussed.

In situ characterization of irradiation-induced microstructural evolution in urania single crystals at 773 K

Implantations with low-energy ions (Xe, La) on UO2 single crystals at 773 K were performed to investigate the role played by both the radiation damage and the incorporation of foreign elements on the matrix destabilisation. The radiation damage was monitored by both in situ RBS-C and in situ TEM during ion irradiation experiments performed at 773 K. RBS-C data shows a similar regular increase of the radiation-induced disorder in crystals for both Xe and La ions followed by a saturation plateau at about 3–4 dpa. An unexpected difference of the value of the saturation plateaus is observed, with a higher value recorded for Xe-irradiated crystals. In situ TEM images show the apparition and evolution of several defects as a function of the ion dose up to 40 dpa, irrespective of the nature of the bombarding ion: ‘black dots’ defects, dislocation loops and lines, and finally a dislocation network at high dpa. Nanometre-sized gas bubbles were observed at 773 K for the Xe-implanted crystal for doses larger than 3 dpa. Neither precipitate nor cavity were observed on La-implanted crystals. The difference in the saturation plateau as seen by RBS-C can be ascribed to the formation of the Xe aggregates that lead to an increase of the dechannelling yield.

The natural aging of austenitic stainless steels irradiated with fast neutrons

Much of today's research in nuclear materials relies heavily on archived, historical specimens, as neutron irradiation facilities become ever more scarce. These materials are subject to many processes of stress- and irradiation-induced microstructural evolution, including those during and after irradiation. The latter of these, referring to specimens “naturally aged” in ambient laboratory conditions, receives far less attention. The long and slow set of rare defect migration and interaction events during natural aging can significantly change material properties over decadal timescales. This paper presents the results of natural aging carried out over 15 years on austenitic stainless steels from a BN-350 fast breeder reactor, each with its own irradiation, stress state, and natural aging history. Natural aging is shown to significantly reduce hardness in these steels by 10–25% and partially alleviate stress-induced hardening over this timescale, showing that materials evolve back towards equilibrium even at such a low temperature. The results in this study have significant implications to any nuclear materials research program which uses historical specimens from previous irradiations, challenging the commonly held assumption that materials “on the shelf” do not evolve.

Irradiation Induced Microstructure Evolution in Nanostructured Materials: A Review.

Nanostructured (NS) materials may have different irradiation resistance from their coarse-grained (CG) counterparts. In this review, we focus on the effect of grain boundaries (GBs)/interfaces on irradiation induced microstructure evolution and the irradiation tolerance of NS materials under irradiation. The features of void denuded zones (VDZs) and the unusual behavior of void formation near GBs/interfaces in metals due to the interactions between GBs/interfaces and irradiation-produced point defects are systematically reviewed. Some experimental results and calculation results show that NS materials have enhanced irradiation resistance, due to their extremely small grain sizes and large volume fractions of GBs/interfaces, which could absorb and annihilate the mobile defects produced during irradiation. However, there is also literature reporting reduced irradiation resistance or even amorphization of NS materials at a lower irradiation dose compared with their bulk counterparts, since the GBs are also characterized by excess energy (compared to that of single crystal materials) which could provide a shift in the total free energy that will lead to the amorphization process. The competition of these two effects leads to the different irradiation tolerance of NS materials. The irradiation-induced grain growth is dominated by irradiation temperature, dose, ion flux, character of GBs/interface and nanoprecipitates, although the decrease of grain sizes under irradiation is also observed in some experiments.

Heavy ion irradiation-induced microstructural evolution in pyrochlore Lu2Ti2O7 at room temperature and 723 K

Polycrystalline pyrochlore Lu2Ti2O7 pellets were irradiated with 600keV Kr3+ at room temperature and 723K to a fluence of 4×1015ions/cm2, corresponding to an average ballistic damage dose of 10 displacements per atom in the peak damage region. Irradiation-induced microstructural evolution was examined by grazing incidence X-ray diffraction, and cross-sectional transmission electron microscopy. Incomplete amorphization was observed in the sample irradiated at room temperature due to the formation of nano-crystal which has the identical structure of pyrochlore, and the formation of nano-crystal is attributed to the mechanism of epitaxial recrystallization. However, an ordered pyrochlore phase to a swelling disordered fluorite phase transformation is occurred for the Lu2Ti2O7 sample irradiated at 723K, which is due to the disordering of metal cations and anion vacancies.

Radiation tolerance of neutron-irradiated model Fe–Cr–Al alloys

The Fe–Cr–Al alloy system has the potential to form an important class of enhanced accident-tolerant cladding materials in the nuclear power industry owing to the alloy system's higher oxidation resistance in high-temperature steam environments compared with traditional zirconium-based alloys. However, radiation tolerance of Fe–Cr–Al alloys has not been fully established. In this study, a series of Fe–Cr–Al alloys with 10–18 wt % Cr and 2.9–4.9 wt % Al were neutron irradiated at 382 °C to 1.8 dpa to investigate the irradiation-induced microstructural and mechanical property evolution as a function of alloy composition. Dislocation loops with Burgers vector of a/2〈111〉 and a〈100〉 were detected and quantified. Results indicate precipitation of Cr-rich α′ is primarily dependent on the bulk chromium composition. Mechanical testing of sub-size-irradiated tensile specimens indicates the hardening response seen after irradiation is dependent on the bulk chromium composition. A structure–property relationship was developed; it indicated that the change in yield strength after irradiation is caused by the formation of these radiation-induced defects and is dominated by the large number density of Cr-rich α′ precipitates at sufficiently high chromium contents after irradiation.

Irradiation-induced microstructural evolution and swelling of 3C-SiC

In this study, an ion-irradiated single crystal 3C-SiC under fluences of up to 20dpa at 400–1350°C was examined using synchrotron based X-ray diffraction and high resolution transmission electron microscopy. Interstitial clusters, dislocation loops, Frank loops, stacking fault loops, and voids in 3C-SiC were investigated. The high resolution TEM results show that clusters collapsed to {111} small loops when their size reached few nm with increasing temperature, and gradually develop into Frank loops with an added atomic layer along {111} at 1000°C.Interplanar spacing information of single crystal SiC was obtained from synchrotron XRD radial scan measurements. Irradiation-induced volume swelling at 400–1350°C was measured, and the anisotropic (a=b<c) swelling behavior of SiC was confirmed. In addition, humps on the right side of SiC (002) were observed, which suggested that C+/Si+-Si〈100〉 and/or C+/Si+-C〈100〉 dumbbells gave rise to diffuse scattering.