Number Density Of Dislocation Loops Research Articles

Enhancing strength and irradiation tolerance of Mg alloy via co-addition of Mn and Ca elements

The Mg alloys with combination of high strength and excellent irradiation resistance are currently required for research reactors. In this work, the novel Mg-3Mn-0.5Ca alloy with a high strength of over 300 MPa has been fabricated via co-addition of Mn and Ca elements. Moreover, the Xe ion implantation (300°C/4.5 × 1015 ions/cm2) is conducted for pure Mg, Mg-3Mn and Mg-3Mn-0.5Ca alloys. Microstructure characterization shows that the number density of dislocation loops is comparable between Mg-3Mn and pure Mg, while the phase boundary of nano-Mn particles could act as the sink to absorb more Xe atoms, resulting in the abundant formation of Xe bubbles in the matrix of irradiated Mg-3Mn alloy. With further minor addition of Ca element, the formation of Xe bubbles and dislocation loops in Mg-3Mn-0.5Ca alloy has been obviously suppressed, and the abundant grain boundaries (GBs) due to grain refinement then act as the sink region for Xe precipitation. The limited number density of Xe bubbles leads to the extremely low swelling ratio in Mg-3Mn-0.5Ca alloy, ∼0.08%. The result above suggests that the Mn and Ca co-addition could enhance the mechanical properties and irradiation tolerance of Mg alloys, simultaneously. The present low-alloying strategy would provide a new design reference for novel nuclear materials.

Cluster dynamics study on nano damage of RPV steels under proton irradiation at 290°C

Irradiation-induced defects such as dislocation loops, cavities or solute clusters and chemical composition segregation of reactor pressure vessel (RPV) steel are the root causes of irradiation embrittlement. Combining two nucleation mechanisms, namely, the uniform nucleation and non-uniform nucleation of solute clusters (such as Cu-rich phase), a cluster kinetic simulation was established based on the reaction rate theory, and the co-evolution of matrix damage and Cu-rich phase in low-copper RPV steel was simulated under irradiation. And the average size and number density of defective clusters and solute clusters were established with irradiation dose. Compared with the average size and number density of dislocation loops observed by transmission electron microscopy (TEM) of proton irradiated RPV steel at 290°C, the verification results show that the cluster dynamics model considering both the nucleation mechanism of interstitial dislocation loops and vacancy clusters can well simulate the irradiation damage behavior of materials.

The effect of stress state and He concentration on the dislocation loop evolution in Ni superalloy irradiated by Ni+ & He+ dual-beam ions: In-situ TEM observation and MD simulations

In-situ TEM observation was conducted during Ni+&He+dual-beam irradiation to monitor the evolution of dislocation loops accompanied by He bubbles in the Ni-based alloy GH3535. Two distinct evolutions of dislocation loops, driven by residual stresses, were observed within the monitored grains. Hence, molecular dynamics (MD) simulations were employed to reveal the effects of stress magnitude and direction on loop evolution, including size, number density, type and variation. The simulations revealed that the presence of compressive stress reduced the formation energy of perfect dislocation loops, thus promoting their formation. Stress state was found to influence the preferential orientation of the loops, and compressive stress resulted in a decreased number density of dislocation loops but an increase in their size. This establishes a clear relationship between stress state and magnitude and the evolution of dislocation loops during ion beam irradiation. Additionally, the nature and characteristics of dislocation loops were quantified to explore the effects of He concentrations on their evolution. The higher He concentration not only promotes the nucleation of dislocation loops, leading to their higher number density, but also facilitates the unfaulting evolution by increasing the stacking fault energy (SFE). Moreover, the accumulation of He in the lower-He-concentration sample led to the growth of dislocation loops in multiple stages, explaining their nearly identical average sizes when compared to the higher-He-concentration sample.

Defects analysis in low displacement damage neutron-irradiated austenitic stainless steels

Analysis of neutron irradiation-induced defects is carried out using Transmission Electron Microscope (TEM) on three variants of austenitic stainless steels, SS 304 L(N), SS 316 L(N), and SS 316 grid plate material irradiated in the Indian Fast Breeder Test Reactor at Kalpakkam to low displacement damage levels of 1.76–6.75 dpa at an irradiation temperature of 380 °C-400 °C and dpa rate in the order of 10−7dpa/s. There is a significant variation in the faulted dislocation loop size distribution and number density as a function of irradiation dose among these stainless steels due to their different stacking fault energies which is evident from the quantified defects data presented. At ∼5.08 dpa, SS 304 L(N), SS 316 L(N) and SS 316 exhibited faulted dislocation loop number densities of 2.3 × 1022/m3, 4.8 × 1021/m3, and 1.2 × 1022/m3 with mean loop sizes of 15.5 ± 6.3 nm, 35.4 ± 13.7 nm, and 16.1 ± 5.6 nm respectively. Thus, at an identical dpa level, SS 304 L(N) and SS 316 L(N) have the highest and the lowest faulted loop number densities respectively. The general size-distribution behaviour, the increment in the loop size with enhanced size distribution as a function of dpa is observed only for SS 316 L(N). The dislocation density data previously determined using X-ray diffraction analysis is correlated to the number density of dislocation loops along with their size distribution determined using TEM analysis. Finally, the change in the yield strength due to the increment in the faulted dislocation loop number density is determined.

The effect of hydrogen-helium synergism on dislocation loops in RAFM steel at different temperatures by dual-beam ion irradiation

Simultaneous Fe+H ion irradiation (denoted as Fe+H) and Fe+H ion irradiation with pre-implantation of He (denoted as He/Fe+H) were carried out on Chinese Low-Activation Martensitic (CLAM) steels at 350 ℃-550 ℃ to investigate the synergistic effect between H and He. The peak dose was 15 dpa, the He concentration in the observation area was 11 appm/dpa and the H concentration was 44 appm/dpa. The microstructure and hardening of the irradiated samples were evaluated by TEM and nanoindentation, respectively. Bubbles were not observed in irradiated CLAM steels at all temperatures, which implies that the promotion of bubbles by H may not be as strong as that by He and that H attenuates the promotion of bubbles by He. The hardening induced by He/Fe+H irradiation at 450 °C is about 1.6 times that of Fe+H irradiation, the synergistic effect of H and He is most significant at this temperature. Comparison with Fe+He irradiation and single Fe irradiation reveals that Fe+H irradiation at 350 °C induces lower interstitial defect damage than Fe+He irradiation, but the level of interstitial defects induced by the synergistic effect of both H and He with displacement damage is comparable at higher temperatures. The average size of dislocation loops induced by Fe+H irradiation is between that of single Fe irradiation and Fe+He irradiation, but the number density of dislocation loops induced by Fe+H irradiation is higher than that of Fe+He irradiation. The synergistic effect of He with displacement damage tends to promote the growth of dislocation loops, while the synergistic effect of H tends to promote the nucleation of dislocation loops.

Microstructure Characterization and Hardening Evaluation of Ferrite/Martensitic Steels Induced by He2+ Irradiation

Two ferrite/martensitic (F/M) steels with different Si concentrations (0 and 0.4 wt.%) were irradiated by 250 keV He2+ ions with different fluences of 2 × 1016 ions/cm2 and 1 × 1017 ions/cm2. Transmission electron microscopy and a nanoindenter were employed to investigate their microstructure evolution and irradiation hardening effects induced by high-energy He2+ ions. A large number of He bubbles formed in the Si-free and Si-containing F/M steels, which preferentially nucleated and grew at the lath and phase boundaries. Owing to the inhibiting effect of Si addition on He bubble growth, the He bubbles in the Si-containing sample exhibited smaller size and higher density at the same He2+ fluence. Nanoindenter measurement revealed that typical irradiation hardening was observed in the F/M steel, and 1/2<111> and <100> type dislocation loops formed by He2+ irradiation was recognized as the dominant mechanism. The addition of Si induced an increase in the number density of dislocation loops, leading to the exacerbation of the irradiation hardening, and the results are basically in agreement with the theoretical analysis based on the dispersion barrier hardening (DBH) and Friedel–Kroupa–Hirsch (FKH) models.

Enhanced irradiation tolerance of a medium entropy alloy via precipitation and dissolution of nanoprecipitates

High/medium entropy alloys (HEAs/MEAs) are good candidates for nuclear applications due to the excellent mechanical properties, good corrosion resistance and radiation resistance. In this work, a novel cobalt-free MEA was developed by introducing L12 nanoprecipitates. The microstructure evolution and radiation tolerance were evaluated after bombarded using 3 MeV Fe11+ ions at 500 °C. The evolution of nanoprecipitates was closely related to the irradiation dose, and dominated by irradiation-enhanced diffusion and ballistic dissolution mechanism. For the solid-solution MEA (without L12 nanoprecipitates), irradiation hardening occurred due to the irradiation-induced formation of precipitates, voids and dislocation loops. However, in MEA with L12 nanoprecipitates after aging, irradiation induced dissolution and reprecipitation of nanoprecipitates were observed. Different from the solid-solution MEA, the hardness kept almost unchanged in the aging sample after irradiation. The swelling rate of the solid-solution sample after irradiation is 9.4 × 10−6 %/dpa, while no swelling occurs in the aging sample under the same irradiation condition. Besides, the average size and number density of dislocation loops in the aging sample decreased by ∼ 40% and ∼ 28%, compared with the solid-solution sample. The precipitation and dissolution of nanoprecipitates substantially improved the radiation tolerance of the cobalt-free MEA.

An in-situ TEM investigation of microstructure evolution in Cr-coated zirconium alloy under heavy ion irradiation: Simultaneous evolution of both Cr coating and Zr-4 matrix

Cr-coated zirconium (Zr) alloy is one of the potential candidate materials for accident tolerant fuel (ATF) cladding. However, little attention has been paid in the simultaneous microstructure evolution of Cr coating and Zr alloy matrix under irradiation, which is necessary for comprehensive evaluation of the alloy irradiation performance. In the present work, in-situ ion irradiation experiment was conducted to capture the major features during simultaneous microstructure evolution of Cr coating and Zr alloy at various temperatures. It is found that at 300 ℃ the size and number density of dislocation loops increase with irradiation dose, but at higher temperature the tendency seems different, which is related to the formation of dislocation networks. Besides, the extent of precipitate amorphization in Zr alloy matrix decreases with temperature due to the recovery of irradiation defects. The present work may provide useful insights into the irradiation property of the Cr-coated Zr alloy for ATF applications.

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.

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 helium and its synergism with displacement damage on microstructure of CLAM steel investigated by single, sequential and simultaneous Fe and He ion irradiations

The reduced-activation ferritic/martensitic steel, China Low-Activation Martensitic (CLAM) steel, was irradiated with single Fe, sequential He/Fe, sequential Fe/He, and simultaneous Fe+He ion beam irradiations at 350 ℃–550 ℃ to a peak displacement damage dose of 15 dpa with 0 - 110 appm/dpa He to explore the effect of He and its synergism with displacement damage on microstructure evolution. The microstructures of all the irradiated specimens were investigated by transmission electron microscope (TEM). For Fe/He sequential irradiation, He inhibited the growth of dislocation loop induced by high-temperature He post-irradiation and reduced the number density of loops. Non-spherical He bubbles were observed only at post-injection He at 550 °C. The size of non-spherical He bubbles produced at 110 appm/dpa was about twice as large as 11 appm/dpa, and the number density increased about 6.4 times. The number density of dislocation loops induced by sequential He/Fe irradiation is the highest, and the average size of loops is comparable to that caused by simultaneous Fe+He irradiation. Moreover, the sequential Fe/He irradiation induced the lowest number density of loops of approximately 64% of the simultaneous Fe+He irradiation. The spherical bubble was observed in all irradiation with He injection above 350 ℃. The average size of spherical He bubbles induced by simultaneous Fe+He irradiation is the largest, and the number density is about 2.6 times that of the sequential Fe/He irradiation. The synergisms of He in different irradiation sequences were analyzed systematically, which helps to better understand the evolutionary processes of the synergistic effect of He and displacement damage.

In-situ TEM investigation of dislocation loop evolution in Al-forming austenitic stainless steels during Fe+ irradiation: Effects of irradiation dose and pre-existing dislocations

Alumina-forming austenitic (AFA) stainless steel with excellent properties has been regarded as a promising candidate of fuel claddings in Supercritical carbon dioxide (S-CO2) gas cooled nuclear reactor. In current work, in-situ irradiation with 400 keV Fe+ in transmission electron microscope were performed on AFA stainless steel to investigate the influences of irradiation dose and pre-existing dislocation lines on loop evolution at 773 K. The irradiation damage caused by irradiation-induced loops was also analyzed. In-situ TEM observation displayed the evolution and characteristics of dislocation loops, including initiation, migration, aggregation, growth, annihilation, and combination. Both Frank loops with b = 1/3<111> and perfect loops with b = 1/2<110> were formed, and at 0.54 dpa, 1/2<110> loops accounted for the majority and was up to 64.6%. Pre-existing dislocation lines obviously affected not only the evolution and distribution of dislocation loops, but also the degree of irradiation hardening. At the same irradiation dose, not only the size and number density of dislocation loops in the region with dislocation lines were smaller than those in the region without dislocation lines, but also the increment of yield strength and irradiation damage was lower before the formation of dislocation network. Increasing pre-existing dislocation density should be beneficial to improve the irradiation resistance of materials, but it needed to be comprehensively considered with the mechanical properties of the material. Al component played an important role in irradiation behavior of AFA steels and the corresponding mechanism was discussed subsequently.

In-situ TEM study of loop evolution in FeCrAl alloy under Fe+ irradiation: Effects of temperatures, dislocations and precipitates

Fe13.5Cr4.7Al alloy was irradiated in-situ with 400 keV Fe+ in TEM at RT and 748 K. The evolution of dislocation loops under the influence of temperatures, irradiation dose, dislocation lines and precipitates were characterized in detail. The loop rafting was found at RT irradiation for the first time and the growth direction of the rafted loops was along g = 110 direction. However, this phenomenon was not observed at 748 K and only the dislocation complex was appeared. The formation conditions of loop rafting were low temperature and a certain value of irradiation dose. The average size and number density of dislocation loops as a function of irradiation dose in the regions with dislocation lines, precipitates and no dislocation lines/precipitates were constructed and analyzed comparatively. At the same irradiation dose, sample irradiated at higher temperature has larger average loop size and lower loop number density. There was a complex interaction between dislocation line-loop at 748 K, but few interactions were observed at RT. At any irradiation dose, the loop number density in the region with dislocation lines was smaller than that in the region without dislocation lines, but the average loop size was slightly opposite. Moreover, the higher the irradiation temperature, the more obvious the effects of precipitates and dislocation lines on loop evolution were. Our current results provide a fundamental insight into how the dislocations, precipitates and dose affect loop evolution and characteristics in FeCrAl alloy at elevated temperatures.

In-situ TEM study of the effect of pre-existing dislocation on loop evolution in 508-III steel during Fe+ irradiation

This study explores the evolution processes of initiation, migration, growth, annihilation, and aggregation for dislocation loops and dislocation complex formation in regions with and without pre-existing dislocations. In-situ TEM experimental observations are conducted for 508-III steel, commonly used as a pressure vessel steel in nuclear reactors, during 400 keV Fe+ irradiation at 573 K. Loop initiation is always observed, even following the formation of dislocation complex. The average size and number density of dislocation loops, in regions both with and without pre-existing dislocations, are obtained as a function of the irradiation dose. At the initial stage of irradiation, pre-existing dislocations significantly influence the evolution of dislocation loops. However, with an increase in the irradiation dose, this influence gradually weakens. The in-situ TEM experimental observations of microstructure evolutions in this study are of significant value for furthering our understanding of irradiation damage to 508-III steel.

Effect of solute elements (Ni, Mn) in Fe-based alloys on dislocation loop evolution under Fe2+ ion irradiation

The dislocation loop is known as one of the irradiation defects that contributes to hardening and accompanying embrittlement in reactor pressure vessel steels. In this study, the effect of solute elements such as Ni and Mn in Fe-based alloys on dislocation loop evolution was investigated using ion irradiation. Pure Fe and Fe model alloys, Fe-0.6Ni and Fe-0.6Ni-1.4Mn, were irradiated by 2.8 MeV Fe ions to a peak dose of 3 displacements per atom (dpa) at 400 °C. Quantitative analysis of transmission electron microscopy (TEM) was conducted to analyze the evolution of dislocation loops. As a result, the presence of solute elements in Fe alloys increased the number density of dislocation loops significantly and decreased their mean size slightly. Both types of dislocation loops, ½<111> and 〈100〉 Burgers vectors, were observed in all specimens. Solute elements reduced the mobility of ½<111> dislocation loops, resulting in an increase in their number density and restricting the growth of <100> dislocation loops. Furthermore, an increase in the number density of <100> dislocation loops of small size was found with higher solute contents.

Characterization of hardening behavior of carbon steels at low temperature irradiations

• Embrittlement factors of carbon steels irradiated at low temperature were fine dislocation loops and solute clusters of Mn-Si-Cu atoms. • Carbon steels and their weld metals containing high concentrations of Cu, Mn and Si had relatively high radiation hardening due to promotion of solute atom clustering. • Low temperature irradiation resulted in higher radiation hardening due to the higher number density of dislocation loops, despite the lower number density of solute clusters. • Low dose rate irradiation (electron irradiation) resulted in greater radiation hardening due to the higher number density of solute clusters than high dose rate irradiation (ion irradiation) did. Overpacks used for geological disposal of high-level radioactive waste will be irradiated by neutrons and γ rays with very low dose rate for a long time. A possible way to predict effects of such radiations is that microstructure changes under the expected irradiation conditions are estimated by both experiments and analytical modeling and then the changes in hardness and fracture toughness from the microstructure changes are estimated. In the present study, carbon steels, which are one of the candidate materials for overpacks, model alloys and a weld of carbon steel were irradiated with either Fe ions or electrons at 90 or 290 °C, and then the microstructure and hardness were investigated to gain an understanding about embrittlement factors of carbon steels and effects of both material composition and irradiation conditions (temperature, dose rate) on radiation-induced changes in microstructure and hardness. Formations of fine dislocation loops and solute clusters containing Mn, Si and Cu atoms were confirmed by microstructure analysis using transmission electron microscopy and atom probe tomography. Hardness increase was well correlated with the evolution of dislocation loops and solute clusters, indicating that these defects are embrittlement factors of carbon steels. Higher radiation hardening was confirmed for the materials containing a high density of Mn, Si and Cu, and the irradiation conditions of low temperature and low dose rate.

Effects of aging and irradiation on Fe-Ni-Al alloy

To explore the effects of aging and irradiation on the formation of solute clusters, Fe-2.8Ni-2.0Al (at.%) alloys after solid-solution treatment were aged at 500 ℃ for 16 h and 32 h and irradiated with Ar ions to the fluence of 4 × 1016 ions·cm−2 at 290 ℃, respectively. Transmission electron microscope and atom probe tomography were used to characterize the microstructure and solute atoms distribution. Results showed that B2-NiAl clusters were formed in aged alloys and the number density increased with time. The number density of clusters was 8.35 × 1022 m−3 after 32 h at 500 ℃ aging. However, high number density of dislocation loops and B2-FeNi clusters were found after irradiation at 290 ℃. The number density of clusters reached to 3.61 × 1023 m−3. The nucleation of B2-NiAl clusters under aging is driven by the phase transform free energy with the lower mixing enthalpy of Ni and Al atoms, while the formation of B2-FeNi clusters is due to the combination of the interaction between dislocation loops and Ni solutes with the enhanced diffusion rate of Ni atoms due to supersaturated vacancy.

The Effect of Yttrium Addition on the Microstructure and Irradiation Hardening in V-4Cr-4Ti Alloy under Self-Ion Irradiation

Microstructure and irradiation hardening of V-4Cr-4Ti alloys with different yttrium (Y) contents were studied by self-ion irradiation at 550 °C via TEM and nano-indentation test technology. The peak damage of the V-4Cr-4Ti-xY alloy (x = 0, 0.1, 0.5, and 1, wt.%) irradiated by 2.5 MeV self-ion (V2+) is 8 dpa. Dense dislocation loops were observed in all vanadium alloy samples after irradiation. With the increase of Y content, both average size and number density of dislocation loops using g = &lt;110&gt; near the pole [001] decreased, while the irradiation hardness increment first decreased and then increased. In order to better reduce the irradiation hardening, it is considered that the addition of 0.1 wt.% Y in V-4Cr-4Ti alloy is reasonable.

The effects of irradiation on CrMnFeCoNi high-entropy alloy and its derivatives

High-entropy alloys (HEAs), based on equiatomic or near-equiatomic mixture of usually four or more elements, have attracted considerable attention as there are indications that along with surprising microstructural simplicity of some alloys, they may also offer intriguing combinations of mechanical and other properties. Amongst these properties, there is growing interest in the irradiation response of HEAs and their potential to withstand the neutron bombardment environment of future civil nuclear power plants. One of the first proposed HEAs is face-centred cubic CrMnFeCoNi, also known as the Cantor alloy, and the irradiation response of the Cantor alloy and its sub-systems are the focus of this review. Using irradiation analogues (electrons, heavy ions and He) to neutron bombardment and considering simulations, advanced microstructural analysis and property measurement, the Cantor alloy and its derivatives are shown to exhibit encouraging irradiation resistance that, in many instances, is superior to more traditional dilute alloys of the same elements. The beneficial aspects are high phase stability and resistance to radiation-induced segregation, smaller size but higher number density of dislocation loops, significantly lower extent of swelling and improved resistance to He bubble growth. The future research directions for irradiation resistant HEAs are also suggested.

In-situ TEM observation of the evolution of helium bubbles & dislocation loops and their interaction in Pd during He+ irradiation

The microstructural evolution of purity Pd under 30 keV He+ irradiation at 573 K was investigated by in-situ transmission electron microscopy. The nucleation, growth, merging, annihilation, size change, number density variation, and types of dislocation loops were analyzed under the influence of irradiation fluence and sample thickness. Both perfect dislocation loops with b = 1/2 < 110> and faulted dislocation loops with b = 1/3 < 111> were formed. However, at low irradiation fluence, most of the loops were 1/3 < 111> loops. The thickness of TEM foil obviously affected the ratio of 1/3 < 111> loop variants, the size and number density of dislocation loops, and the characteristics of bubble-loop complexes. With the increase of irradiation fluence, the size of dislocation loops increased, but loop volume number density remained almost constant until dislocation loops merged and evolved into dislocation network. There was an obvious interaction between dislocation loops and bubbles, indicating that 1/3 < 111> loop was first formed at the initial stage of irradiation, and when the loop grew to a certain size, obvious helium bubbles appeared inside its region.