Evolution Of Dislocation Loops Research Articles

Effect of hydrogen, helium and dose rate on the evolution of dislocation loop in pure iron during in-situ ion irradiation

Simulating neutron irradiation with ion irradiation requires considering the comprehensive effects of hydrogen, helium and dose rate to fully understand the microstructure evolution of material in an irradiation environment. Here, the characterization and evolution of dislocation loops in pure iron were investigated under heavy ion irradiation with four different injected gas ion types (Fe+, Fe++H2+, Fe++He+and Fe++H2++He+) and under dual-beam (Fe++He+) irradiation with four different dose rates using in-situ transmission electron microscopy at 723 K. Comparing to the single-beam (Fe+) irradiation, the additional injection of H or He, or co-injected H and He, mainly promoted the nucleation of dislocation loops at the early stage of irradiation and enhanced loop growth after that. Among them, the simultaneous irradiation of triple beams (Fe++H2++He+) was most pronounced. Moreover, triple-beam (Fe++H2++He+) irradiation resulted in the highest ratio of the rotation of loop habit-plane. Under dual-beam (Fe++He+) irradiation, the results demonstrated that as the dose rate increased, the loop density increased and the loop size decreased. Moreover, the fraction of 〈100〉 loops at low dose rate was noticeably greater than that at high dose rate. The results of the current work provide important references for a better understanding of the effectiveness of ion irradiation in simulating neutron irradiation.

Fluctuations in misfit volume by interstitial carbon atoms contribute to the unusual dislocation loop evolution in high-entropy alloys under irradiation

Dislocation loops play a pivotal role for the irradiation behavior and associated mechanical properties of materials. This study employs advanced transmission electron microscopy, nanoindentation and first-principles calculations to elucidate the impact of interstitial carbon atoms on the evolution of dislocation loops in an Fe49.5Mn30Co10Cr10C0.5 high-entropy alloy under ion irradiation at temperatures ranging from 420°C to 540°C with a peak dose of 50 DPA. Interstitial carbon atoms have been revealed to broaden the size distribution of dislocation loops, rendering their morphologies more jagged, and consequently inducing a more substantial pinning effect. These outcomes can be attributed to the enhancement of misfit volume fluctuations of the alloying elements, which drives more severe local lattice distortions and larger variation of local chemical environments. As a result, these fluctuations promote the local pinning of dislocation loops, lead to their rugged movement and contribute to the rise in hardening. Our findings bring previously unrecognized insights into the role of interstitial carbon atoms in influencing the nature of dislocation loops in high-entropy alloys, and provide important clues to the design of alloys for nuclear applications.

MICROSTRUCTURAL EVOLUTION OF ION IRRADIATED COMMERCIALLY PURE TITANIUM

Radiation-induced dislocation loop evolution in ion irradiated commercially pure (CP) Ti was investigated through both ex situ 4 MeV Ar beams and by in situ TEM irradiation with 1 MeV Kr ion beams at different temperatures. The dominant mechanism for dislocation loop growth was loop coalescence. Both <a>- and c-component loops exhibited an increase in size and density with increased irradiation dose and temperature. C-component loops were observed only after reaching a threshold incubation dose (TID). In CP Ti, these loops initiated at lower doses compared to previous observations in Zr, and the TID decreased with increasing temperature.

Extending damage accumulation of commercial reactor irradiated 316 stainless steel with ion irradiation

Austenitic 316 stainless steel flux thimble tubes (FTTs) removed from a commercial pressurized water reactor (PWR) were re-irradiated using nickel ions to evolve the irradiated microstructure to higher damage levels than those achieved in reactor. The microstructures of the neutron-plus-ion irradiated samples were compared with those of neutron irradiated only samples at the same doses to evaluate the effectiveness of ion irradiation to extend the damage range and match that created in reactor. Ion irradiations effectively captured, both qualitatively and semi-quantitatively, the evolution of dislocation loops, nanocavities, nanocluster size and density, and radiation-induced segregation (RIS) with dose. Only Ni-Si clusters were observed in ion irradiated FTT samples while Ni-Si-Mn clusters that were frequently observed in reactor irradiated samples above ∼41 dpa were absent in ion irradiated samples probably due to ballistic mixing at the high ion irradiation damage rate. Two samples were ion irradiated to ∼160 dpa to predict the irradiated microstructure that may develop due to an increase of the damage accumulation by extending a PWR life from 40 years to 60 years. The extended ion irradiation results indicated that significant microstructure changes were unlikely to occur even to very high doses near common PWR relevant temperature conditions. Overall, the microstructure resulting from ion irradiation following neutron irradiation matched that of neutron to the same dose reasonably well in most cases.

A spatially resolved analysis of dislocation loop and nanohardness evolution in proton irradiated Zircaloys

Proton irradiation is increasingly used as a surrogate for neutron irradiation, providing similar damage structures at significantly lower costs and less time compared to neutron irradiation. However, in contrast to neutrons, protons produce a depth dependent damage profile that incorporates a plateau region and a Bragg peak in the tens of microns region depending on the material and proton energy. Here we demonstrate that this depth dependent damage can be utilised to obtain new understanding about irradiation induced damage at different dose levels from a single sample by combining spatially resolved micro-beam synchrotron X-ray diffraction-based dislocation analysis and nanoindentation. For this purpose, and to study the damage evolution starting from very early stages, we have investigated microstructure evolution and hardening behaviour of Zircaloy-2 and Zircaloy-4 proton irradiated samples between 0.01 dpa and 17 dpa at 350 °C. The results highlight good agreements of SRIM-based damage depth predictions with irradiation-induced dislocation appearance and nanohardness. However, at even relatively low damage levels within the plateau region, dislocation line density and nanohardness profiles appear saturated across the entire irradiated region, even when the dpa levels differed significantly. When relating dislocation loop line densities to nanohardness, it was found that from a certain point hardness increased only very slightly with increasing line densities. This apparent discrepancy can be explained by a decreasing loop size for the highest line densities identified by from the diffraction line profile analysis and the application of a Dispersed Barrier Hardening model that relates obstacle strength to dislocation loop size.

Evolution of dislocation loops in irradiated α-Uranium: An atomistically-informed cluster dynamics investigation

An atomistically-informed mean field cluster dynamics model has been developed to investigate the nucleation and growth of prismatic loops in irradiated α-Uranium. TEM analysis of neutron irradiated α-Uranium shows the evolution of self-interstitial atom and vacancy loops on (010) and (100) crystallographic planes, respectively, resulting in an anisotropic lattice swelling of its face-centered orthorhombic crystal. To provide model parameters, the crystallography of loops and the binding energy of point defects to these loops were studied using an angular dependent EAM potential and classical molecular dynamics (MD) simulations. Furthermore, using the bond-boost hyperdynamics method, the anisotropic diffusion of interstitials and vacancies in α-Uranium was investigated. The mechanisms of point defect diffusion and the associated migration energies were reported and compared with previous DFT studies. The energetics and kinetic quantities mentioned above were used in the cluster dynamics model to predict the cluster density as a function of dose rate, dose and temperature and the results were compared to the reported neutron irradiation experiments. The model predictions reveal an accumulation of small sized vacancy loops along with a population of large and growing self-interstitial loops, which closely corresponds to the TEM observations.

The design of Pd-containing high-entropy alloys and their hardening behavior under He ion irradiation

Imparting additional chemical heterogeneities of high-entropy alloys (HEAs) is considered as a novel strategy to improve their irradiation resistance on dislocation loop growth; however, there is still a lack of knowledge about the effects of this strategy on bubble evolution and irradiation-induced hardening. In this study, a series of Pd-containing FeCrNiCo HEAs with proven potential for chemical short-range order were designed and irradiated by 400 keV He ions to the fluences of 5 × 1016 cm−2, 1 × 1017 cm−2, and 5 × 1017 cm−2 at 723 K. Microscopic characterization demonstrates a remarkably inhibited growth of dislocation loops and promoted bubble coarsening in the Pd-containing HEAs compared to the HEA without Pd alloying. The underlying mechanism is that although the alloying of Pd can suppress the evolution of dislocation loops through severe local lattice distortion, it also significantly reduces the vacancy formation energy of the entire HEA system, leading to the encouraged bubble evolution. Pd-containing HEAs exhibit improved resistance to irradiation-induced hardening at low doses, but this advantage is unexpectedly absent with increasing dose. The hardening mechanism is explained by identifying the loop-dominated hardening at low dose and the bubble-induced hardening which makes an increased contribution at high dose. In addition, the Ni segregation around bubbles and loop punching mechanism are also discussed to provide essential evidence. These insights can help to shed light on the effects of Pd alloying on bubble evolution and irradiation-induced hardening behavior in HEAs, and thus contribute to the design of irradiation-resistant HEAs.

Evolution of dislocation loops and voids in post-irradiation annealed ThO2: A combined in-situ TEM and cluster dynamics investigation

The effect of isochronal annealing on the evolution of dislocation loop and void population in proton irradiated ThO2 has been investigated. Post-irradiation annealing in other actinide oxides like UO2 shows significant loop coarsening. ThO2 samples were irradiated with 2 MeV protons up to a dose of 0.1 dpa at 600 °C. Post-irradiation isochronal annealing was performed at 600, 800, 1000 and 1100 °C for 1 h at each temperature using in-situ TEM. Only faulted 1/3<111> type dislocation loops were observed, and their sizes and distribution were characterized. The population of self-interstitial atom (SIA) dislocation loops did not show any significant growth and coarsening. Additionally, nanometric voids were observed at annealing temperatures of 1000 and 1100 °C. Using cluster dynamics (CD), we have studied the nucleation and growth of point defects and defect clusters, i.e., SIA prismatic dislocation loops and nanometric and sub-nanometric voids in proton irradiated ThO2. The CD model was further utilized to predict the growth and coarsening of loops and voids during isochronal annealing at the experimental and higher temperatures. The model did not predict significant SIA loop growth which closely corresponds to the TEM observations. CD predicted SIA loop coarsening is insignificant even at high annealing temperature of 1500 °C because the model only considers the growth of defect clusters by absorption of like point defects, i.e., SIA loops absorb interstitials and voids absorb vacancies, and cannot account for their migration and coalescence due to elastic interaction. The CD model also predicts the evolution of nanometric voids having mean size within the error bounds of TEM observations.

Helium role in Fe9Cr1.5W0.4Si F/M steel during Fe++He+ dual-beam irradiation

In materials exposed to neutron irradiation, helium atoms produced by transmutation reactions are prevalent and play a significant role in the evolution of dislocation loops, causing mechanical property degradation. However, systematic investigation on the helium role is lack and possible origin of helium effects is unclear. Here, by performing 400 keV Fe ion and 30 keV He in-situ irradiation, this study found that about 9.2×1012 cm−3 loops nucleated for every increase of 1 appm He at 0.9 dpa in Fe++He+ dual beam irradiation and 87% loop density was produced with the help of helium at 2262 appm/dpa. The different loop evolution in single He irradiation and Fe++He+ irradiation indicated the presence of high-density TEM-invisible (≤ 2 nm) small loops after Fe+ irradiation at high temperatures. The large number of loops produced by helium irradiation were possibly related with the small bubbles.

Effect of ion flux on one-dimensional migration of dislocation loops in Fe9Cr1.5W0.4Si F/M steel during in-situ Fe+ irradiation

The study of one-dimensional (1D) migration of dislocation loops is essential for understanding the evolution of material microstructure in the irradiation environment. The 1D migration of dislocation loops under thermal activation and electron irradiation has been widely studied, but the effect of ion flux on the 1D migration of dislocation loops has not been concerned. In this study, the 1D migration and evolution of dislocation loops in the Fe9Cr1.5W0.4Si ferrite/martensite steel during 400 keV Fe+ irradiation was investigated by in-situ transmission electron microscopy at room temperature. Compared with 1/2<111> dislocation loops, 〈100〉 dislocation loops had weak migration ability and low migration frequency. The effect of ion flux on the 1D migration of dislocation loops were explored, which showed that increasing the ion flux was not only beneficial to the migration of large dislocation loops, but also increased the migration frequency. In addition, the higher the ion flux, the higher the density of mobile dislocation loops per unit irradiation dose. The knowledge gained here provides an important reference for understanding and predicting the degradation process of irradiated materials.

Pressure Effect on Diffusion of Native Defects and Mg Impurity in Mg‐Ion‐Implanted GaN during Ultrahigh‐Pressure Annealing

Herein, transmission electron microscopy direct observations are used to examine the time evolution of dislocation loops in Mg‐ion‐implanted GaN during annealing in N2 atmospheres of 2.0 and 0.3 GPa. It is indicated in the results that the diffusion of the native defects, for both interstitials and vacancies, is retarded during annealing at the higher pressure. Furthermore, secondary ion mass spectrometry shows that annealing at the higher pressure retards the migration of Mg and increases Mg‐acceptor concentration in the ion‐implanted region. These findings provide a design principle of the postimplantation annealing process to activate ion‐implanted Mg in GaN.

Molecular dynamics simulations of displacement cascades in vanadium: Generation and types of dislocation loops

Vanadium(V)-based alloys have received extensive attention as promising first-wall and blanket candidates for fusion reactors. In the present study, molecular dynamics (MD) simulations have been performed to study the displacement cascades in V bulk with the recoil energy up to 80 keV at 300 K. The results indicate that the clustering fraction and cluster size of vacancies are higher than those of interstitials. The nucleation, evolution, and energy of dislocation loops are studied. It’s found that 1/2 〈1 1 1〉 interstitial and vacancy dislocation loops are the primary ones, while low concentration 〈1 0 0〉 vacancy loops are also formed. In addition, the nucleation of 1/2 〈1 1 1〉 interstitial dislocation loops is earlier than that of 1/2 〈1 1 1〉 vacancy ones. The reason for the appearance of 1/2 〈1 1 1〉 vacancy dislocation loops is that the formation energy of the 1/2 〈1 1 1〉 vacancy dislocation loops is close to that of voids containing the same number of vacancies. These results provide a new understanding of applying V-based alloys in fusion reactors.

In-situ TEM investigation on the evolution of He bubbles and dislocation loops in Ni-based alloy irradiated by H2+ &amp; He+ dual-beam ions

• Synergistic evolution mechanism of He bubbles and loops in Ni-based alloy irradiated by dual-beam ions were investigated via in-situ TEM. • Several evolution stages of the loop-punching mechanism and its effect on evolution of neighboring bubbles and loops were revealed. • The details of the unfaulting processes of the Frank loops and the formation mechanism of rhombic loops were discussed. • Four Frank loop variations were identified, and their different evolution behaviors were studied. The synergistic evolution mechanisms of He bubbles and dislocation loops under 30 keV H 2 + & H e + dual-beam ions irradiation at 650°C in the Ni-based alloy GH3535, which is the most promising candidate structure material for molten salt reactors (MSRs), were revealed via in-situ TEM. The nucleation, merging, and change in the size of the dislocation loops and He bubbles were characterized in detail to study the influences of irradiation fluence and pre-existing dislocation loops on their evolutions. The number density of both the He bubbles and dislocation loops increases rapidly and subsequently saturates, whereas their size continuously increases with the increasing ion fluence. Pre-existing dislocation loops with strong absorption characteristics grow preferentially and suppress the nucleation of dislocation loops during the dual-beam ions irradiation. Moreover, the bubbles tend to nucleate within the dislocation loops to form bubble-loop complexes, and then decrease in their number density. The details of the unfaulting processes of the Frank loops were discussed, where the energy difference between the two types of loops as well as the evolution of the inside Shockley dislocation loops dominates the unfaulting behavior. The several evolution stages of the loop-punching mechanism are revealed, and the emitted loops can directly form perfect loops as well as unfault the neighboring Frank loops. The He bubbles inside the loops provide corresponding stress for the formation of rhombic loops, which can achieve rapid growth and sweep ability by merging with the neighboring loops. Additionally, its dissociation to Shockley dislocation can unfault the Frank loops along their slip direction. Among the four Frank loop variations, the edge-on Frank loop variations have the highest growth rate, followed by the perfect loop. The related mechanisms based on in-situ experimental observation are discussed in depth. Synergistic evolution mechanism of He bubbles and dislocation loops, including loop-punching mechanism and bubble-loop complexes, in Ni-based alloy irradiated by dual-beam ions were investigated via in-situ TEM.

In-situ TEM investigation of dislocation loop evolution in HR3 steel during Fe+ ion irradiation

The evolution and characteristics of dislocation loops, which are affected by the value of temperature play a crucial role in the mechanical properties of structural materials. This study explores the evolution processes of the dislocation loops with the increasing values of the temperature, and the effect of precipitate on the dislocation loop. In-situ TEM experimental observations are performed using HR3 steel, which is commonly used as a structural material in nuclear reactors, during 400 keV Fe+ irradiation at room temperature (RT), 300 °C, 400 °C and 500 °C. The evolution of the dislocation loop at different values of the temperature is characterized in detail. The average size and density of dislocation loops as a function of displacement per atom (dpa) and values of temperature were constructed and analyzed comparatively. Irradiated samples at higher temperatures have a larger average size and a lower density of dislocation loops. The pre-existing dislocation lines have a profound effect on the density and Burgers vector of dislocation loops. Our present results provide a fundamental insight into the evolution and characteristics of dislocation loops in HR3 steel in a gradually changing of temperature environment.

Effect of dose rate on the characteristics of dislocation loops in palladium: In-situ TEM analysis during 30 keV H2+ irradiation

Ion irradiation is usually used to simulate neutron irradiation to accelerate the evaluation of the irradiation behavior of reactor materials. However, the validity of using a high damage rate of ion irradiation to simulate a low damage rate of neutron irradiation has always been a controversial topic. Here, the effect of two dose rates (2.94 × 10–6 and 7.35 × 10–5 dpa s–1) on the characteristics and evolution of dislocation loops in palladium was studied in situ during 30 keV H2+ irradiation using transmission electron microscopy. The dose rate obviously affected the nucleation rate and growth rate of dislocation loops, the types (Frank loops or perfect loops) of dislocation loops, and the irradiation hardening and total damage obtained from the product of average loop size and loop density. At the same irradiation dose, a high dose rate would lead to high loop density, small average loop size, low loop growth rate, and low irradiation hardening and damage induced by loops in pure Pd. Meanwhile, it was found for the first time that a high dose rate was beneficial to the generation of perfect dislocation loops. The effect of dose rate was attributed to the different dynamic equilibrium results between the effective generation rate of point defects and their absorption rate by existing sinks. The present results show that the effect of dose rate should be considered when using ion irradiation to simulate neutron irradiation to evaluate the irradiation damage to materials.

Deterioration of irradiation resistance of ODS-F/M steel under high concentration of helium

Simultaneous Fe+He ion irradiations of China Low Activation Martensitic (CLAM) steel and oxide-dispersion-strengthened ferritic/martensitic (ODS-F/M) steel were carried out at 350 °C-550 °C to the damage dose of 11 dpa with different He concentrations to investigate the effect of oxide particles on the evolution of He bubbles and dislocation loops. The microstructures of all irradiated samples were observed by transmission electron microscopy (TEM). The average size of oxide particles decreased and the density increased after irradiation. The aggregation of He bubbles was observed at the edge of oxide particles. As the He concentration increased, the aggregation of helium bubbles appeared at the edge of more oxide particles. Furthermore, He bubbles were not observed inside ODS-F/M steel grains or at grain boundaries, which suggests that the oxide particle interface is more capable of capturing helium atoms than other sinks. The average dislocation loop size of ODS-F/M steels is only 58% of that in CLAM steel at 450 °C, showing that ODS-F/M steel has excellent inhibition on the growth of dislocation loops. Nevertheless, with the increase of temperature and He concentration, the number density and size of dislocation loops in ODS-F/M steels gradually approach those in CLAM steels. The ability of ODS-F/M steels to inhibit irradiation hardening relative to CLAM steels increases with increasing temperature and decreases with increasing He concentration. The result of first-principles calculations shows that the formation energy of interstitial Fe atom at the interface increases with the increasing of the number of He atoms, which means that the aggregation of excessive He atoms at the interface will inhibit superior irradiation resistance of ODS-F/M steel.

Effect of different pre-existing dislocation densities on the microstructure evolution of W-0.5ZrC alloy during in-situ He+ irradiation

The evolution of dislocation loops and helium bubbles in W-0.5ZrC alloy with different pre-existing dislocation densities was investigated by in-situ transmission electron microscopy during 30 keV He+ irradiation at 800 °C. The nucleation, migration, aggregation, and growth of dislocation loops and helium bubbles occurred with the whole irradiation process. The variation trends of average size and volume number density of dislocation loops and helium bubbles with the increase of irradiation dose in the regions with different pre-existing dislocation densities were obtained. It was found that the growth of dislocation loops was inhibited in the region with a low pre-existing dislocation density (about 5.2 × 1013 /m2), and the nucleation of dislocation loops was inhibited in the region with the dispersed medium density (about 1.6 × 1014 /m2, dislocations were crossed but dispersed), while both the nucleation and growth of dislocation loops were inhibited in the region with the compact medium density (about 1.5 × 1014 /m2, dislocations were crossed and compact). During the irradiation process, the pre-existing dislocations would interact with the irradiation-induced dislocation loops and helium bubbles, but their influence weakened with the increase of irradiation dose, and finally resulted in the disappearance or deformation of pre-existing dislocations. The irradiation hardening caused by dislocation loops and helium bubbles was increased with the increase of irradiation dose, and the corresponding mechanism was analyzed.

On continuum simulations of the evolution of faulted and perfect dislocation loops in silicon during post-implantation annealing

Predictive models of the damage evolution during post-implantation annealing are important for predicting the performance of final devices. An existing model of the thermal evolution of dislocation loops in silicon during post-implantation annealing has been recalibrated to better capture the distinction between faulted and perfect dislocation loops. The calibration was based on experimental findings from the literature on the simultaneous presence of both faulted and perfect dislocation loops after post-implantation annealing.Graphical abstract

In-situ TEM study on the evolution of dislocation loops and bubbles in CeO2 during Kr+ single-beam and Kr+-H2+ dual-beam synergetic irradiation

CeO2 is widely used as the nonradioactive surrogate fuel of UO2 when studying the irradiation performance of UO2. The evolution and characteristics of dislocation loops and bubbles in CeO2 foils were studied by in-situ transmission electron microscopy (TEM) observation during 400 keV Kr+ & 30 keV H2+ dual-beam synergetic irradiation and 400 keV Kr+ single-beam irradiation at 1073 K. The rotation of the habit plane of dislocation loops induced by above ion irradiation was found in the CeO2 for the first time, such as from [2¯11¯] to [3¯11¯] and then to [1¯00]. The rafted loops were first observed under Kr+ & H2+ dual-beam synergetic irradiation, which not only had similar [1¯11¯] direction, but also belonged to perfect dislocation loops (PDLs). The rafted loops were formed not only by the growth of loops that absorbed irradiation defects, but also by the combination of loops. But, this phenomenon of loop rafting was not obvious during Kr single-beam irradiation. It was first found that Kr+ irradiation induced the change of Burgers vectors of PDLs. The absorption of large PDLs to small PDLs was also observed. The average size and areal number density of dislocation loops and gas bubbles as a function of irradiation dose were constructed, which showed that the addition of H2+ obviously affected the characteristics of dislocation loops and gas bubbles and the swelling of CeO2.

Effect of dose rate on dislocation loop evolution in tungsten: Combination of defect generation rate and elastic interaction

Ion irradiation is often used to simulate neutron irradiation in reactor components, while the dose rate discrepancy makes it difficult to directly compare ion irradiation with neutron irradiation. A better understanding of dose rate effect is thus necessary to emulate neutron irradiation effect with ions. Here, based on in-situ 30 keV He+ irradiation in tungsten in TEM, we find that defect generation rate and elastic interaction are two key factors controlling dislocation loop evolution at different dose rates. By adjusting defect generation rate, the elastic interaction between the loops will play different roles. Finally, a schematical illustration is given to explain the different in loop density, size and spatial distribution caused by the combination of these two factors. Our results provide the underlying insights for understanding the dose rate effects of energetic ion irradiation.