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CVD SiC samples exposed to 400 keV He ion irradiation at 750 °C were subsequently corroded in FLiNaK molten salt at 750 °C for 166 h. After corrosion, the dissolution of Si led to the formation of a carbon-rich layer and obvious denudation of the SiC. The microstructure of the C-rich layer was first visualized and characterized via TEM, confirming that a graphitic structure was present inside the layer. Furthermore, the AFM images indicate that the denudation depth increases with increasing irradiation dose. Moreover, the Ni impurities in the salt play a key role in the irradiation-promoted corrosion of SiC. These findings suggest that Ni can preferentially react with irradiation-included homonuclear Si‒Si bonds to promote SiC corrosion. Moreover, the number density of He bubbles decreases as their size increases near or within the C-rich layer, and it is speculated that the vacancies created by the loss of Si result in the migration and coalescence of He bubbles. In conclusion, irradiation and corrosion at high temperatures induced synergistic damage behavior in SiC.

Contrary to the conventional belief that, in transmission electron microscopy (TEM), selective and sharp imaging of dislocation loops can be realized only by accurate tilting of a specimen from the condition that the symmetrical axis of incident electron beam distribution is parallel to a zone axis of the TEM specimen (on-zone condition), we demonstrate that selective dark-field (DF) imaging of dislocation loops at on-zone condition is possible with a scanning TEM (STEM) mode that uses an objective lens aperture to select a diffraction disk of interest. Diffraction-selected on-zone STEM (DsoZ-STEM) has been applied to selective DF imaging of dislocation loops with a short axis length of <2 nm in a single-crystal aluminum irradiated by argon ions and an electron beam at room temperature. It was found that a Kikuchi line enhances the contrast among the dislocation loops and the matrix of DsoZ-STEM images. DsoZ-STEM obeyed g·b invisibility criterion and showed good agreement with a typical visibility change of a dislocation line and a loop in conventional DF images with a specific pair of $\pm$ ± g. In addition, dislocation loops always showed much higher brightness in the inner side compared to the outer side in DsoZ-STEM images, simplifying the distinction of dislocation loops with apparently the same long-axis direction but different b. Thus, DsoZ-STEM can simplify the selective DF imaging for the determination of the number and the character of dislocation loops. [ABSTRACT FROM AUTHOR]

Nuclear energy is essential for the future development of countries. However, both structural and functional components of nuclear power equipment are facing severe challenges of nuclear irradiation damage after experiencing irradiation growth and irradiation creep. How to avoid irradiation damage to nuclear power equipment has become a hotspot in international research and development of surface protection technology. Deposition of protective coatings on the underlying object surface or in bulk materials has been considered as a near-term solution to enhanc functional components. Different substrate materials are selected according to other service conditions within the reactor. Suitable material selection combined with relevant optimization can significantly increase the service life of materials. This review summarizes recent research on several categories of anti-irradiation coatings prepared by physical vapor deposition technology for current industrial applications. These includes metallic, ceramic, composite and high entropy alloy coatings. The review endeavors to impart a thorough understanding of the properties of these selected anti-irradiation coatings, from the fundamental aspects of their substrate materials to their practical applications across diverse settings. It explores not only the current research progress but also the potential avenues for future advancements. Additionally, the intricate relationships between coating formulations, their resistance to irradiation, and their ultimate performance in various environments are illuminated in this paper.

An accurate description of short-range interactions among atoms is crucial for simulating irradiation effects in applications related to ion modification techniques. A smooth integration of the Ziegler–Biersack–Littmark (ZBL) potential with the adaptive intermolecular reactive empirical bond-order (AIREBO) potential was achieved to accurately describe the short-range interactions for carbon-based materials. The influence of the ZBL connection on potential energy, force, and various AIREBO components, including reactive empirical bond-order (REBO), Lennard–Jones (LJ), and the torsional component, was examined with configurations of the dimer structure, tetrahedron structure, and monolayer graphene. The REBO component is primarily responsible for the repulsive force, while the LJ component is mainly active in long-range interactions. It is shown that under certain conditions, the torsional energy can lead to a strong repulsive force at short range. Molecular dynamics simulations were performed to study the collision process in configurations of the C-C dimer and bulk graphite. Cascade collisions in graphite with kinetic energies of 1 keV and 10 keV for primary knock-on atoms showed that the short-range description can greatly impact the number of generated defects and their morphology. [ABSTRACT FROM AUTHOR]

Refractory metal single crystals have been applied in key high-temperature structural components of advanced nuclear reactor power systems, due to their excellent high-temperature properties and outstanding compatibility with nuclear fuels. Although electron beam floating zone melting and plasma arc melting techniques can prepare large-size oriented refractory metals and their alloy single crystals, both have difficulty producing perfect defect-free single crystals because of the high-temperature gradient. The mechanical properties of refractory metal single crystals under different loads all exhibit strong temperature and crystal orientation dependence. Slip and twinning are the two basic deformation mechanisms of refractory metal single crystals, in which low temperatures or high strain rates are more likely to induce twinning. Recrystallization is always induced by the combined action of deformation and annealing, exhibiting a strong crystal orientation dependence. The irradiation hardening and neutron embrittlement appear after exposure to irradiation damage and degrade the material properties, attributed to vacancies, dislocation loops, precipitates, and other irradiation defects, hindering dislocation motion. This paper reviews the research progress of refractory metal single crystals from three aspects, preparation technology, deformation behavior, and irradiation damage, and highlights key directions for future research. Finally, future research directions are prospected to provide a reference for the design and development of refractory metal single crystals for nuclear applications. [ABSTRACT FROM AUTHOR]

Understanding irradiation-induced degradation processes of nuclear structural materials is essential for creating methodologies and procedures for nuclear reactor safety. Due to the time- and resource-intensive property of both experiments and multiscale simulations of irradiation damage, the trial-and-error approach is completely inefficient. Recently, machine learning techniques have been employed to predict the properties of reduced activation ferritic martensitic (RAFM) steels, such as yield strength and elongation, as well as irradiation embrittlement in steel pressure vessels, with encouraging progress. In this work, void swelling is predicted using a machine learning method for the first time, taking into account the synergistic effects of displacement damage, helium, and hydrogen. Assisted by the analysis of feature engineering, seven machine learning models are trained and compared by multicriteria evaluation methods. Finally, the parameter-optimized gradient-boosting model is selected as the mapping function with the highest accuracy and universality to predict void swelling. In particular, the dependence of the void swelling and the injection amount of helium and hydrogen in the continuous parameter variation range is predicted beyond the existing experimental data. This work demonstrates the feasibility of machine learning to predict material irradiation damage by synergistic effects and has practical significance in nuclear material optimization and reactor safety. [ABSTRACT FROM AUTHOR]

In future fusion reactors, the first wall will be irradiated with hydrogen fuel (deuterium (D), tritium (T)) and helium (He) with low energies. In this work, the effects of He plasma irradiation on D permeation through tungsten (W) coated reduced activation ferritic/martensitic (RAFM) steel were systematically investigated by performing plasma-driven permeation (PDP) measurements in the temperature range of 523–833 K. The W coatings were prepared by magnetron sputtering. He plasma irradiation with different fluences was carried out in a linear plasma device. The results showed that as He irradiation fluence increased, the surface damage of W coatings intensified, resulting in a high steady-state D-PDP flux and effective diffusion coefficient. The effect of incident ion energy on D-PDP was also influenced by He pre-irradiation. After He irradiation with a fluence of 6.0 × 10 ^24 He m ^−2 , the D-PDP flux was found to decrease with increasing incident ion energy despite the fact that the sample temperature was gradually increasing. The He pre-damage affected the overall D transport by re-balancing the bulk diffusion and surface recombination processes, therefore determining the PDP flux. D-PDP through W + RAFM under simultaneous D + He mixture plasma exposure was studied as well. The results indicated that the presence of He reduced the ionization rate of the D plasma and decreased D-PDP flux. D-PDP regime under the mixture plasma exposure was verified by calculating the transport parameter. It was found that the D-PDP regime was not changed by He introduction, which remained in the RD regime (recombination-limited D release from upstream surface and diffusion-limited D release from downstream surface).

Irradiation injury, especially caused by UVB, of the skin is one of the critical reasons for skin inflammation and damage. The present study aimed to explore the protective effect of Syzygium formosum leafy extract (SFLE) and its mechanism of action against UVB-induced damages of human keratinocytes. In this study, SFLE was prepared from 100 kg dried leaves using industrial-scale processes. We found that SFLE markedly reduced markers of the skin inflammation in UVB-induced pro-inflammatory cytokines. Only 2 μg/mL of SFLE exhibited significantly stronger anti-inflammatory effects than the fivefold concentration of positive control. Intriguingly, an anti-inflammatory enzyme, heme oxygenase-1 expression was significantly induced by SFLE treatment. MMP-3 and -9 were, but not MMP-1, significantly reduced. SFLE inhibited the expression of the MAPK pathway, resulting in a decrease on UVB-induced reactive oxygen species. In conclusion, SFLE can potentially be used to treat skin inflammatory diseases. [ABSTRACT FROM AUTHOR]

Along with the globalization of environmental problems and the rapid development of the field of nuclear technologies, the severe irradiation damage of materials has become a big issue, restricting the development of advanced nuclear reactor systems. Refractory high-entropy alloys (RHEAs) have the characteristics of a complex composition, a short-range order, and lattice distortion and possess a high phase stability, outstanding mechanical properties, and excellent irradiation resistance at elevated temperatures; thus, they are expected to be promising candidates for advanced nuclear reactors. This review summarizes the design, preparation, and irradiation resistance of irradiation-tolerant RHEAs. It encompasses a comprehensive analysis of various aspects, including the evolution of defects, changes in microstructure, and the degradation in properties. Furthermore, the challenges and insufficiently researched areas regarding these alloys are identified and discussed. Building on this foundation, the review also provides a forward-looking perspective, outlining potential avenues for future research. [ABSTRACT FROM AUTHOR]

Photoluminescence (PL) properties of Er2O3 specimens were examined by using visible lasers (532 nm and 635 nm) and a UV LED light source (365 nm) to investigate the applicability for irradiation damage monitoring of materials in fusion reactors. Both in the laser induced and UV light induced PL spectra, green (510–590 nm) and red (630–725 nm) luminescence was observed. In the spectrum measurements on specimens with different crystallinities, it was confirmed that an intensity of the red luminescence weakened significantly compared with that of the green luminescence in an Er2O3 specimen with a lower crystallinity. The results indicate that the PL measurements of Er2O3 materials could be applicable for the irradiation damage monitoring in fusion reactors. The luminescence property of ion beam irradiated Er2O3 showed that information of irradiation damages could be kept up to ∼ 300 ℃ and almost recovered at 700 ℃. Based on the obtained luminescence properties, positions in a fusion reactor where Er2O3 materials could be used as irradiation damage sensors are proposed. Changes in PL spectra at high temperatures up to ∼ 400 ℃ indicate the possibility that the Er2O3 materials might be applicable also for temperature monitoring of in-vessel components during reactor maintenance periods.

Subject to intense high-energy particle irradiation, various effects manifest within a material. Specifically, when high-energy particles collide with lattice atoms in a crystal material, a sequence of interactions is set in motion, initiating irradiation effects. Using GaAs solar cells as an example, this study investigates how potential barriers in different directions of atoms in sphalerite structures affect atomic displacement and establishes a probability model for lattice atomic displacement under proton irradiation. By combining Markov chains, changes in displacement threshold energy with different crystal orientations are described, and a damage model for cascading collision relationships that cause irradiation effects is established. Finally, the new model is compared to classical models, and differences in defects caused by proton impacts at several energies on GaAs crystals are simulated. [ABSTRACT FROM AUTHOR]

Excellent irradiation resistance is the basic property of nuclear materials to keep nuclear safety. The high-entropy design has great potential to improve the irradiation resistance of the nuclear materials, which has been proven in alloys. However, whether or not high entropy can also improve the irradiation resistance of ceramics, especially the mechanism therein still needs to be uncovered. In this work, the irradiation and helium (He) behaviors of zirconium carbide (ZrC)-based high-entropy ceramics (HECs), i.e., (Zr0.2Ti0.2Nb0.2Ta0.2W0.2)C, were investigated and compared with those of ZrC under 540 keV He ion irradiation with a dose of 1×1017 cm−2 at room temperature and subsequent annealing. Both ZrC and (Zr0.2Ti0.2Nb0.2Ta0.2W0.2)C maintain lattice integrity after irradiation, while the irradiation-induced lattice expansion is smaller in (Zr0.2Ti0.2Nb0.2Ta0.2W0.2)C (0.78%) with highly thermodynamic stability than that in ZrC (0.91%). After annealing at 800 ℃, ZrC exhibits the residual 0.20% lattice expansion, while (Zr0.2Ti0.2Nb0.2Ta0.2W0.2)C shows only 0.10%. Full recovery of the lattice parameter (a) is achieved for both ceramics after annealing at 1500 ℃. In addition, the high entropy in the meantime brings about the favorable structural evolution phenomena including smaller He bubbles that are evenly distributed without abnormal coarsening or aggregation, segregation, and shorter and sparser dislocation. The excellent irradiation resistance is related to the high-entropy-induced phase stability, sluggish diffusion of defects, and stress dispersion along with the production of vacancies by valence compensation. The present study indicates a high potential of high-entropy carbides in irradiation resistance applications.

Understanding the irradiation damage mechanisms of GaN is of great importance for improving irradiation resistance and the ion implantation processes of GaN‐based devices. The understanding of the damage mechanisms, which are normally simulated using molecular dynamics (MD), is bottlenecked by the dilemma that ab initio MD is limited by relatively high computational cost, whereas classical MD suffers from low accuracy. In this paper, a global neural network (G‐NN) potential is constructed for GaN using random stochastic surface walking global optimization combined with global neural network potential (SSW‐NN). By benchmarking the properties of intrinsic defects and defect‐pairs, as well as the threshold displacement energies along different crystallographic directions, this potential is found to provide accuracy similar to ab initio calculations. Furthermore, based on the large‐scale simulations of the knock‐on process, Ga and N vacancies, as well as N interstitials are found to be the major generated defects, whereas only Ga vacancies are predicted in the previous device simulation studies that have been widely recognized. This potential provides an efficient and accurate tool to gain a fundamental understanding of the irradiation damage mechanisms of GaN and refine the parameters in the related device simulations. [ABSTRACT FROM AUTHOR]

In nuclear fusion (ITER and the future DEMO), those components that face the plasma are exposed to high temperature and irradiation which, in the long term, modifies their thermal and mechanical properties and tritium retention. Tungsten is a candidate material and is the subject of many studies of microstructure evolution under various irradiation and temperature conditions. One milestone is the characterization of its defect properties. We here readdress the diffusion of nanocavities on broad ranges of size and temperature and compare it with dissociation, a competing process during nanocavity growth. First, at the atomic scale, we used molecular dynamics to explore the variety of elementary events involved in the nanocavity diffusion. Second, an experimental study of ion-irradiated samples, annealed at different temperatures up to 1,773 K, revealed the creation and growth of nanocavities on transmission electron microscopy images. Third, we performed multi-objective optimization of the nanocavity diffusion input of our object kinetic Monte Carlo model to reproduce the experimental results. Finally, we applied a sensitivity analysis of the main inputs of our model developed for these particular conditions—the source term which combines two cascade databases and the impurities whose interaction with the defects is characterised with a supplemented database of density functional theory calculations. Three domains of nanocavity size were observed. The first is the small vacancy clusters, for which atomistic calculations are possible and dissociation is negligible. The second is the small nanocavities, for which we provide new diffusion data and where a competition with the dissociation can take place. The third domain is the large nanocavities, for which, in any case, the dissociation prevents their existence above 1,500 K in the absence of a stabilizing interface.

Ceramic composites are promising candidates as structural materials for future fission and fusion reactors. In present work, Al 2 O 3 -ZrO 2 -SiC ternary ceramic composites were irradiated with 2.0 MeV He-ions at 300 and 800 ℃. Grazing incidence X-ray diffraction results confirmed that there was irradiation induced shift and broadening of diffraction peaks, but no amorphization of Al 2 O 3 -ZrO 2 -SiC composite were observed up to fluence of 1.72 × 1018 ions/cm2. Transmission electron microscopy observations showed that throughout the entire irradiation region, nano-sized helium bubbles are mostly distributed in Al 2 O 3 grains and partly in ZrO 2 grains, while no detectable bubbles are observed in SiC grains. No obvious agglomeration of bubbles was found at grain/phase boundaries. By using nanoindentation technique, slight hardening or softening was confirmed for the samples irradiated at 300 and 800 ℃ respectively. The absence of amorphization and surface exfoliation indicating the Al 2 O 3 -ZrO 2 -SiC composite exhibits remarkable resistance to He-ions irradiation. [ABSTRACT FROM AUTHOR]

The increasingly harsh environment of the nuclear reactors and the insurmountable flaws of in-service materials have created an urgent need for the development of the brand-new alloys. For last decade, the high-entropy alloys (HEAs), a novel composition-design strategy, have received much attention due to their promise for the nuclear fields. The application of the multiscale modelling is to explore the irradiation performance and underlying mechanisms of HEAs. Abundant results and data deepen the understanding of the irradiation response, and accelerate the development of advanced irradiation-resistant HEAs. This review introduces the state-of-art multiscale modelling used for studying the irradiated properties of HEAs. Representative irradiation-induced microstructures and properties, as well as damage, are summarized. By strengthening the application of multiscale modelling, a rational design of high irradiation-resistant HEAs is expected. [ABSTRACT FROM AUTHOR]

In this paper, the total ionize dose (TID) irradiations for NPN bipolar transistors were carried out by 60Co γ rays. Obvious degradations were observed after irradiation, which are manifested as the increase of base current and the decrease of current gain. A more obvious increase of base current was observed at lower emitter junction bias. It shows that the degradation of the base current mainly origins from the increase of the recombination current of the emitter junction, since the recombination current component in base current is dominant at lower emitter junction bias. It can be concluded that the performance degradations of bipolar transistors are mainly due to the increase of recombination current caused by radiationinduced traps in the oxide. It is found that the bias condition during irradiation is the key factor affecting the TID effect of NPN transistor. The reverse bias is a worse bias condition than the zero bias during irradiation. Furthermore, the TID effect of two kinds of NPN bipolar transistors was compared. The two kinds of transistors have the same structure, except that one of the devices is radiation hardened by improving the surface state of the base region. The radiation induced degradations in radiation hardened devices are less than that in unhardened devices. For the unhardened device, the base current increases from 882×10-8 A to 367×10-7 A at |VEB|=05 V after 50 krad(Si) irradiation, increased by 316%. While for the hardened device, the base current increases from 751×10-8 A to 119×10-7 A at |VEB|=05 V after 50 krad(Si) irradiation, increased by 584%. Finally, the deep level traps in NPN transistors were measured by deep level transient spectrum (DLTS) before and after irradiation. The radiation will increase the trap densities and alter the energy level of the traps for the unhardened devices. By comparing the DLTS results of hardened and unhardened devices, it is found that there are great differences in their native traps. The energy level of the traps in the hardened devices is farther away from the center of the band gap than the unhardened devices. It may implies that the radiation hardness of the device can be achieved by improving the native trap states of the device.

In this paper, weld metal from unique material of a decommissioned boiling water reactor pressure vessel is investigated. The reactor was in operation for 23 effective full power years. The elemental distribution of Ni, Mn, Si and Cu in the material is analysed using atom probe tomography. There are no well-defined clusters of these elements in the weld metal. However, some clustering tendencies of Ni was found, and these are interpreted as a high number density of small features. Cu atoms were found to statistically be closer to Ni atoms than in a fully random solid solution. The impact of the non-random elemental distribution on mechanical properties is judged to be limited.

Because any fuel assembly material will be exposed to extensive irradiation during reactor operation, understanding the effects of irradiation on the microstructure of coatings and structural materials is critical to improving nuclear reactor safety and performance. In the present work, the point defect and behavior of H in the MAX phase Cr2AlC coating are investigated using first-principles calculations. It is found that the electronic structure of MAX phase Cr2AlC material determines its special properties. The formation of Frenkel pairs leads to the expansion of interlayer spacing along the c-axis, which is the initial cause of the phase transition of the MAX phase structure caused by irradiation damage. Meanwhile, the relationship between the energy of intrinsic defect formation and the chemical potential was analyzed, which showed that sufficient C content was conducive to the structural stability of CrAl bonds. Additionally, the obtained a lower energy barrier of the H atom moving parallel to the Al plane, indicating that the H-bubble tends to aggregate near the Al atomic panel. This study provides a fundamental understanding of the point defect properties and the behavior of impurities in the MAX phase and provides theoretical support for the study of anti-radiation performance in nuclear applications.

This work aims at exploring the irradiation resistance of oxide-dispersion-strengthened reduced-activation ferritic/martensitic (ODS-RAFM) steel prepared by a rapid and cost-efficient method, wire and arc additive manufacturing (WAAM). WAAM ODS-RAFM steel and RAFM steel were prepared using flux cored wires with and without addition of Y2O3, respectively. Both specimens were irradiated with 2.5 MeV Fe ions at 450 °C to a fluence of 1.84 × 1016 ions cm−2, corresponding to a peak damage dose of 20 dpa. The microstructure, hardness and defect structure of WAAM ODS-RAFM steel before and after irradiation were investigated and compared with RAFM steel. It is found that nano-Y2Ti2O7 particles with a high number density (∼10 nm, 1022/m3) are formed in WAAM ODS-RAFM steel, which pin grain boundary and dislocations motion, thus rendering a stable matrix structure after irradiation. The particle size of the stable nano-Y2Ti2O7 slightly increases to ∼ 15 nm while its number density is still high (1022/m3) after irradiation. The grain boundaries, dislocations and nano-Y2Ti2O7 interfaces act as defect sinks to absorb irradiation-induced defects such as self-interstitial atoms and vacancies. Therefore, small dislocation loops (3.6 nm) are found to be the main irradiation defects while no voids are observed. As a result of the inhibited formation and growth of defects, the irradiation hardening is insignificant (ΔH0 = 0.16 GPa), showing great promises of WAAM ODS-RAFM steel as structural components in future fusion reactors.

Abstract Understanding the irradiation damage mechanisms of GaN is of great importance for improving irradiation resistance and the ion implantation processes of GaN‐based devices. The understanding of the damage mechanisms, which are normally simulated using molecular dynamics (MD), is bottlenecked by the dilemma that ab initio MD is limited by relatively high computational cost, whereas classical MD suffers from low accuracy. In this paper, a global neural network (G‐NN) potential is constructed for GaN using random stochastic surface walking global optimization combined with global neural network potential (SSW‐NN). By benchmarking the properties of intrinsic defects and defect‐pairs, as well as the threshold displacement energies along different crystallographic directions, this potential is found to provide accuracy similar to ab initio calculations. Furthermore, based on the large‐scale simulations of the knock‐on process, Ga and N vacancies, as well as N interstitials are found to be the major generated defects, whereas only Ga vacancies are predicted in the previous device simulation studies that have been widely recognized. This potential provides an efficient and accurate tool to gain a fundamental understanding of the irradiation damage mechanisms of GaN and refine the parameters in the related device simulations.

In this paper, experimental work is supported by multi-scale numerical modeling to investigate nanomechanical response of pristine and ion irradiated with Fe2＋ ions with energy 5 MeV high purity iron specimens by nanoindentation and Electron Backscatter Diffraction. The appearance of a sudden displacement burst that is observed during the loading process in the load–displacement curves is connected with increased shear stress in a small subsurface volume due to dislocation slip activation and mobilization of pre-existing dislocations by irradiation. The molecular dynamics (MD) and 3D-discrete dislocation dynamics (3D-DDD) simulations are applied to model geometrically necessary dislocations (GNDs) nucleation mechanisms at early stages of nanoindentation test; providing an insight to the mechanical response of the material and its plastic instability and are in a qualitative agreement with GNDs density mapping images. Finally, we noted that dislocations and defects nucleated are responsible the material hardness increase, as observed in recorded load–displacement curves and pop-ins analysis. [ABSTRACT FROM AUTHOR]

• The irradiated sample with higher W content shows swelling and melting for all discharges, the CrNbTaVW only shows blisters. • The effect of irradiation was more severe in CrNbTaVW 1.7 when compared to CrNbTaVW. • The deuterium retention was evidenced to be in a deeper depth between 1.3 and 3 μm. • Deuterium retention was higher for CrNbTaVW 1.7 when compared with CrNbTaVW for 3 discharges applied. CrNbTaVW x high-entropy alloys have been developed for plasma facing components to be applied in nuclear fusion reactors. The CrNbTaVW x (x = 1 and 1.7) compositions were prepared by ball milling and consolidated at 1600 °C under 90 MPa. To study the irradiation resistance of these materials, deuterium plasmas were used to irradiate the samples in the PF-1000U facility with 1 and 3 discharges. Structural changes before and after irradiation were analyzed by scanning electron microscopy coupled with energy dispersive X-ray spectroscopy. Nuclear reaction analysis was carried out with 1000 and 2300 keV 3He+ ion beams to evaluate the profile and amount of retained deuterium on the irradiated samples. After irradiation, the sample with higher W content revealed swelling and melting for all discharges, while in the case of CrNbTaVW only blisters were observed. The deuterium retention was higher for CrNbTaVW 1.7 when compared with CrNbTaVW for 3 discharges applied. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]

Combining the advantages of a wet chemical method and spark plasma sintering, carbide-doped materials W-1wt%TiC and W-1wt%ZrC were prepared. Microstructural evolution in W-1wt%TiC and W-1wt%ZrC under irradiation of 5 keV He+ at 600 °C to fluences up to 5.0 × 1021 ions/m2 with ion flux of about 8.8 × 1017 ions/m2s was investigated by transmission electron microscopy (TEM). The dislocation loop number density of W-1wt%TiC was higher than that of W-1wt%ZrC, but the average loop size of the W-1wt%TiC was in average smaller. There were no observable helium bubbles in W-1wt%TiC and W-1wt%ZrC, exhibiting higher radiation resistance to He+ compared to pure W. He+ pre-damaged and undamaged W-1wt%TiC and W-1wt%ZrC samples were irradiated by 5 keV D2+ to estimate the D retention in doped W materials. The irradiation damage impact of He+ on deuterium retention was examined by a method of thermal desorption spectroscopy (TDS). Compared with the undamaged samples, it was illustrated that D2 retention of W-1wt%TiC and W-1wt%ZrC increased after He+ pre-irradiation. [ABSTRACT FROM AUTHOR]

Although alloying can significantly increase the hydrogen permeability of palladium (Pd) and solve the problem of hydrogen embrittlement in service, the intrinsic mechanism has not been revealed. Here, we report intuitive differences in the nucleation, growth and distribution of dislocation loops in Pd and PdAg alloys by in-situ hydrogen irradiation in TEM. The results reveal that alloying element Ag not only modifies the behavior of H, but also induces defects to change from the original long-range one-dimensional (1D) diffusion to short-range three-dimensional (3D) diffusion, which significantly delays the aggregation and growth of dislocation loops and reduces the total irradiation damage and H retention. A new strategy to improve hydrogen embrittlement resistance is proposed, i.e., constructing dense dislocation loop bands (or other sinks that can efficiently trap H) in front of grain boundaries to prevent H from diffusing into the grain boundaries. [Display omitted] [ABSTRACT FROM AUTHOR]

The yttria-stabilized zirconia (YSZ) coatings with 4–6 mol.% Y 2 O 3 content were irradiated by 6 MeV Au ions with a fluence 6.0 × 1015 ions/cm2. The cubic to monoclinic transformation of ZrO 2 was identified in the Au-irradiated samples. The cavities were observed in the YSZ layer, while larger size cavities were found inside the 6YSZ layer. LBE corrosion results showed that irradiation improved the adhesion and corrosion resistance of the coating due to the fact that irradiation released the stresses in the coating. Overall, the 4YSZ coating showed the best corrosion resistance and irradiation resistance in the case of heavy ion irradiation. • The YSZ coatings with 4–6 mol.% Y 2 O 3 content were irradiated by 6 MeV Au ions with a fluence 6.0 × 1015 ions/cm2. • For YSZ coatings, the irradiation damage of zirconia increases with increasing yttria content. • Larger size cavities were found in the 6YSZ layer. [ABSTRACT FROM AUTHOR]

The paper investigated the irradiation damage of CZT detectors caused by high flux an X-ray from X-ray tube operated at a tube voltage of 140 kV and an accumulated dose of 130 kGy. Deep-level transient spectroscopy (i-DLTS) was employed to characterize the defects. Four types of irradiation-induced defects were identified, corresponding to the energy levels at Ec-0.1eV, Ev+0.17eV, Ev+0.27eV and Ec-0.55eV, which are associated with the point defects, In Cd +/0, V Cd -/0, V Cd 2−/− and Te Cd 2+/+ respectively. After the irradiation, the concentration of V Cd -/0 changed from 5.63 × 1012 cm−3 to 3.30 × 1013 cm−3 and V Cd 2−/− increased from 1.13 × 1012 cm−3 to 1.88 × 1013 cm−3. The resistivity of all samples showed an upward trend after the irradiation, with the most significant change observed in sample 1, where it doubled from 1.09 × 1011 Ω cm to 1.94 × 1011 Ω cm. However, the energy resolution for the spectrum peak of 241Am (59.5 keV) decreased from 6.2% to 10.7% and the signal noise to ratio decreased from 42.24 to 27.33 when irradiated from anode. The counting performance of the photon counting module decreased by approximately 1.81% after irradiation on the cathode side. [ABSTRACT FROM AUTHOR]

In the design of the spallation target of CiADS, T91 steel was selected for the most critical component, the beam window. A Si-modified martensitic steel (SIMP) has been currently developed, which has better corrosion resistance and is planning to replace T91 as a material of beam window in the future. In order to assess the effect of irradiation damage on the microstructure evolution and corrosion behavior of the SIMP, the pre-irradiated samples with irradiation doses from 5 × 1015 to 3 × 1017 He/cm2 exposed in static lead-bismuth eutectic (LBE) with saturated oxygen at 350 °C for 4000 h were investigated. Results show that pre-irradiation neither change the double-layer structure of the oxide layer nor significantly affect the corrosion rate. This may be due to the slow diffusion of elements at low temperatures, and the low irradiation dose not reaching the threshold for triggering accelerated corrosion. [ABSTRACT FROM AUTHOR]

Abstract Objective This study aimed to reveal the protective effect of hydrogen storage nanomaterial MgH2 on radiation-induced male fertility impairment. Methods The characterization of MgH2 were analyzed by scanning electron microscopy (SEM) and particle size analyzer. The safety of MgH2 were evaluated in vivo and in vitro. The radioprotective effect of MgH2 on the reproductive system were analyzed in mice, including sperm quality, genetic effect, spermatogenesis, and hormone secretion. ESR, flow cytometry and western blotting assay were used to reveal the underlying mechanisms. Results MgH2 had an irregular spherical morphology and a particle size of approximately 463.2 nm, and the content of Mg reached 71.46%. MgH2 was safe and nontoxic in mice and cells. After irradiation, MgH2 treatment significantly protected testicular structure, increased sperm density, improved sperm motility, reduced deformity rates, and reduced the genetic toxicity. Particularly, the sperm motility were consistent with those in MH mice and human semen samples. Furthermore, MgH2 treatment could maintain hormone secretion and testicular spermatogenesis, especially the generation of Sertoli cells, spermatogonia and round sperm cells. In vitro, MgH2 eliminated the [·OH], suppressed the irradiation-induced increase in ROS production, and effectively alleviated the increase in MDA contents. Moreover, MgH2 significantly ameliorated apoptosis in testes and cells and reversed the G2/M phase cell cycle arrest induced by irradiation. In addition, MgH2 inhibited the activation of radiation-induced inflammation and pyroptosis. Conclusion MgH2 improved irradiation-induced male fertility impairment by eliminating hydroxyl free radicals. Graphical abstract Mice fertility and function were evaluated with or without MgH2 treatment after 5 Gy irradiation. MgH2 had the ability of hydroxyl radicals scavenging and MDA suppressing in testicular tissue induced by irradiation. Further, MgH2 could participate in spermatogenesis and protect sperm development in three stages: the generation of Sertoli cells (Sox-9+), spermatogonia (Stra8+) and round sperm cells (Crem+). Moreover, MgH2 alleviated the decrease of testosterone secreted by interstitial cells after irradiation. In addition, MgH2 suppressed apoptosis, pyroptosis and inflammatory response and alleviated cell cycle arrest by mediating IR-induced ROS.

Focused Ion Beam patterning has become a widely applied technique in the last few decades in the micro- and nanofabrication of quantum materials, representing an important advantage in terms of resolution and versatility. However, ion irradiation can trigger undesired effects on the target material, most of them related to the damage created by the impinging ions that can severely affect the crystallinity of the sample, compromising the application of Focused Ion Beam to the fabrication of micro- and nanosized systems. We focus here on the case of Bi2Se3, a topological material whose unique properties rely on its crystallinity. In order to study the effects of ion irradiation on the structure of Bi2Se3, we irradiated with Ga+ ions the full width of Hall-bar devices made from thin films of this material, with the purpose of inducing changes in the electrical resistance and characterizing the damage created during the process. The results indicate that a relatively high ion dose is necessary to introduce significant changes in the conduction. This ion dose creates medium-range lateral damage in the structure, manifested through the formation of an amorphous region that can extend laterally up to few hundreds of nanometers beyond the irradiated area. This amorphous material is no longer expected to behave as intrinsic Bi2Se3, indicating a spatial limitation for the devices fabricated through this technique. [ABSTRACT FROM AUTHOR]

This work researched the effects of irradiation on the current-voltage characteristics and voltage magnetic sensitivity of the silicon magnetic sensitive transistor (SMST). The 1-MeV electron irradiation source was used to irradiate the SMST. The irradiation fluences were 1 × 1012 e/cm2, 1 × 1013 e/cm2 and 1 × 1014 e/cm2, respectively (the irradiation flux was 1 × 1010 cm−2·s−1). The experimental results demonstrate that the collector current (IC) of the SMST occurs attenuation after irradiation under the same collector voltage (VCE) and the base current (IB). The attenuated rate of the IC increases obviously with the enhance of electron irradiation fluence when the IB is the same. Moreover, the attenuated rate of the IC increases slight with the rise of the IB when the electron irradiation fluence is the same. When the supply voltage is 5.0 V (RL = 1.5 kΩ) and the IB is 4.0 mA, the voltage magnetic sensitivity (SV) of the SMST occurs attenuate after irradiation. The attenuated rate of the SV increases with the enhance of electron irradiation fluence. [ABSTRACT FROM AUTHOR]

Along with the globalization of environmental problems and the rapid development of the field of nuclear technologies, the severe irradiation damage of materials has become a big issue, restricting the development of advanced nuclear reactor systems. Refractory high-entropy alloys (RHEAs) have the characteristics of a complex composition, a short-range order, and lattice distortion and possess a high phase stability, outstanding mechanical properties, and excellent irradiation resistance at elevated temperatures; thus, they are expected to be promising candidates for advanced nuclear reactors. This review summarizes the design, preparation, and irradiation resistance of irradiation-tolerant RHEAs. It encompasses a comprehensive analysis of various aspects, including the evolution of defects, changes in microstructure, and the degradation in properties. Furthermore, the challenges and insufficiently researched areas regarding these alloys are identified and discussed. Building on this foundation, the review also provides a forward-looking perspective, outlining potential avenues for future research.

The clustering behaviors and stable configurations of vacancies with the size up to 30 in vanadium were studied by first-principles investigations. The average formation energies per vacancy of vacancy clusters decreases with the sizes. When Vn cluster with the odd size forms Vn+1 cluster, the average formation energy per vacancy has a lager reduction relative to the case of even size. The stabilities of vacancy clusters increase with the sizes in terms of the total binding energy (n = 2–30), the incremental binding energies of vacancy clusters show an increasing trend of fluctuation with the sizes, which can be explained by both the size characteristics and terrace-ledge-kink model. We also estimated the relationship between vacancy formation/incremental binding energy and Wigner-Seitz area. Finally, we established a machine learning model based on Linear-SVM to predict formation energies of larger vacancy clusters. These results deepened the understanding of the formation and evolution of nanovoids in vanadium metal under irradiation.

Subject to intense high-energy particle irradiation, various effects manifest within a material. Specifically, when high-energy particles collide with lattice atoms in a crystal material, a sequence of interactions is set in motion, initiating irradiation effects. Using GaAs solar cells as an example, this study investigates how potential barriers in different directions of atoms in sphalerite structures affect atomic displacement and establishes a probability model for lattice atomic displacement under proton irradiation. By combining Markov chains, changes in displacement threshold energy with different crystal orientations are described, and a damage model for cascading collision relationships that cause irradiation effects is established. Finally, the new model is compared to classical models, and differences in defects caused by proton impacts at several energies on GaAs crystals are simulated.

In this paper, a combined transport-defect evolution multiscale model describing the generation and the evolution of microstructure damage in silicon carbide (SiC) induced by focused helium ion beams is developed. In the proposed model, the transport of helium ions and displaced atoms in the SiC substrate and the generation of point defects are described by the Boltzmann transport equations, while the subsequent defect evolution is characterized by a set of rate equations with the contributions of the modeling of the bubble coalescence as well as the substrate swelling. The validity and superiority of the transport equations are verified by comparing the simulation results with the data from experimental measurements and available simulation methods. The subsurface amorphous profile, onsurface swelling profile, and the spatial and size distribution of helium bubbles in a SiC substrate irradiated by focused helium ion beams are simulated using the proposed multiscale model. The damage morphology simulated by the proposed model is in good agreement with the transmission electron microscopy images at different beam energies and doses. This work provides an effective tool for full-stage modeling of complex evolutionary mechanisms of microstructure damage induced by precise and high-throughput helium irradiation. • A multiscale model linking the complex evolution stages of defects in silicon carbide induced by precisely localized irradiation of He + FIB was developed. • The model is based on the combination of Boltzmann equations and defect rate equations supplemented by bubble coalescence and substrate swelling mechanisms. • The combined model has favorable prediction reliability for the microstructure damage morphology under different beam parameters. [ABSTRACT FROM AUTHOR]

In current work, the effect of nanoparticle size in Cu-Y 2 O 3 alloy on the microstructure response after He/D sequential irradiation is studied using a transmission electron microscope (TEM) and nanoindentation. The results indicated that the direct contribution of nanoparticle/matrix interface to defect evolution and irradiation hardening of ODS Cu is limited when the nanoparticle size is larger than 20 nm. The difference in the proportion of ∑3 grain boundary and small angle grain boundaries was observed with different nanoparticle sizes, which suggests that the indirect effects on the sink strength caused by changes in the nature of grain boundary cannot be ignored, especially for the large particles. Besides, the distribution of bubbles after sequential He/D irradiation was highly similar to that in single He irradiation. • Nanoparticles with a size of >20 nm have limited effect on the defect evolution. • Change on the nature of grain boundary has significant impact defect evolution. • The He pre-implantation has significant impact on the distribution of bubbles. [ABSTRACT FROM AUTHOR]

Understanding irradiation damage involves a multi-scale and multi-physics approach, integrating data from experiments, simulations and phenomenological models. This paper focuses on its early stages, specifically neutron-induced collision cascades in zirconium, as nuclear-grade zirconium alloys are widely used in fuel assemblies. We have gathered and analysed a significant sample of results from high-fidelity, large-scale molecular dynamics (MD) simulations, employing existing interatomic potentials and the two-temperature model to account for electron–phonon coupling. Our data can be directly applied to higher-scale methods. Furthermore, we carried out a comprehensive statistical analysis of the features associated with the defect production, including the number of defects, their distribution and size of the affected area. As a result, we developed a generative model of collision cascades that is parametric, hierarchical and stochastic, i.e. it takes into account a statistical nature of the phenomenon, is interpretable and shareable. This model has been developed with three primary objectives: to provide a sufficient descriptor of a cascade, to interpolate data obtained from high-fidelity simulations, and to demonstrate that the statistical model can produce representative distributions of primary irradiation defects. The results can be used to generate synthetic inputs for models at longer length and time scales, to provide fast approximations that take into account the morphology of introduced defects, and in general to serve as a powerful analytical tool. [ABSTRACT FROM AUTHOR]

MAB phases (M = transition metal, A = aluminium (Al), B = boron (B)), a new family of nano-laminated ternary transition metal borides with an excellent combination of functional and structural properties, are currently being investigated as potential radiation shielding materials in nuclear fusion reactors. Here, we report the fabrication of a solid solution MAB phase Mo 0.75 W 0.25 AlB (Mo - molybdenum, W - tungsten), in the form of dense and high purity pellets, synthesised for the first time using a reactive hot-pressing method. The shielding and irradiation damage properties of the material were also evaluated. Monte Carlo N-Particle (MCNP) calculations showed that W doping in the MoAlB phase is likely to enhance the shielding capacity against neutrons and gamma rays of different energies. The experimental damage tolerance of Mo 0.75 W 0.25 AlB to silicon ion (Si+) irradiation indicated that this solid solution compound has a high resistance to crack formation and other radiation damage effects compared to that of pure MoAlB. Thermal annealing of the amorphous Mo 0.75 W 0.25 AlB layer formed by irradiation showed that the damaged structure easily recrystallizes at high temperatures, completely returning to the virgin state. The results show that this solid solution should be considered as a promising candidate material for radiation shielding components. • A high density and purity Mo 0.75 W 0.25 AlB solid solution was synthesised by sintering at 1100 °C. • Mo 0.75 W 0.25 AlB exhibits excellent shielding capacity against gamma rays and neutrons of different energies. • Unlike MAX phases, no cracks or voids were observed on the surface of the amorphous MAB phase. • The radiation-induced amorphization of the solid solution was almost completely recovered by thermal annealing at 700 °C. [ABSTRACT FROM AUTHOR]

Single-phase Al 4 SiC 4 powder with a low neutron absorption cross section was synthesized and mixed with SiC powder to fabricate highly densified SiC ceramics by hot pressing. The densification of SiC ceramics was greatly improved by the decomposition of Al 4 SiC 4 and the formation of aluminosilicate liquid phase during the sintering process. The resulting SiC ceramics were composed of fine equiaxed grains with an average grain size of 2.0 μm and exhibited excellent mechanical properties in terms of a high flexure strength of 593 ± 55 MPa and a fracture toughness of 6.9 ± 0.2 MPa m1/2. Furthermore, the ion-irradiation damage in SiC ceramics was investigated by irradiating with 1.2 MeV Si5+ ions at 650 °C using a fluence of 1.1 × 1016 ions/cm2, which corresponds to 6.3 displacements per atom (dpa). The evolution of the microstructure was investigated by X-ray photoelectron spectroscopy (XPS) and Raman spectroscopy. The breaking of Si–C bonds and the segregation of C elements on the irradiated surface was revealed by XPS, whereas the formation of Si–Si and C–C homonuclear bonds within the Si–C network of SiC grains was detected by Raman spectroscopy. [ABSTRACT FROM AUTHOR]