Radiation-induced Defects Research Articles

EXAFS Study of Radiation-Induced Defects in Molybdenum and Tantalum Metals

EXAFS Study of Radiation-Induced Defects in Molybdenum and Tantalum Metals

Effect of native carbon vacancies on evolution of defects in ZrC1- under He ion irradiation and annealing

The dynamic study of radiation-induced defects with annealing is critical for the material design for next-generation nuclear energy systems. The native vacancy could affect the development of defects, which lacks study. In the present work, the as-hot pressed ZrC1-x (x=0, 0.15, 0.3) ceramics which comprised crystallites of a few microns in size with different amounts of carbon vacancies were irradiated by 540 keV He2+ ions at room temperature with a fluence of 1 × 1017/cm2. The radiation-induced lattice expansion was found to be a common phenomenon in a sequence of ZrC0.85≥ZrC1.0>ZrC0.7. Both X-ray and electron diffractions confirmed maintenance of structural integrity without amorphization after irradiation. Inside the irradiated region, only “black-dot” type defects, i.e., clusters of point defects were observed while no helium-induced cavities, cracks, or extended dislocations were detected. The as-irradiated ZrC1-x were then annealed at different high temperatures. Upon annealing at 800 °C, very tiny helium-induced cavities were found to be generated and the crystal lattice recovered to a great extent, especially for the sub-stoichiometric samples. While annealed at 1500 °C, all the samples almost fully recovered the crystal lattices close to those of as-hot pressed ones. Meanwhile, large cavities and extended dislocations were generated. With increasing amount of native carbon vacancies, the size of cavities increased while the length and density of extended dislocations decreased. Inverse changes of lattice parameters during irradiation and annealing processes have been interpreted by the kinetics of defects. Finally, the correlation between native vacancies and damage behavior is discussed.

Review of Radiation-Induced Effects on β-Ga2O3 Materials and Devices

β-Ga2O3 has become an ultimate choice of emerging new-generation material for its wide range of compelling applications in power electronics. In this review, we have explored the available radiations in the atmosphere and the effects of radiation on the β-Ga2O3 material and devices. The focus in this review summarizes various studies covering different radiation such as swift heavy ions, protons, neutrons, electrons, Gamma, and X-rays to understand the radiation-induced effects on the structure and their reliable performance in harsh environments. In addition, we focused on the various pre-existing defects in β-Ga2O3 and the emergence of radiation-induced defects that provoke a severe concern, especially from the device performance point of view. This review presents the irradiation-induced effects on the devices such as high-power devices such as Schottky barrier diodes (SBDs), field-effect transistors (FETs), metal-oxide-semiconductor (MOS) devices, and photodetectors. Some key studies including the changes in carrier concentration with a removal rate, Schottky barrier height (SBH), ideality factor, defect dynamics dielectric damage, interface charge trapping, a thermally activated recovery mechanism for charge carriers at elevated temperature, and diffusion length of minority charge carriers. These reports show that β-Ga2O3-based devices could be deployable for space or high-radiation terrestrial applications. These results provide/suggest a better device design based on the radiation degradation studies in the state-of-the-art β-Ga2O3 devices.

EPR-active dimer centers with S=1 in α-Al2O3 single crystals irradiated by fast neutrons

Corundum (α-Al2O3) possesses a number of unique properties, including high tolerance to harsh radiation environment. Material functionality is strongly affected by radiation-induced structural defects, single interstitial-vacancy Frenkel pairs and their aggregates. Three novel paramagnetic two-electron defects with total spin S = 1 have been revealed in fast neutron irradiated α-Al2O3 single crystals by means of the EPR method. Analysis of the EPR characteristics of these defects allows to conclude that the main novel D1 defect (initial concentration after irradiation about 2.8 × 1016 cm−3) is oriented along one of three equivalent b crystal axes, that connect two adjacent anion sites of oxygen triangles on the base corundum plane. The D1 defect is ascribed to the F22+ center (two adjacent single-charged oxygen vacancies), while the D2 and D3 have the same dimer structure with an additional defect in their vicinity. The suggested defect structure is supported by pulse annealing of the EPR signal of relevant defects: defect concentration dependence contains a rising stage at 500–630 K, while complete thermal destruction due to the recombination with complementary oxygen interstitials occurs by 720–750 K. Such behavior qualitatively coincides with earlier studied thermal annealing of the F22+ center optical absorption in neutron-irradiated corundum.

Characterization of ion-induced microstructural changes in oxygen irradiated Ti–6Al–4V

The change in microstructure and mechanical properties ofTi–6Al–4V alloy due to 160 MeV O6+ ion irradiation has been reported here. Detailed X-ray diffraction line profile analyses have been used to evaluate the microstructural parameters and characterize the nature of defects and their arrangements. The results indicate saturation of size of the coherent region, the dislocation density and its arrangement with dose, which points to an optimized final microstructure of irradiated material. This was also supported by the Vicker’s microhardness measurements. Electron back scattered diffraction study was used to assess the changes in microstructure in terms of the evolution of misorientation boundaries within the grains due to irradiation-induced defects. All the results support the radiation resistance nature of Ti–6Al–4V alloy. Finally, comparison has been made with earlier proton irradiation results, and the difference has been related to the nature of the displacement cascade produced by light and heavy ions.

The effect of helium on cavity swelling in dual-ion irradiated Fe and Fe-10Cr ferritic alloys

Ferritic-martensitic (FM) steels for in-vessel components in proposed fusion reactors are expected to suffer from high levels of displacement damage and helium (He) generation by neutron transmutation. However, a thorough understanding of the He synergistic effects on the cavity swelling in FM steels is still not well established. To gain fundamental insights into the He effect on cavity swelling, high purity Fe and Fe-10 wt.% Cr ferritic model alloys were irradiated with 8 MeV Ni ions and co-implanted He ions at 400 to 550 °C up to 30 displacements per atom (dpa) with He implantation rates of 0.1, 10 and 50 appm He/dpa. The current study focuses on the 50 appm He/dpa, 500 °C behavior vs. 0.1 and 10 appm He/dpa. Irradiation-induced defects, including cavities, dislocation loops, and dislocation networks were characterized using transmission electron microscopy (TEM). In the grain interior, a bimodal cavity size distribution was observed in the 10 and 50 appm He/dpa samples, but not for 0.1 appm He/dpa. Cavity swelling was maximized at intermediate He implantation rates of ∼10 appm He/dpa for both ion-irradiated Fe and Fe-10Cr alloys. The cavity swelling behavior as a function of He implantation rate appears to be controlled by the He/dpa-dependent variation of cavity sink strengths. Treating small bubbles as biased sinks for interstitial absorption can significantly increase the ratio of biased to unbiased sink strengths (Q) and results in maximized cavity swelling for a Q ratio close to one.

Influence of ion irradiation-induced defects on phase formation and thermal stability of Ti0.27Al0.21N0.52 coatings

The influence of changes induced by ion irradiation on structure and thermal stability of metastable cubic (Ti,Al)N coatings deposited by cathodic arc evaporation is systematically investigated by correlating experiments and theory. Decreasing the nitrogen deposition pressure from 5.0 to 0.5 Pa results in an ion flux-enhancement by a factor of three and an increase of the average ion energy from 15 to 30 eV, causing the stress-free lattice parameter to expand from 4.170 to 4.206 Å, while the chemical composition of Ti0.27Al0.21N0.52 remains unchanged. The 0.9% lattice parameter increase is a consequence of formation of Frenkel pairs induced by ion bombardment, as revealed by density functional theory (DFT) simulations. The influence of the presence of Frenkel pairs on the thermal stability of metastable Ti0.27Al0.21N0.52 is investigated by scanning transmission electron microscopy, differential scanning calorimetry, atom probe tomography and in-situ synchrotron X-ray powder diffraction. It is demonstrated that the ion flux and ion energy induced formation of Frenkel pairs increases the thermal stability as the Al diffusion enabled crystallization of the wurtzite solid solution is retarded. This can be rationalized by DFT predictions since the presence of Frenkel pairs increases the activation energy for Al diffusion by up to 142%. Hence, the thermal stability enhancement is caused by a hitherto unreported mechanism - the Frenkel pair impeded Al mobility and thereby retarded formation of wurtzite solid solution.

Au-ion irradiation effects on microstructure and deuterium permeation resistance of Al/Al2O3 coating

Al/Al2O3 coating with a total thickness of approximately 3 μm was irradiated by 6 MeV Au-ions with the damage level from 13 to 39 dpa at room temperature. The results showed that the surface roughness of the Al/Al2O3 coating decreased with increasing irradiation damage levels. As the irradiation damage level is up to 39 dpa, the crystallinity of the Al2O3 layer is enhanced, and the irradiation-induced defects such as voids and dislocations appeared in the Al/Al2O3 coating. The decreased hardness and the deterioration of deuterium permeation resistance are strongly associated with the changed microstructure. The underlying mechanism for the evolution of the microstructure and properties induced by Au-ion irradiation is discussed.

Molecular dynamics study of interactions between edge dislocation and irradiation-induced defects in Fe–10Ni–20Cr alloy

Irradiation-induced defects frequently impede the slip of dislocations, resulting in a sharp decline in the performance of nuclear reactor structural materials, particularly core structural materials. In the present work, molecular dynamics method is used to investigate the interactions between edge dislocations and three typical irradiation-induced defects (void, Frank loop, and stacking fault tetrahedron) with the sizes of 3 nm, 5 nm, and 7 nm at different temperatures in Fe–10Ni–20Cr alloy. The critical resolved shear stresses (CRSSs) are compared among different defect types after interacting with edge dislocations. The results show that the CRSS decreases with temperature increasing and defect size decreasing for each defect type during the interaction with edge dislocations, except for the case of 3-nm Frank loops at 900 K. According to a comparison, the CRSS in Frank loop is significantly higher than that of others of the same size, which is due to the occurrence of unfaulting and formation of superjog or stacking-fault complex during the interaction. The atomic evolution of irradiation-induced defects after interacting with dislocations can provide a novel insight into the design of new structural materials.

Formation of stacking fault pyramid in zirconium

Zirconium alloys were widely used as fuel cladding in nuclear reactors. Stacking fault pyramid (SFP) is an irradiation-induced defect in zirconium. In this work, the formation process of SFP from a hexagonal vacancy plate on basal plane is studied by molecular dynamics (MD) simulations. The results show that, during the SFP formation from a basal vacancy plate, the 12[0001] (<c>/2) dislocation is firstly dissociated into two partial dislocations 118<202¯3> and 19<1¯013>. The former one resides on the basal plane, while the latter one glides on the first-order pyramidal plane. The partials on adjacent pyramidal planes react further and form a partial dislocation on the pyramidal edge, i.e. 19[101¯3¯]+19[01¯13]→19[11¯00]. A critical edge length of the initial vacancy plate is observed, below which perfect SFP is formed while above which truncated SFP is formed. The critical edge length increases with the increasing temperature. Under a compressive stress, the SFP collapses into an <c>/2 dislocation loop and then becomes a faulted loop. Under shear stress, the formation of SFP is facilitated, i.e. the critical edge length increases with the increasing shear stress. The current work is useful for understanding the irradiation effects in zirconium alloys.

Molecular dynamics simulations of displacement cascade near precipitate in zirconium alloys

Precipitates play an important role in the evolution of irradiation-induced defects and mechanical property of irradiated metals. In this work, the effects of a Zr2Cu precipitate on the production and subsequent evolution of cascade-induced point defects (vacancies and interstitials) in ZrCu alloy were investigated by molecular dynamics simulations at room temperature. The simulation results show that the precipitate increases the number of residual point defects at the end of cascade. However, most of the residual defects reside in the precipitate and near precipitate boundary. In the matrix, more interstitials survive than vacancies. In addition, a defect-free region is seen in matrix and near precipitate boundary, indicating that the precipitate boundary traps vacancies/interstitials in matrix and retains them in the precipitate. To explain these observations, the formation energy and migration energy of point defects are calculated. The formation energies of both vacancy and interstitial decrease as they are moving to the precipitate boundary from matrix and precipitate, indicating that they can be absorbed by precipitate boundary and result in the defect-free region. However, the migration barrier of vacancy is lower than that of interstitial. Therefore, more vacancies in matrix are absorbed. The current work provides new insights into understanding the irradiation effects in zirconium alloys.

Evaluation of interface thermal transport of Cu-Graphene nanocomposite under cascade overlaps and its effects on irradiation-induced defects

The materials used in neutron irradiation experiments are typically irradiated up to high doses. Thus, the density of defects in these materials is so high that subsequent collision cascades are overlapping with previously cascade-induced defects. The Cu-Graphene nanocomposite (NC) has been proposed as a high irradiation tolerance material. Its interface region can act as an effective sink to point defects. The interfacial thermal resistance (ITR) and interfacial thermal conductance (ITC) of NC are calculated using non-equilibrium molecular dynamics (NEMD) simulations. Single cascade and cascade overlaps induced by 6 keV primary knocked-on atoms (PKAs) in NC are studied using molecular dynamics (MD) simulations. Results show that Graphene loses its morphology during irradiation. Thus, these phenomena may be the main reason for the reduction and rise of ITC and ITR after the first cascade relative to those of un-irradiated NC, respectively. We found that with increasing the number of cascade overlaps the number of irradiation-induced defects in the upper bulk (bulku) region that cascade ends up in this region of NC increases, but the values of ITR and ITC exhibit a slight fluctuation. By evaluating the effects of increasing the ambient temperature and number of cascades on ITC, and defects in the bulku region, we found that the suspended carbon chains in Graphene break by irradiation, and the damaged area of Graphene gradually aggravates with increasing the system temperature. Also, at higher temperatures, the number of defects in the bulku region for the cases two and three cascades increases with increasing temperature. These results imply that the damage of Graphene does not destroy the role of the interface region in healing the irradiation-induced defects in the lower bulk (bulkl) region at different temperatures, but can affect thermal transport properties and consequently, change the number of defects in the bulku region of NC.

Superior radiation tolerance via reversible disordering-ordering transition of coherent superlattices.

Materials capable of sustaining high radiation doses at a high temperature are required for next-generation fission and future fusion energy. To date, however, even the most promising structural materials cannot withstand the demanded radiation environment due to irreversible radiation-driven microstructure degradation. Here we report a counterintuitive strategy to achieve exceptionally high radiation tolerance at high temperatures by enabling reversible local disordering-ordering transition of the introduced superlattice nanoprecipitates in metallic materials. As particularly demonstrated in martensitic steel containing a high density of B2-ordered superlattices, no void swelling was detected even after ultrahigh-dose radiation damage at 400-600 °C. The reordering process of the low-misfit superlattices in highly supersaturated matrices occurs through the short-range reshuffling of radiation-induced point defects and excess solutes right after rapid, ballistic disordering. This dynamic process stabilizes the microstructure, continuously promotes in situ defect recombination and efficiently prevents the capillary-driven long-range diffusion process. The strategy can be readily applied into other materials and pave the pathway for developing materials with high radiation tolerance.

Computational Study of Low-Energy Pt-Ion Implantation into Graphene for Single-Atom Catalysis

Single-atom catalysts (SACs) represent the ultimate goal of nanocatalysis fields. However, complex synthesis processes and pyrolysis inactivation problems are the two main challenges that plague the development of SACs. In this work, we propose that the ultralow-energy ion-implantation (ULEII) method could be utilized to simply and efficiently synthesize stable SACs. Our simulation results of Pt-ion implantation into graphene indicate that the total doping efficiency, including direct displacement doping and indirect trap doping, can be effectively optimized by delicately adjusting the energy of incident ions. Further systematic molecular dynamics simulations and first-principles calculations demonstrate that irradiation-induced vacancy defects can effectively capture and anchor adsorbed metal atoms on the graphene surface. The stability and migration characteristics of various defects are also clearly elucidated. Theoretically, by selecting an optimal ion energy, the ULEII method can achieve a doping efficiency as high as 73.4% .

The role of carbon allotropes on the radiation resistance of Cu-based nanocomposites: An atomistic, energetic, and thermodynamic perspective

Carbon allotropes can be considered as an excellent radiation resistance enhancer in metal-Graphene nanocomposites (NCs). Many research groups studied the radiationresistanceand interface stability of metal-Graphene NCs and revealed that Graphene can improve the radiation resistance of NCs under irradiation. Other allotropes of carbon have not been studied up to now. Therefore, in this work, four Cu-based NCs such as Cu-Graphene, Cu-Graphyne, Cu-Graphdiyne, and Cu-Graphane were studied by molecular dynamics (MD) to understand the role of carbon allotropes on the radiation resistance of Cu-based NCs. Compared with pure copper; four Cu-based NCs have fewer residual defects in the bulk region after irradiation. The results demonstrated that the interface of these NCs acts as a sink for radiation-induced defects, and preferentially traps interstitials over vacancies. The results of energetic calculations indicate the defect formation energy is reduced in the vicinity of interface regions. Compared with Cu-Graphyne and Cu-Graphdiyne, Cu-Graphene and Cu-Graphane have low segregation energy for interstitial emission mechanism to annihilate vacancies. Also, Cu/Graphane/Cu interface has a higher strength of interaction and attraction with vacancies due to the low value of vacancy formation energy (Evac). Thermodynamic and structural analyses reveal that Graphdiyne plane isn’t a stable plane during collision cascades. The results of this study can provide a fundamental perspective on the radiation resistance of Cu-based NCs including different carbon allotropes to select the best allotrope to improve the radiation resistance of NC for using in extreme radiation environments.

Radiation-induced segregation on dislocation loops in austenitic Fe-Cr-Ni alloys

Radiation-induced segregation of alloying elements to crystallographic defects is commonly observed in irradiated austenitic stainless steels. The interaction between solutes and radiation-induced defects changes the physical distribution of solutes and thus affects the formation and growth of defects. The change of the microstructure consequently affects the mechanical properties of the material. A qualitative and quantitative understanding of the interaction between solutes and defects is desirable to better predict the service lifetime of nuclear materials. We used atom probe tomography to measure the distribution of solutes at dislocation loops in 304L stainless steel, irradiated with 2 MeV protons up to 1.5 displacements per atom at 373 and 633 K. No segregation at dislocation loops was found in samples irradiated at 373 K, whereas Ni and Si enrichment and Cr depletion were detected at dislocation loops irradiated at 633 K. The experimentally observed perfect and faulted dislocation loops in vacancy and interstitial types were reproduced by molecular dynamics (MD). A hybrid MD/Monte Carlo method was used to predict the redistribution of alloying atoms at all possible types of dislocation loops in face-centered cubic Fe-Cr-Ni alloys at the same irradiated temperatures (373 and 633 K). The simulations show that, at both temperatures, Cr clusters were formed and distributed randomly, and Ni atoms enriched or depleted at interstitial or vacancy dislocation loops, respectively. The change of solute concentration reaches the highest at the edge of the loop. Ni profiles exhibit characteristic behavior in terms of the stress field of the loops: tension inside of vacancy loops showing depletion of Ni atoms compared with compression inside of interstitial loops showing enrichment of Ni atoms. In addition, the stress field is reduced after solute redistribution. The absence of alloying segregation observed in experiments at a lower temperature (373 K) is explained by a rate theory model: Low-temperature irradiation requires significantly longer irradiation time to see the same amount of segregation as at high temperatures because of the extremely low diffusion of vacancies at low temperatures.

Nanostructured TiC dispersion-strengthened tungsten composite with remarkably improved He ion irradiation resistance

Improving the irradiation resistance of W alloys is vital to their application of plasma facing materials in future fusion reactors. In this work, quasi-spherical tungsten nanoparticles synthesized by thermal plasma were mixed with nanosized TiC to sinter nanostructured TiC dispersion-strengthened tungsten alloy. As-obtained W-TiC composite exhibited high relative density of 99.14% with suppressed grain size of about 1.12 μm, together with well-dispersed TiC with size 50–100 nm into W grains. More importantly, semi-coherent W/TiC interface with diffusion layer of about 25 nm was formed, which highlights the tight connection between W and TiC. Furthermore, He ions irradiation results indicated that W-TiC composite exhibited excellent irradiation resistance with a damage depth of about 9.53 μm, which is much lower than 18.26 μm of pure W. Mechanism investigation indicated that the remarkably improved He ion irradiation resistance was attributed to the efficient capture and annihilation of irradiation-induced defects by grain boundaries and phase interfaces.

Characterization of microstructure and hardening of SLM nickel-based alloy irradiated by He ions

For the application of additively manufactured nickel-based alloys in molten salt reactors (MSRs), it is crucial to evaluate their performance in nuclear environments, especially under He-induced damage. Therefore, in this study, the irradiation damage behaviour of selective laser melted (SLM) nickel-based GH3536 alloy was studied by irradiating it with He ions at various doses. Traditionally manufactured (TM) materials were used as reference materials to evaluate the resistance of SLM GH3536 to irradiation. Transmission electron microscopy (TEM) was performed to reveal the depth distribution of irradiation-induced He bubbles, as well as the characteristics of dislocation loops, using both on-zone scanning transmission electron microscopy (STEM) and rel-rod modes. Nanoindentation tests were performed on both alloys to investigate their irradiation hardening. The TEM graphs and quantitative analysis showed that, owing to the cellular sub-grain structures, the He bubbles in the SLM alloy exhibited a lower number density but larger size than those in the TM alloy. Additionally, the SLM alloy showed inferior He tolerance but better resistance to irradiation hardening, as revealed by the nanoindentation test results. Consistency between the experimental hardness increment and calculated yield stress increment showed that the irradiation hardening of the sample was mainly caused by the presence of irradiation-induced defects. A comparison between the on-zone STEM and rel-rod modes suggested that the rel-rod mode is more effective if only the effect of irradiation-induced defects on mechanical properties is studied. Furthermore, this study provides insight into the design of SLM nickel-based alloys for nuclear industry applications.

Observation of gamma radiation-induced defects in radiation-intolerant and radiation-tolerant Si/SiO2 by electron spins resonance and quantitative interpretation

The gamma radiation-induced defects in radiation intolerant and radiation tolerant Si/SiO2 were observed by electron spin resonance (ESR). The Pb and È defects appeared in radiation-intolerant Si/SiO2 after radiation, but only Pb defects appeared in radiation-tolerant Si/SiO2 before and after radiation. The experimental results showed that these defects were correlated with the way of oxidation, the dosage of 60Co radiation and the electrical bias field during radiation. The higher the radiation dosage, the higher the defect concentration. A quantitative relationship was found between defect concentration and radiation dosage. Besides, the △B (peak to peak) of Pb and È indicated that Pb is the defect with a slow electron spin relaxation time, while È is the defect with a fast electron spin relaxation time. Finally, the experimental results are explained qualitatively.

Microstructure evolution and lead-bismuth eutectic corrosion behavior of 11Cr1Si ferritic/martensitic steel after proton irradiation

The effect of proton irradiation fluence on microstructure evolution and lead-bismuth eutectic (LBE) corrosion behavior of 11Cr1Si ferritic/martensitic (F/M) steels was investigated. The results revealed that dislocations and black dots were the two main types of TEM-visible defects of all irradiated samples. Proton irradiation did not alter the duplex structure of the oxide layer but significantly increased the corrosion rate, and this was attributed to the irradiation-induced defects responsible for the irradiation-enhanced outward diffusion of Fe and inward diffusion of O atoms.