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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]

With the rapid development of nuclear industry, advanced niobium alloys with high strength and irradiation resistance are urgently required. In this work, a new multi-component Nb 80 Mo 10 Zr 10 Cr 10 Al 5 B 0.1 alloy has been designed, and the yield strength in compression can reach ∼1100 MPa, and plastic strain is ∼11.3%. To reveal the irradiation damage mechanism, both the NbMoZrCrAlB alloy and pure Nb samples have been irradiated by Xe ion at 350 °C and 750 °C, with the fluence of 2.1 × 1015 ions/cm2. The average diameter of Xe bubble under 750 °C irradiation is only ∼2.92 nm in NbMoZrCrAlB alloy, which is smaller than that of pure Nb (∼ 9.22 nm). Consequently, the volume swelling ratio is only 0.04% in NbMoZrCrAlB alloy, which is largely lower than that of ∼0.65% in pure Nb. Meanwhile, both the average diameter and number density of dislocation loops in NbMoZrCrAlB alloy are obviously decreased compared with pure Nb at the same irradiation temperature. This result should arise from the severe lattice distortion and sluggish diffusion in present NbMoZrCrAlB alloy, which is similar with the high entropy alloy. The present work provides a reference for designing novel Nb based alloys for advanced nuclear materials. [Display omitted] • The yield stress of the novel Nb alloy can reach ∼1100 MPa. • The novel Nb alloy shows a low swelling(0.04%) at elevated temperature irradiation. • The novel Nb alloy shows a distinct decrease in number density of dislocation loops compared with pure Nb. [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.

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.

Understanding irradiation damage and effects in α-uranium (α-U) is critical to modeling the behavior of U-based metallic fuels. The aim of this review is to address the renewed interest in U-based metallic fuels by examining the state-of-the-art knowledge associated with the effect of irradiation on the microstructure, dimensional changes, and properties of α-U. We critically review the research progress on irradiation-induced growth and swelling, the enhancement of plastic flow and superplasticity by irradiation, and the effect of irradiation on thermal and electrical properties of α-U. Finally, we outline the research directions that require advancements, specifically the need to carry out fundamental research on several of the less understood mechanisms of irradiation damage and effects in α-U. [ABSTRACT FROM AUTHOR]

Tungsten (W) has been considered as candidate plasma facing material (PFM) in nuclear fusion reactor. However, the severe work environment will cause radiation damage of W, which will affect its thermal energy transport. We constructed the structural model of undamaged and damaged W, and used the non-equilibrium molecular dynamics (NEMD) simulations to quantitatively study the thermal conductivity of W. The results show a decrease in lattice thermal conductivity with the increase of neutron radiation energy. Interstitial atoms trend to segregate at grain boundary, and the interstitial W atoms have a greater effect on lattice thermal conductivity than vacancies. Nanoscale-sized He bubbles cause a significant decrease in the lattice thermal conductivity of W, which down to - 25 % of its perfect crystal value when the porosity is 2.1%. These different defects cause different degrees of lattice deformation, which increase the phonon scattering and result in the decrease of lattice thermal conductivity. [ABSTRACT FROM AUTHOR]

Molecular dynamics simulation has become a powerful tool to deepen the understanding of the radiation damage mechanism of nuclear materials. Extracting point defects, analyzing their diffusion, and visualizing the defect dynamics in atomic simulation data are important, but challenging tasks to understand irradiation behavior. In the past, irradiation defects have been detected using the so-called Wigner–Seitz cell method and analyzed by the statistics of Frenkel pairs. However, traditional analysis modes blur the fine details of defect dynamics. In this paper, we present a visual analysis pipeline for domain scientists to comfortably explore radiation damage simulation data. We couple defect identification, defect clustering, molecule visualization, and tracking graph to form an integrated visual exploration approach. We describe the application of our approach in practice to study defect clustering in Ni–Fe alloy. With our proposed pipeline, defects can be extracted in a robust way, clusters can be visualized with a favorable representation, in-depth data analysis can be setup, and defect dynamics can be demonstrated in greater detail than previously possible. [ABSTRACT FROM AUTHOR]

Irradiation damage in materials for nuclear reactors, particularly for fusion reactors, is a serious problem. For example, the pressure vessel in a fission power plant becomes brittle after exposure to neutron irradiation for many years. In the case of fusion reactors, in addition to the increase in ductile-to-brittle transition temperature, irradiation-induced swelling occurs in structural materials that need to tolerate high-dose irradiation of several hundreds of displacements per atom (dpa). The irradiation of particles (such as neutrons, ions, and electrons) with high energy introduces a large number of point defects, i.e., self-interstitial atoms and vacancies, into materials. Such point defects aggregate together to form self-interstitial atom clusters as interstitial loop and vacancy clusters as void, stacking fault tetrahedra, or vacancy loop. Then, these clusters affect the microstructure and properties of materials. Moreover, these clusters play a more important role than individual point defects during the irradiation damage process. Even after research for decades, many questions about clusters remain unanswered partially because of the difficulties in irradiation test and cluster observation. This review paper explained the structures of clusters and the effect of clusters on irradiation damage in materials. As a unique research of this author’s group, the formation of vacancy-type dislocation loops with sizes of up to 100 nm in iron was introduced, including the effect of hydrogen and its isotope and the effect of alloy elements on the formation of vacancy-type dislocation loops. There are two different kinds of vacancy-type dislocation loops, i.e., those having a Burgers vector of b= and those having a Burgers vector of b=1/2. The density of the first type is approximately one order of magnitude higher than that of the second type. The one-dimensional motion of self-interstitial atom clusters and the research activities in this field were also discussed in detail, and the one-dimensional motion would be a key factor effecting the irradiation damage of high entropy alloys.

Salivary gland hypofunction causes significant morbidity and loss of quality of life for head and neck cancer patients treated with radiotherapy. Preventing hypofunction is an unmet therapeutic need. We used an adeno-associated virus serotype 2 (AAV2) vector expressing the human neurotrophic factor neurturin (CERE-120) to treat murine submandibular glands either pre- or post-irradiation (IR). Treatment with CERE-120 pre-IR, not post-IR, prevented hypofunction. RNA sequencing (RNA-seq) analysis showed reduced gene expression associated with fibrosis and the innate and humoral immune responses. We then used a minipig model with CERE-120 treatment pre-IR and also compared outcomes of the contralateral non-IR gland. Analysis of gene expression, morphology, and immunostaining showed reduced IR-related immune responses and improved secretory mechanisms. CERE-120 prevented IR-induced hypofunction and restored immune homeostasis, and there was a coordinated contralateral gland response to either damage or treatment. CERE-120 gene therapy is a potential treatment for head and neck cancer patients to influence communication among neuronal, immune, and epithelial cells to prevent IR-induced salivary hypofunction and restore immune homeostasis.

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.

With valuable physicochemical properties, yttria-stabilized zirconia crystal has promising advantages in optical applications. In this paper, the waveguide effect is observed in yttria-stabilized zirconia crystal irradiated by energetic oxygen ions. The waveguide properties and the field intensity are analyzed using prism and end-face coupling method arrangements, and the results show that monomode is found in the near-surface region and the light beam can be well confined in the waveguide structure, which shows refractive index distribution of the barrier-wall and enhanced-well type. The lattice damage induced by irradiation is investigated by the Rutherford backscattering/channeling experiment and high-resolution X-ray diffraction techniques. The simulation results are in good agreement with the experimental data. [ABSTRACT FROM AUTHOR]

An improved version of a recently developed stochastic cluster dynamics (SCD) method (Marian and Bulatov, 2012) [6] is introduced as an alternative to rate theory (RT) methods for solving coupled ordinary differential equation (ODE) systems for irradiation damage simulations. SCD circumvents by design the curse of dimensionality of the variable space that renders traditional ODE-based RT approaches inefficient when handling complex defect population comprised of multiple (more than two) defect species. Several improvements introduced here enable efficient and accurate simulations of irradiated materials up to realistic (high) damage doses characteristic of next-generation nuclear systems. The first improvement is a procedure for efficiently updating the defect reaction-network and event selection in the context of a dynamically expanding reaction-network. Next is a novel implementation of the τ-leaping method that speeds up SCD simulations by advancing the state of the reaction network in large time increments when appropriate. Lastly, a volume rescaling procedure is introduced to control the computational complexity of the expanding reaction-network through occasional reductions of the defect population while maintaining accurate statistics. The enhanced SCD method is then applied to model defect cluster accumulation in iron thin films subjected to triple ion-beam (Fe3+, He+ and H+) irradiations, for which standard RT or spatially-resolved kinetic Monte Carlo simulations are prohibitively expensive.

An improved version of a recently developed stochastic cluster dynamics (SCD) method (Marian and Bulatov, 2012) [6] is introduced as an alternative to rate theory (RT) methods for solving coupled ordinary differential equation (ODE) systems for irradiation damage simulations. SCD circumvents by design the curse of dimensionality of the variable space that renders traditional ODE-based RT approaches inefficient when handling complex defect population comprised of multiple (more than two) defect species. Several improvements introduced here enable efficient and accurate simulations of irradiated materials up to realistic (high) damage doses characteristic of next-generation nuclear systems. The first improvement is a procedure for efficiently updating the defect reaction-network and event selection in the context of a dynamically expanding reaction-network. Next is a novel implementation of the τ-leaping method that speeds up SCD simulations by advancing the state of the reaction network in large time increments when appropriate. Lastly, a volume rescaling procedure is introduced to control the computational complexity of the expanding reaction-network through occasional reductions of the defect population while maintaining accurate statistics. The enhanced SCD method is then applied to model defect cluster accumulation in iron thin films subjected to triple ion-beam (Fe3+, He+ and H+) irradiations, for which standard RT or spatially-resolved kinetic Monte Carlo simulations are prohibitively expensive.

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High entropy ceramics (HECs) are a novel class of ceramics that exhibit high melting point, excellent high-temperature mechanical performance, and superb corrosion resistance, making them promising candidates for advanced nuclear reactors. Notably, encouraging irradiation tolerance has been found in a few HECs based on the scattered reports in the current literature. In this article, we systematically review the recent progress in the irradiation response of HECs, including high entropy carbides, high entropy pyrochlore oxides, high entropy MAX phases, and high entropy nitride films. The influence of chemical complexity on the irradiation properties of HECs is examined by comparing them with their corresponding single-component ceramic counterparts. Besides, the similarities and differences between HECs and widely-studied high entropy alloys are discussed regarding their irradiation responses. In the end, we envisage prospects for the future development of irradiation-tolerant HECs. [ABSTRACT FROM AUTHOR]

Object kinetic Monte Carlo (OkMC) is a fundamental tool for modeling defect evolution in volumes and times far beyond atomistic models. The elastic interaction between defects is classically considered using a dipolar approximation but this approach is limited to simple cases and can be inaccurate for large and close interacting defects. In this work a novel framework is proposed to include exact elastic interactions between defects in OkMC valid for any type of defect and anisotropic media. In this method, the elastic interaction energy of a defect is computed by volume integration of its elastic strain multiplied by the stress created by all the other defects, being both fields obtained numerically using a FFT solver. The resulting interaction energies reproduce analytical elastic solutions and show the limited accuracy of dipole approaches for close and large defects. The OkMC framework proposed is used to simulate the evolution in space and time of self-interstitial atoms and dislocation loops in iron. It is found that including the anisotropy has a quantitative effect in the evolution of all the type of defects studied. Regarding dislocation loops, it is observed that using the exact interaction energy results in higher interactions than using the dipole approximation for close loops. [ABSTRACT FROM AUTHOR]

This study explores the delaying of the formation of helium bubbles and blisters in pure titanium by hydrogen pre-implantation. Titanium, implanted with helium (40 KeV, 5 × 1017 ions/cm²), exhibited large bubbles that cause exfoliation after heat treatment, whereas hydrogen pre-implantation inhibited bubble growth at room temperature and reduced the exfoliation after heat treatment. In the samples pre-implanted with hydrogen, we found evidence of helium diffusion delay by: (a) a fourfold reduction in bubble pressure (b) faceted cavities in the samples (c) a smaller increase in titanium lattice parameters (d) a 16-fold reduction in average bubble size and a sixfold reduction in bubble area fraction (e) a more than twofold decrease in exfoliation (f) a tendency toward the formation of larger bubbles as a result of heat treatment. We believe that it is reasonable to assume that the inhibition of helium diffusion between tetrahedral interstitial lattice sites takes place because of the occupation of the intermediate octahedral sites by hydrogen atoms. Evidence for the opposite effect, that is inhibition of the diffusion of hydrogen in the presence of helium, is found in the retention of hydrogen in the specimens at elevated temperatures. This retention allowed the existence of titanium hydride after heat treatment at 680 °C. The present study sheds light on the intricate interplay between hydrogen and helium in titanium, providing insights into mechanisms that can potentially mitigate helium-induced damage in materials. [ABSTRACT FROM AUTHOR]

• GB improves irradiation resistance via absorbing irradiation defects. • Universally exponent function between GB absorption rate and energetic state is captured. • GB resistance is strongly improved attributed to disorder-induced entropy change. • Ni segregation is particularly enhanced, contributing to high GB absorption rate. Multi-principal element alloys (MPEAs) demonstrate significant promise as structural materials for nuclear energy equipment owing to their exceptional mechanical properties and radiation-resistant performances. In these alloys, the grain boundary (GB) serves as a crucial microstructure that typically mitigates irradiation damage by absorbing the irradiation-induced defect. However, the micromechanisms governing the anti-irradiation performance of GBs in MPEAs remain unclear. In this study, we investigate the irradiation defect production during collision cascade in the model NiCoCr bicrystal system through atomic simulations, aiming to unveil the atomic-scale origin of GB to resist irradiation damage in MPEAs. The results reveal that GBs effectively serve as sinks for irradiation defects in NiCoCr. The sink efficiency depends on the GB energetic state, including GB excess energy and defect segregation energy, as well as the energetic difference between interstitial and vacancy segregation. Statistical analysis identifies a universally exponent function between the defect absorption rate at GB and GB energetic state. In NiCoCr, the GB-disorder-induced-entropy increase leads to a biased reduction in interstitial segregation energy, narrowing the gap between interstitial and vacancy segregation energies by approximately 11 % compared to Ni. This improvement enhances the overall resistance of GBs to irradiation damage. Additionally, preferential segregation of Ni interstitial atoms is notably enhanced in NiCoCr, contributing to a high defect absorption rate at GBs. This study provides new insights into the resistance of GBs to irradiation defects in MPEAs and suggests GB engineering as an effective strategy for developing advanced alloys with enhanced radiation tolerance. [Display omitted] [ABSTRACT FROM AUTHOR]

The dense Al 2 O 3 -ZrO 2 ceramic composite prepared by spark plasma sintering was irradiated by 500 keV He ions with different fluences and temperatures. The microstructural evolution and mechanical properties were investigated. The results showed that peak broadening and shifts at RT revealed by GIXRD and Raman are associated with damage induced microstrain and formation of point defects. The recovery at 500 °C suggested the reduction of irradiation induced damage. Compared with α-Al 2 O 3 , t-ZrO 2 exhibited a reverse trend in lattice parameters change and lattice expansion. Many helium bubbles with oblate and ribbon-like shape were mainly formed in α-Al 2 O 3 grains at He concentration peak at 1.0 × 1017 ions/cm2. With increasing of fluence at RT, ribbon-like helium bubbles developed into microcracks at 4.0 × 1017 ions/cm2. Though evident structural changes, no full amorphization was observed at 4.0 × 1017 ions/cm2. Formation of ribbon-like bubbles and microcracks is the main mechanism for degradation of mechanical properties of irradiated Al 2 O 3 -ZrO 2 composite. [ABSTRACT FROM AUTHOR]

The irradiation resistance of pre-deformed FeCu alloy was studied using a 3 MeV Fe ion irradiation experiment at room temperature in comparison with that of the as-received sample. Nanoindentation and atom probe tomography (APT) were used to characterize the mechanical properties and solute distribution. The stress–strain curve obtained by nanoindentation shows that the yield strength (σ0.2) of the pre-deformed sample is unexpectedly reduced with an increase in the irradiation dose to five displacements per atom (dpa). We suggest that it results both from the decrease in the dislocation density and the suppression of defects during irradiation. APT shows that the nucleation of the Cu cluster is suppressed; however, its growth is promoted in the pre-deformed sample, resulting in the formation of sparse and coarse clusters at 1 dpa irradiation. These coarse Cu clusters were then unexpectedly refined to finer grains with an increase in the irradiation dose to 5 dpa. Theoretically, the improvement in the resistance to irradiation in the pre-deformed sample is attributed to the dense point-defect sinks, that is, the dislocations and grain boundaries introduced by pre-deformation. In addition, the contributions of the dislocations and grain boundaries to the sink strength are estimated for both the as-received and pre-deformed samples. The results indicate that dislocations, rather than grain boundaries, play a major role after deformation. [ABSTRACT FROM AUTHOR]

This review paper provides insights the into current developments in additive manufacturing (AM) techniques. The comprehensive presentations about AM methods, material properties (i.e., irradiation damage, as-built defects, residual stresses and fatigue fracture), experiments, numerical simulations and standards are discussed as well as their advantages and shortages for the application in the field of nuclear reactor. Meanwhile, some recommendations that need to be focused on are presented to advance the development and application of AM techniques in nuclear reactors. The knowledge included in this paper can serve as a baseline to tailor the limitations, utilize the superiorities and promote the wide feasibilities of the AM techniques for wide application in the field of nuclear reactors.

We report a map of irradiation-induced recrystallization and polygonization in a CrFeCoNiCu high-entropy alloy obtained by a precession electron diffraction technique. This orientation map enables the visualization of depth-dependent recrystallization at a glance and reveals the effects of irradiation dose and material entropy. Discontinuous dynamic recrystallization preferentially occurs near the surface, where the irradiation dose is low, but radiation-enhanced diffusion dominantly occurs leading to dislocation climb even at room temperature. Interestingly, in the high-entropy phase with a relatively low stacking fault energy, discontinuous dynamic recrystallizastion is suppressed due to the lower stored energy and grain boundary diffusivity compared low-entropy one. Image, graphical abstract [ABSTRACT FROM AUTHOR]

The nearly infinite compositional design space of high entropy alloys (HEAs) presents many opportunities to improve performance in extreme environments, particularly for nuclear reactors. The ability of some HEAs to resist high amounts of radiation damage, while well documented, has not yet been fully exploited. We studied the irradiation effect of different ions (Si and Ni) on the microstructure and mechanical properties of the Al0.5CoCrFeNi alloy at its equilibrium state, and of the individual response of each phase to irradiation. The results show a stronger effect of Si ions and differences in response of the ductile FCC (A1) and brittle ordered BCC (B2) phases towards irradiation. This finding highlights the need to further probe unexplored compositional spaces in HEAs, as further optimization is likely to yield further compositional and microstructural stability with practical applications.

Ion irradiation was used as a surrogate approach to mimic radiation induced modification in iron phosphate glass network. To understand the synergetic effect of electronic and nuclear energy losses in glass network modification, ions at different energies were employed. The iron phosphate glasses were irradiated with Au ions at different ion energies between 750 keV and 20 MeV. The ion beam irradiated samples were characterised by employing different techniques.The Fe L2,3- edge X-ray absorption spectra of irradiated samples reveal reduction of Fe3+ to Fe2+, induced by ion beam irradiation. X-ray diffraction and electron microscopy imaging confirm formation of crystalline Fe3(P2O7)2, Fe4(P2O7)3, Fe(PO)4, and Fe(PO3)3 phases. The origin of these crystalline phases may be attributed to induced stress in irradiated samples. The observed decrease in hardness of irradiated samples was attributed to formation of non-bridging oxygen.

In order to realize low-defect fabrication of optical elements, detection and characterization methods of damage precursors related to laser damage performance were summarized herein. It was proposed that nanoscale damage precursors and chemical structural defects were the focus in low damage fabrication. Combined with the development of optical elements, methods for mitigating nanoscale damage precursor were discussed to effectively decrease nanoscale contaminations and chemical structural defects based on chemical-assisted magnetorheological polishing and ion beam sputtering cleaning technology. Three challenges in low-defect fabrication of optical elements in high influence were introduced, which might be a reference for the clean manufacturing of optical elements. [ABSTRACT FROM AUTHOR]

We have investigated the microstructural and crystallographic evolution of nanocrystalline zirconia under heavy ion irradiation using in-situ transmission electron microscopy (TEM) and have studied the atomic configurations of defect clusters using aberration-corrected scanning transmission electron microscopy (STEM). Under heavy ion irradiation the monoclinic-ZrO 2 is observed to transform into cubic phase, stabilised by the strain induced by irradiation-induced defect clusters. We suggest that the monoclinic-to-cubic transformation is martensitic in nature with an orientation relationship identified to be (100) m ∥(100) c and [001] m ∥[001] c. By increasing the damage dose, both the formation of voids and irradiation-induced grain growth were observed. A model for the formation of voids is proposed, taking defect interactions into consideration. The study has also demonstrated that high resolution orientation mapping by transmission Kikuchi diffraction (TKD) combined with in-situ irradiation in a TEM is a powerful method to probe the mechanisms controlling irradiation-induced processes, including grain boundary migration, phase transformations and texture evolution. Image, graphical abstract [ABSTRACT FROM AUTHOR]

Simulation of irradiation damage in a stainless-steel type 304L was studied by means of cold rolled samples that were then subjected to different annealing cycles to assess their recovery and compare it with the behavior observed in annealed neutron irradiated samples. Changes in hardness as a function of time and temperature were used to compute the activation energy associated with the recovery in hardness; such value coincided with those reported in literature for irradiated samples. The study was coupled with the evaluation of the degree of sensitization in the cold-rolled and annealed samples by means of the electrochemical potentiokinetic reactivation technique. Annealing the cold-rolled samples for 5 h at 500 °C resulted in the optimum treatment for reducing the hardness as it does not sensitize the steel. [ABSTRACT FROM AUTHOR]

Hexagonal boron nitride was irradiated with 140 MeV protons to a fluence of ~6 1020p/cm2 at T irr ~200 °C. Isotropic graphite was also irradiated alongside h-BN under similar conditions and fluence to enable direct comparison of the two similar structures. The effects of proton irradiation on dimensional stability and microstructure were studied using precision dilatometry and energy dispersive Xray diffraction techniques revealing by direct comparison to graphite that h-BN can better resist radiation damage from bombardment when irradiated with protons impinging normal to the crystallographic planes or along the crystallographic c-axis. X-ray diffraction experiments also revealed a preferred orientation of the crystallites in bulk samples near the sample surface, an orientation influenced by irradiation. Thermal studies using differential scanning calorimetry and thermogravimetric analysis to 740 °C, augmented by precision dilatometry, provided evidence of subtle phase transitions attributed to residual w-BN in the matrix. Irradiation appears to induce shifting of such transitions. [ABSTRACT FROM AUTHOR]

During their service in a nuclear power plant, structural materials are subject to high temperature, mechanical stress, corrosive coolant environments, and neutron irradiation. The combination of radiation, corrosion and thermo-mechanical loading represents a unique environment that materials must endure: radiation can significantly alter the material's thermo-mechanical performance and the lifetime of a component. The conditions under which this happens are difficult to study and normally necessitate nuclear reactor experiments followed by post-irradiation examination in hot cells, which renders investigations complex, time consuming and expensive. While few attempts have been made to study the combined effects of radiation and corrosion with lead bismuth eutectic [1] or molten salts [2] none of these experimental set-ups were designed to study the combination of corrosion and irradiation together with mechanical loading. Here the development of a miniature test technique is presented where irradiation, corrosion and strain can synergistically be applied on a material in a controlled way. The experiment is based on scaling down the membrane bulge test [3], using liquid lead as a pressure medium at high temperature. Membrane type specimens are made from the structural material to study, with their back side pressurized thereby inducing corrosion and membrane stretching stress while irradiation damage is generated by proton irradiation at the front surface affecting the entire thickness to the back side of the sample. The paper presents the way how the engineering problems related to the combination of corrosion by liquid metal, stress application and irradiation of the sample have been addressed. [ABSTRACT FROM AUTHOR]

In this paper, we study the mechanical behaviour of silicon carbide at the nanoscale, with a focus on the effects of grain orientation and high-dose irradiation. Grain orientation effect was studied through nanoindentation with the aid of scanning electron microscopy (SEM) and EBSD (electron backscatter diffraction) analyses. Mechanical properties such as hardness, elastic modulus and fracture toughness were assessed for different grain orientations. Increased plasticity and fracture toughness were observed during indentations on crystallographic planes which favour dislocation movement. In addition, for SiC subjected to irradiation, increases in hardness and embrittlement were observed in nanoindentations at lower imposed loads, whereas a decrease in hardness and an increase in toughness were obtained in nanoindentations at higher loads. Transmission electron microscopy (TEM) analyses revealed that the mechanical response observed at a shallow indentation depth was due to Ga ion implantation, which hardened and embrittled the surface layer of the material. With an increased indentation depth, irradiation-induced amorphization led to a decrease in hardness and an increase in fracture toughness of the material. [ABSTRACT FROM AUTHOR]

This research revealed the mechanisms of irradiation damage in the novel high entropy ceramic materials. (Zr 0.25 Ta 0.25 Nb 0.25 Ti 0.25)C high-entropy carbide ceramics (HECC) with a single-phase rock-salt structure was synthesized by spark plasma sintering, which was irradiated by 3 MeV Zr ions to 20 dpa at 25, 300, and 500 °C. X-ray diffraction analysis showed that (Zr 0.25 Ta 0.25 Nb 0.25 Ti 0.25)C maintained a high phase stability without phase transformation after irradiation. About 0.2% lattice parameter expansion was revealed. The irradiation-induced microstructures were comprised of defect clusters with diameters of several nanometers, without void formation or radiation-induced segregation. The defect clusters were characterized by transmission electron microscopy as two types of dislocation loops, including perfect loops with Burgers vectors of b = a /2<1 1 0> and faulted Frank loops with Burgers vectors of b = a /3<1 1 1>. The growth of dislocation loops may be suppressed by the strong local lattice distortion. Nanoindentation tests showed irradiation-induced hardness increase, which was possibly caused by dislocation loops and lattice strain. Overall, the high irradiation resistance, along with other excellent physical properties makes HECC promising structural materials for advanced reactor designs. Image, graphical abstract [ABSTRACT FROM AUTHOR]

The mechanical properties of irradiated iron are studied by molecular dynamics simulation. The initial models are irradiated with the energy of primary knocked-on atoms (PKA) of 10 keV at 100, 300, 500 and 600 K, and then all the irradiated models are subjected to tensile test. The results reveal that the mechanical properties of irradiated iron are changed compared with un-irradiated iron, the yield stress and strain decrease after irradiation, and the irradiation causes the hardening of micro-structure at low temperature and high temperature, but it results in the softening of structure at room temperature. The plastic reduces for irradiated iron under tensile test, more stacking faults are formed in the crystal structure as the temperature increases. [ABSTRACT FROM AUTHOR]

This research revealed the irradiation hardening mechanism in two novel NbZrTiV and VNbMoTaWTi refractory high entropy alloys (RHEAs). The RHEA sample irradiated by a dual-beam of 0.5 MeV protons and 2 MeV He ions at 300 K. Irradiation-induced defects were characterized by transmission electron microscope (TEM). It was discovered that the defect clusters were mostly consisted of nano-sized dislocation loops and bubbles. The density of defects was higher in the peak region than in the plateau region for both irradiated samples. The effects of irradiation-induced defects on hardening were evaluated by the nanoindentation test and dispersed barrier hardening (DBH) model. The results show that the peak region contributes more to the hardening than the plateau region in both irradiated alloys. The DBH hardness increments agree well with those measured by nanoindentation. Dual-beam irradiation led to ∼14% hardening for NbZrTiV and ∼ 18% for VNbMoTaWTi. The effects of the chemical disorder and lattice distortion induced by compositional complexity on defect formation and hardening were discussed. The excellent irradiation resistance makes RHEAs to be potential candidate materials used as core component in advanced nuclear reactors. • The irradiation effects of two novel BCC RHEAs were studied. • The relationship between irradiation hardening and irradiation-induced defects was established. • The mechanism of compositional complexity on irradiation tolerance of BCC RHEAs was discussed. [ABSTRACT FROM AUTHOR]

We studied the morphology of the tracks (irradiation damage) of 44, 66, and 84 MeV Au ions in YBa 2 Cu 3 O 7-x (YBCO) thin films using cross-sectional transmission electron microscopy (TEM). The electronic energy loss (S e = 11.6, 16.7, and 20.1 keV/nm) ranged near the onset of ion-track formation previously determined by plan-view TEM and critical-temperature measurements (about 13 keV/nm). The morphology changed from 1D arrays of large spheres of about 5 to 10 nm diameter in 44-MeV irradiation to their elongated variant (spheroid) that were statistically pierced by thin-lines of about 1 to 2 nm and 3 to 5 nm diameter in 66 and 84 MeV, respectively. In the same series of Au-irradiated YBCO films, we observed a sudden increase in critical current density at self-field and 77 K between 44- and 55-MeV irradiation, in agreement with the appearance of the thin-line damage in TEM (between 44- and 66-MeV). [ABSTRACT FROM AUTHOR]

Self-interstitial atoms (SIA) are generated in collision cascades during high-energy particle irradiation of crystalline materials. In pure metals, SIA generally adopt split configurations and display fast mobilities along one-dimensional trajectories. This differentiates them from vacancies, which move in confined three-dimensional paths. This asymmetry is one of the pillars of irradiation damage theories that have been successful in explaining a number features of the irradiation response of pure metals and dilute alloys. However, in complex concentrated alloys consisting of several metallic elements in equal or near equal proportions, lattice distortions associated with compositional fluctuations change the potential energy landscape on atomic scales, leading to SIA structures not seen in their pure metal counterparts. In this paper, we use atomistic simulations to study the properties of self-interstitial atoms in the quaternary equiatomic Nb–Mo–Ta–W refractory alloy. Chemically, these SIA defects adopt a variety of structures involving all pairs of atoms. We find the 〈 111 〉 orientation to be the most common among all split configurations, but we also observe – surprisingly – a relatively high occurrence of octahedral SIA. In terms of their diffusivities, we find two clearly distinguished regimes below and above 600 K, where the SIA diffusion changes dimensionality from 1D to 3D. We calculate the migration energies and diffusion pre-factors in both regions, from which we extract the translational and rotational components of the defect migration. We find values of 0.25 eV and pre-factors of ∼ 1 0 − 11 m 2 s−1 in the low temperature regime, and 0.57 eV and ∼ 1 0 − 8 m 2 s−1 in the high temperature one. From this, we estimate a rotational energy barrier of 0.37 eV. [ABSTRACT FROM AUTHOR]

• Effect of helium plasma exposure test on the surface and interior microstructure of W-0.5 % wt Y 2 O 3 was studied. • W matrix had thick, fuzzy tendrils on its surface, whereas Y 2 O 3 particles showed no fuzzy tendrils, but rather protrusions and small cracks. • W matrix has a complete tungsten crystal structure with fuzzy tendrils on its surface, and these tendrils are filled with uniformly distributed helium bubbles; the surface of the fuzzy tendrils has undergone amorphization. • In Y 2 O 3 particles, larger and fewer helium bubbles are present near the surface, while the smaller and denser bubbles are located further away. • The distribution of helium bubbles in W matrix is deeper and larger than that of Y 2 O 3 particles, indicating that He in W matrix are more prone to migration and bubbles are easier to growth. Tungsten is the preferred material for the first wall. In this paper, W-0.5 %wt Y 2 O 3 was selected for the exposure test of helium plasma, so that fuzz tendrils of considerable thickness were formed on its surface. By using ultrasonic dispersion, the tendril dispersion was obtained and observed through TEM. The fuzz tendrils on the W matrix surface were seen to have a complete tungsten crystal structure, along with bubbles of different sizes. The outer layer of the tendril is composed of an amorphous material, which is visible on the surface of the fuzz tendril. This is likely due to the subsequent of helium plasma exposure damage. Fuzz tendrils only appeared on the W matrix, whereas Y 2 O 3 particles created nano-protrusions and samll cracks on the surface. The nano-protrusion morphology and EDS analysis, including the lack of Y in fuzz tendrils, indicate that Y 2 O 3 particles were unable to form a fuzzy structure. It is evident that the defects distribution of W matrix and Y 2 O 3 particles is altered due to helium plasma exposure. Near the surface of Y 2 O 3 particles, fewer but larger bubbles were observed, whereas further away, smaller but denser bubbles were seen. In W matrix, bubbles were formed at a deeper level and were larger than the deepest bubbles in Y 2 O 3 particles, likely caused by the lower migration energy. The different responses of W matrix and Y 2 O 3 particles to helium plasma exposure have combined to further reveal the potential of increasing the plasma exposure resistance of plasma-oriented tungsten based materials. [ABSTRACT FROM AUTHOR]

We developed and validated an accurate inter-atomic potential for molecular dynamics simulation in cubic silicon carbide (3C-SiC) using a deep learning framework combined with smooth Ziegler–Biersack–Littmark (ZBL) screened nuclear repulsion potential interpolation. Comparisons of multiple important properties were made between the deep-learning potential and existing analytical potentials which are most commonly used in molecular dynamics simulations of 3C-SiC. Not only for equilibrium properties but also for significant properties of radiation damage such as defect formation energies and threshold displacement energies, our deep-learning potential gave closer predictions to the DFT criterion than analytical potentials. The deep-learning potential framework solved the long-standing dilemma that traditional empirical potentials currently applied in 3C-SiC radiation damage simulations gave large disparities with each other and were inconsistent with ab initio calculations. A more realistic depiction of the primary irradiation damage process in 3C-SiC can be given and the accuracy of classical molecular dynamics simulation for cubic silicon carbide can be expected to the level of quantum mechanics. [Display omitted] [ABSTRACT FROM AUTHOR]

Thrombocytopenia is a critical complication after radiation therapy and exposure. Dysfunction of megakaryocyte development and platelet production are key pathophysiological stages in ionizing radiation (IR)-induced thrombocytopenia. Protein kinase C (PKC) plays an important role in regulating megakaryocyte development and platelet production. However, it remains unclear how PKC regulates IR-induced megakaryocyte apoptosis. In this study, we found that pretreatment of PKC pan-inhibitor Go6983 delayed IR-induced megakaryocyte apoptosis, and inhibited IR-induced mitochondrial membrane potential and ROS production in CMK cells. Moreover, suppressing PKC activation inhibited cleaved caspase3 expression and reduced p38 phosphorylation levels, and IR-induced PKC activation might be regulated by p53. In vivo experiments confirmed that Go6983 promoted platelet count recovery after 21 days of 3 Gy total body irradiation. Furthermore, Go6983 reduced megakaryocyte apoptosis, increased the number of megakaryocyte and polyploid formation in bone marrow, and improved the survival rate of 6 Gy total body irradiation. In conclusion, our results provided a potential therapeutic target for IR-induced thrombocytopenia. [ABSTRACT FROM AUTHOR]