Evolution Of Dislocation Loops Research Articles

Surface effect on <inline-formula><tex-math id="Z-20191230130848">\begin{document}${\langle 100 \rangle }$\end{document}</tex-math><alternatives><graphic xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="3-20191379_Z-20191230130848.jpg"/><graphic xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="3-20191379_Z-20191230130848.png"/></alternatives></inline-formula> interstitial dislocation loop in iron

Formation and evolution of interstitial dislocation loop induced by radiation damage in a material are confirmed to seriously affect the performance of the material under irradiation. For example, in body-centered cubic Fe based alloy, 1/2<inline-formula><tex-math id="Z-20191230113253">\begin{document}$\left\langle 111 \right\rangle $\end{document}</tex-math><alternatives><graphic xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="3-20191379_Z-20191230113253.jpg"/><graphic xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="3-20191379_Z-20191230113253.png"/></alternatives></inline-formula> and <inline-formula><tex-math id="Z-20191230113318">\begin{document}$\left\langle 100 \right\rangle $\end{document}</tex-math><alternatives><graphic xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="3-20191379_Z-20191230113318.jpg"/><graphic xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="3-20191379_Z-20191230113318.png"/></alternatives></inline-formula> are mainly formed during the irradiation, which is related to various degradations of material properties. Thus, the understanding of their effect on radiation damages of material is always one of the hottest topics in nuclear material society. Previous studies have shown the surface effect on 1/2<inline-formula><tex-math id="Z-20191230113405">\begin{document}$\left\langle 111 \right\rangle $\end{document}</tex-math><alternatives><graphic xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="3-20191379_Z-20191230113405.jpg"/><graphic xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="3-20191379_Z-20191230113405.png"/></alternatives></inline-formula> loop through the investigation of the interaction between 1/2<inline-formula><tex-math id="Z-20191230113300">\begin{document}$\left\langle 111 \right\rangle $\end{document}</tex-math><alternatives><graphic xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="3-20191379_Z-20191230113300.jpg"/><graphic xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="3-20191379_Z-20191230113300.png"/></alternatives></inline-formula> loop and {111} surface. Considering the difference in property between 1/2<inline-formula><tex-math id="Z-20191230113308">\begin{document}$\left\langle 111 \right\rangle $\end{document}</tex-math><alternatives><graphic xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="3-20191379_Z-20191230113308.jpg"/><graphic xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="3-20191379_Z-20191230113308.png"/></alternatives></inline-formula> loop and <inline-formula><tex-math id="Z-20191230113327">\begin{document}$\left\langle 100 \right\rangle $\end{document}</tex-math><alternatives><graphic xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="3-20191379_Z-20191230113327.jpg"/><graphic xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="3-20191379_Z-20191230113327.png"/></alternatives></inline-formula> loop, in this work the interaction between a <inline-formula><tex-math id="Z-20191230113322">\begin{document}$\left\langle 100 \right\rangle $\end{document}</tex-math><alternatives><graphic xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="3-20191379_Z-20191230113322.jpg"/><graphic xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="3-20191379_Z-20191230113322.png"/></alternatives></inline-formula> loop and {100} surface is studied in detail through the molecular dynamics method. The simulation results indicate that the factors including Burgers vector of loop, loop-to-surface depth, interaction between pre-existing <inline-formula><tex-math id="Z-20191230113337">\begin{document}$\left\langle 100 \right\rangle $\end{document}</tex-math><alternatives><graphic xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="3-20191379_Z-20191230113337.jpg"/><graphic xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="3-20191379_Z-20191230113337.png"/></alternatives></inline-formula> loops, and temperature, all seriously affect the interaction between loop and surface. Especially, the present results show for the first time the evolution of Burgers vector of <inline-formula><tex-math id="Z-20191230113333">\begin{document}$\left\langle 100 \right\rangle $\end{document}</tex-math><alternatives><graphic xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="3-20191379_Z-20191230113333.jpg"/><graphic xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="3-20191379_Z-20191230113333.png"/></alternatives></inline-formula> loop from <inline-formula><tex-math id="Z-20191230113343">\begin{document}$\left\langle 100 \right\rangle $\end{document}</tex-math><alternatives><graphic xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="3-20191379_Z-20191230113343.jpg"/><graphic xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="3-20191379_Z-20191230113343.png"/></alternatives></inline-formula> to 1/2<inline-formula><tex-math id="Z-20191230113348">\begin{document}$\left\langle 111 \right\rangle $\end{document}</tex-math><alternatives><graphic xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="3-20191379_Z-20191230113348.jpg"/><graphic xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="3-20191379_Z-20191230113348.png"/></alternatives></inline-formula> and its one-dimensional diffusion to surface. According to these results, we also further explore the surface evolution after its interaction with loop. The appearance of atomic island results in the rugged surface morphology. All these results provide a new insight into the radiation damage to the surface of material.

Rate theory study of the proton-irradiation induced dislocation loops in modified 310S steels

The evolution of dislocation loops in two types of modified 310S steel under proton irradiation are investigated with rate theory simulations. The growth of dislocation loops is promoted with the increase of irradiation dose and growth rate decreases at high dose, while the number density of the dislocation loops increases first with dose and then tends to be saturated gradually. At the initial stage of irradiation, large number of dislocation loops nucleate and the density increases rapidly. With the development of irradiation, the growth of dislocation loops gradually takes over. Simulation results are in good agreement with the experimental observations of transmission electron microscope. Present study shows that The irradiation resistance of Zr-added steel is slightly better than that of (Nb, Ta, W)-added steel, which may due to the smaller migration energy of interstitials and vacancies in Zr-added steel than that in the (Nb, Ta, W)-added one.

Helium bubble growth under strain fields in tungsten investigated by atomic method

The evolution of helium (He) bubble and dislocation loop from loop punching process affected by external strain field have been studied at atomic scale with hydrostatic strains of around −2% to 2%. The 1st maximal displacements of tungsten atoms around HenV, the formation energy of HenV clusters and the binding energy between the clusters and He are calculated, indicating that the number of He atoms needed to form the 1st self-interstitial atom (SIA) increases from tensile state to compressive state, and the configurations of the 1st SIAs are all in the form of 〈1 1 1〉 crowdion, independent of applied strains. Comparing to the case without external strain effect, the tensile and compressive strains affect not only the critical number of He atoms to punch the 1st loop in loop-punching process but also the Burgers vectors of formed loops. All these results provide a new understanding to radiation damages related with He bubble growth in nuclear materials.

Molecular dynamics study on the Burgers vector transition of nanometric dislocation loops induced by cascade in bcc-iron

Abstracts Dislocation loops are commonly observed defects in bcc-iron based alloys under radioactive environments. They are also major sources of hardening and embrittlement in such materials. In this paper, a comprehensive molecular dynamics (MD) study is conducted concerning the effects of subsequent cascades on previously formed dislocation loops, particularly changes in their Burgers vectors induced by cascades. We reveal the atomistic mechanism behind the Burgers vector transitions and then investigate the effects of temperature, initial cluster type, loop size, and cascade parameters on the transition rate and the effects of different potentials. The competing effects of loop size and cascade energy result in an overall non-monotonic effect on the transition rate to dislocation loops, which is attributed to the density of the dissolved self-interstitial atoms in the loop. We also find temperature dependence of the transition rate of loops with M07 potential function. This study provides essential information regarding the evolution of nanometric dislocation loops under irradiation.

Effect of helium and hydrogen synergy on vacancy migration energy in Fe-10Cr model alloy

Effect of helium and hydrogen synergy on vacancy migration energy (Emv) and microstructural evolution of Fe-10Cr model alloy were investigated by single-He, H and sequential-(He + H) ion pre-implantation and subsequent in-situ electron irradiation at different temperature. Growth rates of irradiation induced dislocation loops were experimentally measured at temperatures ranging from 573 to 773 K to calculate Emv. The loop's size increased as a function of irradiation dose. Moreover, pre-implantation with sequential-(He + H) ion showed an increase in the temperature dependence of the loop growth rate compared to implantation with single He and H. The Emv were determined to be 1.4–1.5 eV in single He and H implanted sample. In-situ electron irradiation experimental results indicated that the Emv be increased due to He and H synergy. Foil surface effect on the calculation of the migration energy has been considered also by discussing a new derivation. This synergic effect of He and H on the evolution of dislocation loop and the mobility of vacancy have been discussed.

Towards understanding the evolution of dislocation loops and their interaction with vacancies in Fe9Cr alloy during the irradiation swelling incubation period

It is well known that body-centered cubic metals have much longer swelling incubation periods than face-centered cubic metals during neutron or charged particle irradiation. In this paper, Fe9Cr model alloy was irradiated with 2 MeV self-ions to lower doses of 0.5, 1.0 and 3.0 dpa at room temperature. The evolution of microstructural defects was revealed by transmission electron microscopy and positron annihilation spectroscopy. Small interstitial dislocation loops were observed in the form of black spots and they were accompanied by a large number of vacancies after 0.5 dpa. Visible dislocation loops dominated the microstructure after irradiation with 1.0 dpa and 3.0 dpa, but the sharp decrement of the S value means that the vacancies should be annihilated gradually with the irradiation dose increasing. In addition, the vacancy status was further confirmed by the investigation of their occupation behaviors with low-energy deuterium atoms. The vacancies located around loops would be trapped and removed by the coarsening interstitial dislocation loops. We believe that the understanding of the interaction between vacancies and dislocation loops is a possible complementary mechanism, applicable for explaining the long stage of swelling incubation in bcc metals.

Loop-punching suppression induced by growth of helium bubble pair in tungsten

Molecular dynamics and transmission electron microscopy are employed to study the formation and evolution of interstitial dislocation loops due to nearby growing helium bubbles. The study reveals a novel mechanism that suppresses further formation of loops. Mass transport of self-interstitial atoms between the bubbles is observed, followed by the loop formation, resulting in a loop-bubble complex in which the Burgers vector may rotate between <100> and 1/2 <111>. Such a complex can absorb free loops, effectively influencing the evolution of loop number density. These results provide a new understanding of loop-bubble complex formation and saturation of the loop density under continuous helium implantation in materials.

The effect of irradiation temperature on damage structures in proton-irradiated zirconium alloys

A study into the effects of irradiation temperature on the damage structures that form during proton-irradiation has been carried out on two commercial Zr alloys in order to develop a more mechanistic understanding of the effect of niobium on dislocation loop evolution. The two Zr alloys (Zircaloy-2 and Low-Sn ZIRLO™) were proton irradiated to a damage level of ∼2 dpa at 280 °C, 350 °C and 450 °C. Detailed dislocation analysis was carried out using on-axis bright-field scanning transmission electron microscopy combined with spectral imaging and synchrotron x-ray line profile analysis. The analysis revealed a significant difference in the effect of irradiation temperature on loop size between the two alloys. In the case of the Nb-free Zr-alloy (Zircaloy-2), an increase in irradiation temperature results in a marked increase in a-loop diameter, by a factor of ∼7.5 from 280 to 450 °C, and a stark decrease in the dislocation line density. In contrast, the Nb-containing Zr-alloy (Low-Sn ZIRLO™) showed very little variation of loop size and line density over the same radiation temperature range. The STEM-based spectral imaging revealed irradiation-induced nano-clustering found throughout the matrix in Low-Sn ZIRLO™, which is not present in the case of Zircaloy-2. Therefore, it is proposed that Nb plays a crucial role in the evolution of dislocation loops in Zr through the formation of irradiation precipitation throughout the matrix.

Evolution of Dislocation Loops Induced by Different Hydrogen Irradiation Conditions in Reduced-Activation Martensitic Steel.

Hydrogen can be induced in various ways into reduced-activation ferritic/martensitic (RAFM) steels when they are used as structural materials for advanced nuclear systems. However, because of the fast diffusion of hydrogen in metals, the effect of hydrogen on the evolution of irradiation-induced defects was almost neglected. In the present work, the effect of hydrogen on the evolution of dislocation loops was investigated using a transmission electron microscope. Specimens of reduced-activation ferritic/martensitic (RAFM) steels were irradiated with hydrogen ions to 5 × 1020 H+ • m−2 at 523–823 K, and to 1 × 1020 H+ • m−2 − 5 × 1020 H+ • m−2 at 723 K. The experimental results reveal that there is an optimum temperature for dislocation loop growth, which is ~723 K, and it is greater than the reported values for neutron irradiations. Surprisingly, the sizes of the loops produced by hydrogen ions, namely, 93 nm and 286 nm for the mean and maximum value, respectively, at the peak dose of 0.16 dpa under 723 K, are much larger than that produced by neutrons and heavy ions at the same damage level and temperature. The results indicate that hydrogen could enhance the growth of loops. Moreover, 47.3% a0 <111> and 52.7% a0 <100> loops were observed at 523 K, but a0 <111> loops disappeared and only a0 <100> loops existed above 623 K. Compared with the neutron and ion irradiations, the presence of hydrogen promoted the formation of a0 <100> loops.

Study on the irradiation effect of mechanical properties of RPV steels using crystal plasticity model

Abstract In this paper a body-centered cubic(BCC) crystal plasticity model based on microscopic dislocation mechanism is introduced and numerically implemented. The model is coupled with irradiation effect via tracking dislocation loop evolution on each slip system. On the basis of the model, uniaxial tensile tests of unirradiated and irradiated RPV steel(take Chinese A508-3 as an example) at different temperatures are simulated, and the simulation results agree well with the experimental results. Furthermore, crystal plasticity damage is introduced into the model. Then the damage behavior before and after irradiation is studied using the model. The results indicate that the model is an effective tool to study the effect of irradiation and temperature on the mechanical properties and damage behavior.

Mesoscale modeling of irradiation damage evolution in bcc iron and vanadium: A comparative study

Voids and dislocation loops are two major types of damages in irradiated structural materials, which are mainly responsible for the degradation of material properties. Here we use the phase-field model and rate theory to simulate the microstructural evolution of voids and dislocation loops, respectively, in irradiated bcc iron and vanadium. The temperature-dependent material parameters of iron and vanadium are derived from ab initio calculations. The simulated results at different temperatures (513 K, 623 K and 722 K) and irradiation doses (1∼20 dpa) are analyzed to reveal the impact of irradiation conditions on the formation of irradiation-induced defect clusters. A comparison of the results shows larger void porosity and void/loop size in iron and higher void/loop density in vanadium. Then, a dispersed-barrier hardening model is used to correlate the mesoscale simulation results on microstructure with the yield stress change of the materials.

Influence of irradiation temperature on void swelling in NiCoFeCrMn and NiCoFeCrPd

Using ion irradiation and transmission electron microscopy, temperature dependent void swelling and dislocation loop evolution in equiatomic NiCoFeCrMn and NiCoFeCrPd high entropy alloys were investigated. Voids were observed in all tested conditions. An order higher of swelling was observed in both HEAs as temperature increased from 420 to 580 °C. Alloying with Pd was found to have a stronger suppression effect on void and dislocation loop growth than alloying with Mn. Possible contributions to the swelling resistance were explored, including localized lattice distortion and higher defect migration barrier in NiCoFeCrPd. Additionally, no phase decomposition was observed in either alloy.

Understanding irradiation-induced nanoprecipitation in zirconium alloys using parallel TEM and APT

We investigate nano-scale irradiation-induced precipitation in a ZrSnFeCrNi-alloy (Zircaloy-2) by combining atom probe tomography (APT) for chemical detail with scanning transmission electron microscopy (STEM) and high resolution energy dispersive X-ray (EDX) spectroscopy for wider context and complimentary and correlative TEM diffraction techniques for crystallographic relationships. We find that Fe and Cr-rich nano-rods precipitate in Zircaloy-2 following proton irradiation at 350 °C to a low dose of ∼2 dpa. The long-axis of the nano-rods are aligned in a direction 12–15° from the Zr matrix 〈0001〉, align in the basal plane and are of width 1.5–5 nm. Smaller rods are of APT-determined composition Zr4(Fe0.67Cr0.33), tending towards Zr3(Fe0.69Cr0.31) as the rod volume increases to > ∼400 nm3, in agreement with STEM-EDX determination of composition resembling that of Zr3Fe with Cr replacing some of the Fe. The Fe/Cr ratio has been shown to increase with distance from the nearest partially-dissolved Zr(Fe,Cr)2 phase particle. The nucleation of nano rods has implications for macroscopic irradiation-induced deformation phenomena, irradiation-induced hardening and the evolution of dislocation loops and other defects.

Synergistic effects on microstructural evolution and hardening of the Hastelloy N alloy under subsequent He and Xe ion irradiation

Synergistic effects in the evolution of He bubbles and dislocation loops in Hastelloy N alloy were investigated using transmission electron microscopy and nanoindentation. Results show that the sizes of both He bubbles and dislocation loops induced by the He and Xe irradiation were larger than those in the individual irradiation cases. The corresponding hardening degree of irradiated sample was also measured to be higher than the combined hardening value of two individual irradiation cases. The underlying physics controlling the synergistic effects have revealed that the vacancies are preferentially captured by pre-existed He bubbles as compared to their annihilation with the interstitial atoms during the subsequent Xe irradiation.

The coupled atomistic/discrete-dislocation method in 3d. Part III: Dynamics of hybrid dislocations

In two companion papers, the CADD-3d method for coupling atomistic and discrete dislocations was described and validated for (quasi-)static equilibrium. Here, CADD-3d is validated further for the time evolution of straight and curved hybrid dislocations, i.e. dislocations that span both atomistic and continuum domains simultaneously. Studies show physically sensible results when using precise calibration of the discrete dislocation mobility and atomistic core structures, and the sensitivity of results when intentional deviations away from the precise calibrations are used. CADD-3d is then used to model the evolution of dislocation loops from atomistic Frank–Read sources under sustained load. The predicted nucleation and expansion of multiple loops is in excellent agreement with fully atomistic simulations at small scales. CADD-3d is then applied to much larger scales that capture both the nucleation of loops at the atomic scale and the evolution of a dislocation loop pile-ups at the micron scale, with full interactions among all dislocations in the system during the entire time evolution. Collectively, the results here serve to demonstrate the accuracy, robustness, and power of CADD-3d, and point toward its future application in many complex dislocation problems requiring both atomistic resolution and large-scale dislocation evolution.

Review and critical assessment of dislocation loop analyses on EUROFER 97

The understanding of microstructural defects behavior after neutron irradiation is crucial for assessing the applicability of reduced activation ferritic/martensitic (RAFM) steel EUROFER 97 in future fusion reactors. Formation and evolution of dislocation loops is believed to play the major role in material's hardening under neutron irradiation. In this work, transmission electron microscopy (TEM) data on dislocation loop size distribution is provided after different irradiation campaigns to determine the role of neutron dose on the dislocation loop evolution. A comparison of investigations on dislocation loop behavior and appearance yield considerable differences. For a conclusive interpretation, this work reviews available data, and possible reasons for the observed differences are discussed. Recommendation for future TEM investigation are given.

Dislocation loops in ultra-high purity Fe(Cr) alloys after 7.2 MeV proton irradiation

Ultra-high purity Fe(Cr) alloys (from 0 wt% Cr to 14 wt% Cr) were 3D homogeneously irradiated by 0–7.2 MeV protons to 0.3 dpa at nominal temperatures from 270 °C to 500 °C. Microstructural changes were observed by transmission electron microscopy (TEM). The results showed that evolution of dislocation loops depends on the Cr content. Below 300 °C, large ½ a0 <111> loops are dominating. Above 300 °C, a0 <100> loops with a habit plane {100} appear. Loop sizes of both types are more or less the same. At temperatures from 310 °C to 400 °C, a0 <100> loops form clusters with the same {100} habit plane as the one of the loops forming them. This indicates that <100> loops of the same variant start gliding under mutual elastic interaction. At 500 °C, dislocation loops form disc shaped clusters about 1000 nm in diameter and sitting on {111} and/or {100} planes in the pure Fe samples. Based on these observations a quantitative analysis of the dislocation loops configurations and their temperature dependence is made, leading to an understanding of the basic mechanisms of formation of these loops.

Rate Theory for Dislocation Loops Evolution in AL-6XN Austenitic Stainless Steel under Proton Irradiation

The average size and density evolution of dislocation loops in AL-6XN austenitic stainless steel, a candidate fuel cladding material for supercritical water-cooled reactor, under proton irradiation were simulated through a rate theory model. The simulation results exhibit relatively good agreement with the experimental results at 563 K. The size and density of defect clusters are calculated under irradiation temperature between 550 K and 900 K and irradiation doses up to 15 dpa which satisfies the working condition in supercritical water-cooled reactor. The fast nucleation between self-interstitials happens at the initial stage of irradiation. The average size of dislocation loops increases while the average density of these loops reduces with the increasing temperature, and the average density approaches to a constant when irradiated at higher irradiation doses. The mechanism is discussed based on the variation of rate constants of defect reactions and the variation of the diffusion coefficients of interstitials and dislocation loops with dose and temperature.

Effect of post-irradiation annealing on the irradiated microstructure of neutron-irradiated 304L stainless steel

Post-irradiation annealing was performed on a 304L SS that was irradiated to 5.9 dpa in the Barsebäck 1 BWR reactor. Evolution of dislocation loops, radiation-induced solute clusters and radiation-induced segregation at the grain boundary was investigated following thermal annealing at 500 °C and 550 °C up to 20 h. Dislocation loops, Ni-Si and Al-Cu clusters, and enrichment of Ni, Si and depletion of Cr at the grain boundary were observed in the as-irradiated condition. Dislocation loop size did not change significantly after annealing at 550 °C for 5 h but the loop number density decreased considerably and loops mostly disappeared after annealing at 550 °C for 20 h. The average size of Ni-Si and Al-Cu clusters increased while the number density decreased with annealing. The increase in cluster size was due to diffusion of solutes rather than cluster coarsening. Significant volume fractions of Ni-Si and Al-Cu clusters still remained after annealing at 550 °C for 20 h. Substantial recovery of Cr and Ni at the grain boundary was observed after annealing at 550 °C for 5 h but neither Cr nor Ni was fully recovered after 20 h. Annihilation of dislocation loops, driven by the thermal vacancy concentration gradient caused by the strain field and stacking fault associated with the loops appeared to be faster than annihilation of solute clusters and recovery of Ni and Si at the grain boundary, both of which are driven by the solute concentration gradients.

Atomistic Simulation of Interstitial Dislocation Loop Evolution under Applied Stresses in BCC Iron

Evolution of an interstitial 1/2 dislocation loop under tensile, shear, and torsion stresses is studied with molecular statics method. Under a tensile stress, the dependence of ultimate tensile strength on size of loop is calculated. The formation of small shear loops around the initial prismatic loop is confirmed as an intermediate state to form the final dislocation network. Under a shear stress, the rotation of a loop is observed not only by a change of the habit plane but also through a transformation between a shear and a prismatic loop. Under torsion, a perfect BCC crystal may undergo a BCC to FCC or BCC to HCP transformation. The present work indicates that a 1/2 loop can delay these transformations, resulting in the formation of micro‐crack on the surface.