Sinks For Defects Research Articles

Radiation response of multicomponent L12 γ′ precipitates strengthened high entropy alloys: The role of γ/γ′ interface

The advancement of high entropy alloys (HEAs) has stimulated the development of multicomponent L12 γ′ intermetallics, which exhibit both excellent strength and ductility. These intermetallics with long-range order are typically employed as coherent precipitates to enhance the high-temperature performance of HEAs. This study investigates the influence of multicomponent γ′ intermetallics on the irradiation response of the strengthened HEA γ phase, focusing on the role of the multicomponent γ/γ′ interface in defect production and evolution. Our results indicate that the chemical disorder within the multicomponent γ′ phase leads to a broad defect energy spectrum near the interface zone, diminishing its effectiveness as a defect sink. Based on these observations, we propose that, distinct from traditional superalloys, the γ/γ′ interface plays a minor role in the irradiation resilience of multicomponent γ′ intermetallics-strengthened HEAs. Instead, the inherent chemical disorder within the multicomponent γ′ intermetallics emerges as the key factor.

Effects of Cold Rolling–Induced Defects on Proton Irradiation Response in Nb-1Zr-0.1C Alloy

This work highlights the effect of proton irradiation in the presence of preexisting defects in the Nb-1Zr-0.1C alloy. To introduce the initial defects, the Nb-1Zr-0.1C alloy was subjected to cold rolling (10% and 20%) prior to irradiation. Characterization of the irradiated material was performed using detailed X-ray diffraction line profile analyses. The results, in terms of various microstructural parameters, revealed that the preexisting defects in these materials act as effective sinks for irradiation-induced defects during the initial irradiation. Electron backscattered diffraction analyses suggested that the fraction of low-angle grain boundaries plays a crucial role during irradiation and determines the outcome of the competing process between defect annihilation and defect production. Transmission electron microscopy analyses revealed that the black dots were coexisting with carbide precipitate in the cold-rolled Nb alloy after irradiation. Mechanical properties were also assessed by measuring microhardness to correlate with the changes in microstructure after irradiation. The change in yield strength calculated from the change in the microhardness value as a function of dose indicated that the 20% cold-rolled Nb alloy was more radiation resistant than the 10% cold-rolled Nb alloy at ambient temperature and a low-dose regime. To see the effect of alloying elements, similar studies were also carried out on pure Nb, and the results were compared with that of the Nb-1Zr-0.1C alloy. All these findings help to develop a better understanding of the behavior of the proton-irradiated Nb-1Zr-0.1C alloy in the presence of preexisting defects introduced by means of cold rolling prior to irradiation.

The capacity of disclinated non-equilibrium GBs to accommodate point defects in tungsten

Experimental observations reveal that disclinated non-equilibrium grain boundaries (GBs) exist extensively in polycrystalline metals; however, the interaction between these GBs and irradiation-induced point defects is rarely reported. In the present work, Molecular statics simulation was used to evaluate the capacity of disclinated non-equilibrium GBs to accommodate point defects in tungsten. Simulation results showed that disclinated non-equilibrium GBs are more efficient sinks for point defects than their equilibrium counterparts because of long-range stress field around them. Continuous segregation of point defects will change the structure of the disclinated non-equilibrium GBs and leads to the GBs tending to relax to equilibrium state. According to theoretical calculation, the disclinated non-equilibrium GBs can absorb a large number of point defects before transforming into equilibrium state; therefore, disclinated non-equilibrium GBs have very strong capacity to accommodate point defects and can be used as strong defect sinks for developing radiation resistance materials.

A potential mechanism for abnormal grain growth in Ni thin films on c-sapphire

Normal grain growth (NGG) of a (111) textured Ni film on c-sapphire and abnormal grain growth (AGG) of (100) grains at the expense of this (111) texture has been studied as a function of temperature with and without a capping layer. The grain boundaries (GBs) in the Ni film are controlled by the preferred orientation relationships (ORs) adopted by the Ni grains on the sapphire substrate. The 2 variants of a single OR, Ni(111)<11¯0>//Al2O3(0001)<11¯00>, form a (111) mazed bicrystal with Σ3 GBs. The (100) grains have a single OR, Ni(100)<010>//Al2O3(0001)<11¯00> with 3 variants; their GBs within the (111) grains have the (111)<11¯0>//(100)<010> misorientation.(100) AGG within the (111) mazed bicrystal of the 100 nm Ni film takes place above 1023 K. The orientation transition is driven by the biaxial elastic modulus anisotropy which favors the growth of (100) grains over (111) grains, as this reduces the elastic strain energy induced by the thermal mismatch between Ni and sapphire. (100) AGG is suppressed and the NGG of the (111) texture is slowed down when the film is covered by a 10 nm amorphous alumina layer aimed at inhibiting surface diffusion. Thus, it is proposed that as long as the surface can act as a sink for the point defects diffusing along the GBs, the movement of the GBs is correlated to the diffusivity of atoms and vacancies, which is a function of their misorientation and crystallographic GB structure.

Radiation Effects in Tungsten and Tungsten-Copper Alloys Treated with Compression Plasma Flows and Irradiated with He Ions.

The paper presents the results of studying the structure and phase state of tungsten and tungsten-copper alloy after pulsed action of compression plasma flows and irradiation with helium ions. The compression plasma flows were used to modify the surface layer of tungsten, as well as to create an alloy based on tungsten and copper. Using scanning electron microscopy and X-ray structural analysis, the formation of radiation defects on the tungsten surface was detected in the form of local areas of exfoliation and destruction, which begin to form at helium ion irradiation doses of 2 × 1017 cm-2. It is shown that preliminary plasma treatment of the surface in the melting mode leads to the complete disappearance of surface radiation defects up to a dose of 2 × 1017 cm-2, which may be associated with the formation of a fine-crystalline grain structure, the intergranular boundaries of which serve as effective sinks for primary radiation defects.

Formation of uranium nitride nanoparticles via mechanical alloying of uranium-molybdenum alloy fuels in gaseous nitrogen

Uranium-molybdenum (U-Mo) alloys show promise as a nuclear fuel system due to their high thermal conductivity and fuel loading capability. However, U-Mo systems are susceptible to irradiation induced swelling ultimately affecting the cladding via mechanical and chemical interactions. To address these shortcomings, this research investigated the formation of uranium mononitride (UN) nanoparticles within a 90 wt% U/ 10 wt% Mo (U-10Mo) matrix to act as a prospective defect sink for fission products at nanometric hetero-interfaces. To promote the formation of UN, U-10Mo powders were mechanically alloyed under a high purity nitrogen atmosphere. Variations of the milling process investigated included media size, duration of milling, and number of times the milling jar was re-aerated with nitrogen gas. Characterization of the fuel microstructure was completed using light element analysis, X-ray diffraction, scanning and transmission-electron microscopy, electron energy loss spectroscopy, and atom probe tomography. UN nanoparticles measuring 1–5 nm in radius were observed in the U-Mo matrix as early as 1 h into the mechanical alloying process. Milling time in excess of 10 h was found to lead to deleterious effects induced by the stainless-steel milling media.

Effect of pre-existing dislocations and precipitates on microstructure evolution in W-0.5ZrC alloy during in-situ He+ & H2+ dual-beam synergistic irradiation

Bubble evolution in different defect sinks has a crucial effect on the properties of materials for fusion reactor. Here, transmission electron microscopy (TEM) was used to in-situ investigate bubble evolution in different defect sinks, including pre-existing dislocations and precipitates in W-0.5ZrC alloy during He+ & H2+ dual-beam synergistic irradiation at 900 °C. The average size and volume number density of bubbles with increasing irradiation dose were obtained, revealing not only the significant effect of different types of defect sinks, but also the obvious impact of the same type of defect sinks in different states. It was found that linear pre-existing dislocations promoted bubble growth while inhibited nucleation. Pre-existing dislocations characterized by curved form exhibited strong inhibition effect on bubble nucleation, while pre-existing dislocations with staggered-mesh form, not only inhibited bubble growth but also suppressed nucleation compare to the linear pre-existing dislocations. ZrC precipitates had minimal effect on bubble growth, but could decrease the bubble nucleation. Dispersed small precipitates could better reduce the bubble nucleation sites in the early stage of irradiation. Once the absorption capacity of small precipitate was saturated, bubble growth might accelerate sharply. The presence of hydrogen made the linear pre-existing dislocation maintain high mobility. The current research results are of great significance for understanding the irradiation behavior of tungsten alloys and optimizing the preparation process.

Investigations on structure and optical properties of neutron irradiated nanocrystalline La2Zr2O7 pyrochlores

This research work investigates the response of nanocrystalline lanthanum zirconium oxide (LZO) under neutron irradiation, emphasizing grain size-dependent structural stability. LZO samples were synthesized with the chemical precipitation method and the LZO pellets were annealed at two different temperatures (1200 °C and 1400 °C) to obtain two different grain size samples, namely LZO-1200 and LZO-1400. The samples were exposed to neutrons at different doses (ranging from 1.63 × 1012 n/cm2 to 5.68 × 1014 n/cm2) and the effects were studied with X-ray diffraction, scanning electron microscopy, Raman scattering and optical characterization. The structural characterization results reveal that, the grain size and strain are significantly modified in LZO-1400 samples, and such changes are absent inLZO-1200 samples. The band gap and photoluminescence intensity are reduced significantly in LZO-1400 samples, however, this effect is not observed in LZO-1200 samples. The experimental results reveal that in LZO-1200 samples, the grain boundaries act as a sink for irradiation-induced defects and there is no significant change in the grain size, strain and band gap. However, in the case of the LZO-1400 sample, the grain size is larger than the cascades induced by primary knock-on atoms (PKAs) and it results in disorder-driven fast grain growth and changes in band gap and PL intensity. The LZO-1200 samples with small grain sizes are radiation resistant.

Unique irradiation damage behavior and deformation mechanisms in crystalline/amorphous Ag/Cu-Zr nanolaminates

Nanolaminated materials due to the abundant interfaces that are efficient sinks for irradiation defects have received broad attention. In this work, Ag/Cu-Zr crystalline/amorphous nanolaminates (C/ANLs) with a wide modulation ratio η from 0.1 to 9.0 were prepared using sputtering magnetron and implanted to a total influence of 1×1017 ions cm-2 He ions to investigate their irradiation damage behavior and mechanical properties. The irradiated Ag/Cu-Zr C/ANLs enable an interesting distribution behavior of He bubbles that only aggregate in the crystalline Ag layers but are absent in the amorphous Cu-Zr layers and interfaces, and become increasingly obvious with reducing η. This is attributed to the synergistic effect of C/A interfaces and amorphous layers as efficient sinks for irradiation defects. The irradiation-induced hardening is found to go through a minimum at η = 3.0, owing to the transformation of strengthening mechanism from He bubbles strengthening in Ag layers as η ≤ 3.0 to free volume annihilation in Cu-Zr layers as η > 3.0. Moreover, the strain rate sensitivity (SRS) m of irradiated C/ANLs firstly decreases rapidly from positive to negative and then increases slowly with raising η, differing from the monotonic η-dependent SRS m of as-deposited C/ANLs. Relevant deformation mechanisms are elucidated based on the dislocation-bubble interactions in terms of pinning effect on dislocations bow-out and dynamic strain aging effect. Our findings could provide valuable insights for the microstructural design of C/ANLs for applications in nuclear reactors.

A mechanistic study of grain boundary behaviour during irradiation-induced growth in zirconium

Grain boundaries play a significant role during deformation and can exhibit both beneficial and adverse effects on deformation behaviour during irradiation. For instance, macroscopic irradiation growth is increased by the presence of a high density of grain boundaries. Grain boundaries can act as sinks for point defects and barriers to dislocation motion. Thus, a mechanistic assessment of the impact of grain boundaries is essential for better understanding irradiation-induced deformation. The interplay between irradiation-induced microstructure and mechanical properties has been an area of research for many decades, but many uncertainties remain. In the present work, the deformation fields adjacent to grain boundaries in pure zirconium are investigated by using high angular resolution electron backscatter diffraction (HR-EBSD) after a period of irradiation growth. It is observed that the extent of grain boundary deformation is associated with the crystallographic orientation difference between adjacent grains. High-angle grain boundaries function as a deformation constraint, whereas a significant grain boundary migration occurs at some low-angle grain boundaries, which likely act as an active diffusion path for the irradiation-induced point defects. In addition, the change in residual strain fields associated with the irradiation damage is quantified with respect to a non-irradiated reference. A net change in the state of residual elastic strain in the order of ∼10−3 has been estimated for ∼8 dpa proton irradiation. Furthermore, the origin of residual elastic strain and lattice misorientation concentration along (0002) traces are discussed.

Calculation of grain boundary diffusion coefficients in γU-Mo using atomistic simulations

The γU-Mo alloy has been selected for the conversion of U.S. High-Performance Research Reactor (HPRR) fuel from highly enriched uranium to low enriched uranium as a part of the effort to reduce nuclear proliferation risks. Although γU-Mo alloys have the advantage of high uranium density and good overall irradiation performance, the irradiation-induced swelling and creep in the metal remain important design parameters since they influence the mechanical and thermal integrity of the fuel. To account for these design criteria, engineering scale models need fundamental properties, such as diffusion coefficients, as input. In this study, the diffusion of related species along grain boundaries in γU-Mo fuel is quantified considering that grain boundaries act as sinks for point defects, nucleation sites for gas bubbles, avenues for Coble creep, etc. The diffusivities of U and Mo in selected grain boundaries of γU-Mo alloys (γU-7Mo, γU-10Mo, and γU-12Mo) are obtained utilizing molecular dynamics simulations for a temperature range of 600 K - 1200 K with an interval of 100 K. The structures analyzed include symmetric tilt, asymmetric tilt, and twist grain boundaries. The grain boundary diffusion coefficients of U and Mo in the examined γU-Mo alloys are on the order of 10−14 to 10−11 m2s−1. It is observed that the U diffusivity in the grain boundary is higher than the Mo diffusivity in all cases and that the increase in Mo content of the alloy correlates to a decrease in the grain boundary diffusion. Xe diffusion along γU-10Mo grain boundaries is also calculated in this work, and the diffusivity of Xe in the γU-10Mo grain boundaries is found to be 8 to 15 orders of magnitude higher than the intrinsic Xe diffusivity in γU-10Mo depending on the temperature. The information gathered in this work can inform fuel swelling and creep models and help understand various other phenomena related to γU-Mo fuel performance.

Complex dislocation loop networks as natural extensions of the sink efficiency of saturated grain boundaries in irradiated metals.

The development of radiation-tolerant structural materials is an essential element for the success of advanced nuclear energy concepts. A proven strategy to increase radiation resistance is to create microstructures with a high density of internal defect sinks, such as grain boundaries (GBs). However, as GBs absorb defects, they undergo internal transformations that limit their ability to capture defects indefinitely. Here, we show that, as the sink efficiency of GBs becomes exhausted with increasing irradiation dose, networks of irradiation loops form in the vicinity of saturated or near-saturated GB, maintaining and even increasing their capacity to continue absorbing defects. The formation of these networks fundamentally changes the driving force for defect absorption at GB, from "chemical" to "elastic." Using thermally-activated dislocation dynamics simulations, we show that these networks are consistent with experimental measurements of defect densities near GB. Our results point to these networks as a natural continuation of the GB once they exhaust their internal defect absorption capacity.

The radiation resistance and chemical stability optimization of analog actinide nuclides incorporated Gd2Zr2O7 ceramics via grain size control

Gd2Zr2O7 is considered a promising host material for immobilizing high-level radioactive waste. However, its interplay between grain sizes, radiation resistance and chemical stability is not well understood yet. Herein, Gd1.7Nd0.3Zr0.5Ce1.5O7 ceramics with defect-fluorite phase and three different grain sizes (96, 390 and 1959 nm) were successfully fabricated and irradiated by 140 keV He ion beam up to a fluence of 8 × 1017 ions/cm2. Phase evolutions, microstructure changes and chemical stabilities were investigated. The nano-grained ceramic exhibits the lowest degree of amorphization and a delayed irradiation-induced lattice expansion and damage evolution process, demonstrating an enhanced ability in suppressing the coalescence of He bubbles inside the grain. Grain boundaries parallel to the surface were found to be more favorable sinks for irradiation-induced defects. Chemical stability tests showed that the 42 days leaching rates of all elements were between 10−7∼10−5 g/m2·d. All samples have maintained this superior chemical stability even after irradiation. Interestingly, the leaching test results also suggested that grain size may be not the primary factor influencing the long-term leaching rate.

Phase-field simulation of the effect of grain boundary on fission gas migration in UO2 fuel

Grain boundaries are widely recognized as line defect sinks. During reactor operation, vacancies and fission gas atoms within UO2 fuel grains migrate to the grain boundaries, forming bubbles at these locations. In order to better understand the effect of grain boundaries on the migration of fission gas atoms, this study employed a phase-field model to simulate the nucleation and growth process of fission gas bubbles within the UO2 fuel grains and in the vicinity of grain boundaries. This study also investigated the degree to which grain boundaries affect fission gas atoms under different temperature conditions. The results indicated that in models containing grain boundaries, the nucleation of fission gas bubbles occurred earlier, as compared to models without grain boundaries. A noticeable bubble denuded zone was also observed adjacent to grain boundary interfaces. Furthermore, with increasing temperature, the bubble denuded zone becomes thicker. The effects of irradiation, vacancy diffusion, Ostwald ripening, as well as grain boundary trapping were discussed.

Uncovering origin of grain boundary resistance to irradiation damage in NiCoCr multi-principal element alloys

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.

Effects of radiation damage on the yielding and fracture of nanowires.

Since free surfaces act as perfect sinks for radiation-induced defects, nanowires, owing to their high surface-to-volume ratio, are considered to be radiation tolerant. But the question remains on how this tolerance translates to their yielding and fracture behavior. Atomistic simulations of irradiated gold nanowires reported here show the existence of a size regime where the yield stress is affected by the accumulation of radiation damage. Our analysis also shows that, regardless of the diameter of the nanowire, early on during tensile loading, much of the radiation-induced defect content initially present in the wire is rapidly cleaned by the motion of pre-existing dislocations as well as dislocations emitted from the surface of the wire. This defect removal process resets the crystallographic configuration of the wire which subsequently deforms and fractures via the same mechanisms that occur in pristine, un-irradiated nanowires. We conclude that the fracture behavior of nanowires in the size and dose regimes tested is unaffected by radiation damage.

A general scheme for point defect sink strength calculation and related machine-learning-based expressions

Irradiation tends to increase the concentration of point defects (PDs) in crystalline materials, whose consecutive interactions with other types of defects, such as dislocation and void, are recognised highly responsible for the characteristic plastic and damaging behaviours of materials under irradiation. Conventional treatments on evaluating the strength of PD sinks see their limitation with strong regularity requirements over the models used for summarising the key underlying microstructural behaviours, where analytical solutions are bound to be the outcome. The present article serves to introduce a general scheme for PD sink strength evaluation, where constraints on solution analyticity are fully resolved with the use of machine learning. In particular, a neural network representation of the PD sink strength due to void/bubble is derived, where PD transportation tendencies against the hydrostatic pressure gradient surrounding a bubble can be considered in details. The treatment is also applied to analyse PD sink strength due to edge dislocation clusters. For nearly uniformly distributed clusters, upon undertaking a two-scale asymptotic strategy, the corresponding sink strength formulation becomes explicit. For randomly distributed dislocations, the sink strength is found to roughly scale with the onsite dislocation density. But for patterned dislocations, such as dislocation dipoles, their sink strength is suggested to vary with the applied load. The machine-learning-based formulation is also compared well with the results obtained by other multiscale methods.

Interaction of displacement cascades with {10 [formula omitted]2} and {10 [formula omitted]1} twin boundaries in zirconium: A molecular dynamic study

Twinning is a significant deformation mode in zirconium (Zr). Here we investigated the primary displacement cascades near the twin boundary (TB) and mechanical properties in Zr using molecular dynamics simulations. The results show that {10 1‾2} and {10 1‾1} TBs can act as an effective sink for point defects to enhance irradiation resistance in Zr. The {10 1‾2} TB exhibits better sink efficiency than {10 1‾1} TB. Hence, it's recommended to employ Zr alloys with more {10 1‾2} deformation twinning microstructure in irradiation conditions. The present work predicts the introduction of TBs may contribute to a decrease in the density of large-sized clusters and clustered fraction. Further, irradiation leads to localized deformation of the two TBs even in one single collision cascade. The tensile loading of the unirradiated and irradiated samples indicates that the stacking faults serve as an important deformation mechanism in two TB models. The results not only present a reference for understanding the TB structure in the displacement cascades of Zr and its alloys but also provide further support for the design of irradiation-resistant nanotwinned metals.

The effect of screw dislocation on Helium Bubble growth in Tungsten: molecular dynamic simulation study

The effect of a 1/2[111] screw dislocation on the growth of helium bubble in tungsten is investigated by molecular dynamics simulation method. When helium bubble grows at sites less than 3 nm away from the dislocation, tungsten self-interstitial atoms (SIAs) pushed out by the growing bubble are absorbed by and then migrate rapidly along the dislocation line, inducing the screw dislocation to evolve into a helical configuration. The existence of screw dislocation can facilitate the growth of helium bubble. The interaction energies between the screw dislocation and point defects (helium, SIA and vacancy) are calculated. The interaction energies are all negative, indicating that the screw dislocation can attract and serve as a sink for these point defects. Further investigation shows that the interactions between the screw dislocation and these point defects are local, and stem from the special stress distribution pattern of the screw dislocation.

Effect of He-ions irradiation on the microstructure and mechanical properties of TiTaNbZr refractory medium-entropy alloy film

In this work, TiTaNbZr refractory medium-entropy alloy (RMEA) films were irradiated by 60 keV helium (He) ions at the fluence of 5 × 1016 cm−2 - 2 × 1017 cm−2. After irradiation, a structural transformation of amorphous-BCC single phase, namely self-healing behavior, in TiTaNbZr RMEA films was revealed on account of the enhanced diffusion and redistribution of atoms, which could also lead to a decrease of defects. In this process, increased grain boundaries, acting as effective defect sinks, have inhibited the He bubbles within crystals. The short-range migration of atoms did not destroy the uniformity of elements dramatically. In addition, the correlation between the structural transformation and mechanical properties evolution in the RMEA films was also further clarified. No degradation of mechanical properties even for irradiation fluence as high as 2 × 1017 cm−2 due to the increased grain boundaries and self-healing mechanism. These results all indicate that the TiTaNbZr RMEA have outstanding irradiation resistance and potential applications in reactor structural materials. Meanwhile, the microstructural transformation and performance evolution of the TiTaNbZr RMEA explored in this study could provide more understanding of irradiation effect and a new perspective for the development of the alloys with self-healing ability in refractory high/medium entropy alloys (RHEAs/MEAs).