Number Density Of 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.

Correlating microstructure and mechanical properties of harvested high dose Zorita light water reactor internals

In this study, microstructural studies and micro-mechanical testing of an ex-plant material harvested from the decommissioned pressurized water reactor (PWR) is carried out. Irradiated 304 stainless steel (SS) components were harvested by the Electric Power Research Institute, U.S. Nuclear Regulatory Commission, and members of the José Cabrera Nuclear Power Station (Zorita) in Spain. The bi-crystalline micro-tensile specimens, atom probe tomography (APT) tips, and transmission electron microscopy (TEM) lamellae were fabricated from the baffle plate materials irradiated to 0.05, 15 and 50 dpa to elucidate microstructural, microchemical, and local mechanical properties changes as a function of dose. TEM and APT studies reveal radiation induced Ni, Cr, and Si rich precipitates in the 15 dpa sample that become smaller and partially redissolve into the matrix of the 50 dpa sample (decreasing size and number density). Dislocation loop and cavity number density and size were quantified as a function of dose as well as swelling. Intergranular Ni and Si enrichments with concomitant Fe and Cr depletion were observed in the 15 dpa sample and were more pronounced in the 50 dpa sample. Micro-tensile testing performed in the scanning electron microscope (SEM) at room temperature and 300 °C shows that the material's local yield strength and ultimate tensile strength decreases with increased dose and elevated test temperature and provides further insights via localized strain mapping. Current and previous mechanical data was compared with calculated values using the dispersed barrier hardening model with inputs of dislocation loops, precipitates, and cavities from microstructural characterization.

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.

In-situ TEM characterization of dislocation loops in tungsten under simultaneous helium and hydrogen irradiation

The interaction among He, H, and irradiation-induced dislocation loops in pure tungsten was examined through 30 keV simultaneous He+ and H2+ irradiation at 1073 K. Three different He/H concentration ratios (0.1, 1, and 5) were utilized. This study involved the characterization of the number density, average size, and size distribution of loops by varying the dose percentage of He and H. Increasing the dose percentage of He facilitated the loop nucleation. Increasing the total irradiation dose not only increased loop nucleation sites but also changed the loop size distribution from unimodal to bimodal. The specific Burgers vectors and habit planes of loops were carefully determined using the inside-outside method in combination with the g·b invisibility criterion and crystallographic map. The proportion of interstitial 1/2<111> loops exceeded 95 %, in which four Burgers vector variants of 1/2[111], 1/2[1¯11], 1/2[1¯1¯1], and 1/2[11¯1] were identified. Increasing the dose percentage of He led to an even distribution of Burgers vector variants. The {110} habit planes were found to deviate at an angle of 35° from the {111} ones. A majority of {111} habit planes were found compared to {110} for 1/2<111> loops. Altering the dose percentage of He or H showed no apparent effect on the habit planes of {111} or {110} for 1/2<111>loops, nor did it affect their average size when situated on {111} or {110}.

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.

Nano-indentation deformation behavior of heavy ion irradiated reduced activation ferritic/martensitic (RAFM) steels

The Nano-indentation deformation behavior of the heavy ion irradiated RAFMs is investigated in the present work. Specimens of the CLF-1 steel were irradiated with 123.4 MeV 20Ne ion at both -50 °C and 400 °C, respectively, to achieve a mean displacement damage level of 0.15 dpa. A quasi-uniform distribution of the displacement damage was produced in the specimens by utilizing an energy degrader at the irradiation chamber. The results of nano-indentation experiments indicate that during the loading stage, the irradiated materials exhibit a lower strain rate, but during the holding stage after reaching the maximum load, the irradiated materials show a higher strain rate than the un-irradiated material. Transmission electron microscopy observation showed that the low-temperature irradiation produced a high density of irradiation-induced dislocation loops, which impeded the glide of dislocation lines during the loading stage. However, as the load increases gradually to the holding stage, the dislocation lines can overcome the pinning by the dislocation loops. The strain that did not occur due to the pinned effect of the dislocation loops in the loading stage was released during the holding stage, resulting in a higher strain rate in the irradiated specimen during the holding stage. Based on the Peierls mechanism, a quantitative relationship between the number density of the irradiation-induced dislocation loops and the strain rate during the holding stage was established. The calculated results agree well with the experimentally obtained strain rate during the holding stage.

Irradiation dose and temperature effects on BCC material with dislocation based crystal plasticity

Numerical modelling of temperature and strain rate dependent plasticity of body centered cubic (BCC) materials is carried out based on dislocation based crystal plasticity model. In view of the strong influence of radiation effects on the response of BCC materials, the constitutive equations are further extended to incorporate the effect of irradiation-induced defects. Size of irradiation defects produced (mainly dislocation loops) is mostly controlled by the irradiation temperature, whereas the defect number density is governed by the irradiation dose. The present constitutive model is introduced to predict the irradiation conditions and temperature dependent plasticity for BCC materials. The model is used to simulate the different irradiation conditions in terms of dislocation loop size and number density. To have a quantitative effect of irradiation conditions on material toughness, the Defect Induced Apparent Temperature Shift (ΔDIAT) is estimated and compared with available experimental evidence, for validation.

Dislocation loops, segregation and hardening induced by high-dose ion irradiation of NbMoTaW and VCrTaW high-entropy alloy coatings on the T91 substrate

NbMoTaW and VCrTaW high-entropy alloys (HEA) coatings with near-equal atomic ratios were deposited onto the T91 substrate by magnetron sputtering. The two HEA coatings and the T91 substrate were irradiated to the peak damage 50 displacements per atom (dpa) and 100 dpa using 2.7 MeV Si2+ at 550 °C. X-ray diffraction (XRD), Scanning electron microscopy (SEM), transmission electron microscopy (TEM), energy dispersive X-ray spectroscopy (EDS) and nanoindentation techniques were used to study the changes in grain morphology, microstructure, element distribution and hardness before and after irradiation. The results showed that the grain structure was stable and the coatings were closely bonded to the substrate after irradiation. The elements (Ta and W) near the grain boundary in NbMoTaW coating were depleted before irradiation, but they became homogenized after irradiation. A large number of dislocation loops were observed in both the T91 substrate and HEA coatings after irradiation, the HEA coatings significantly increased the doses required for the full evolution of the dislocation loops. With the increase of damage doses, the dislocation loops size of HEA coatings increased and the number density decreased. The size of the dislocation loops of the two coatings was similar, while the number density of the dislocation loops of NbMoTaW coating was lower than that of VCrTaW coating. Both HEA coatings were significantly hardened after irradiation and the hardening rate of NbMoTaW coating was lower than that of VCrTaW coating. Interestingly, the hardening rate of 50 dpa was higher than 100 dpa for both HEA coatings, which was determined by softening factor and hardening factor in the irradiation process. Based on the defects and hardening rate caused by irradiation, it was preliminarily concluded that NbMoTaW coating had better irradiation resistance, and the above results provide an important reference for the subsequent optimization of lead-based fast reactors structural materials.

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&lt;111&gt; and &lt;100&gt; 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.

Microstructural Characterization of Reactor Pressure Vessel Steels

Ion irradiation is a promising tool to emulate neutron-irradiation effects on reactor pressure vessel (RPV) steels, especially in the situation of limited availability of suitable neutron-irradiated material. This approach requires the consideration of ion-neutron transferability issues, which are addressed in the present study by comparing the effect of ions with neutron-irradiation effects reported for the same materials. The first part of the study covers a comprehensive characterization, based on dedicated electron microscopy techniques, of the selected unirradiated RPV materials, namely a base metal and a weld. The results obtained for the grain size, dislocation density, and precipitates are put in context in terms of hardening contributions and sink strength. The second part is focused on the depth-dependent characterization of the dislocation loops formed in ion-irradiated samples. This work is based on scanning transmission electron microscopy applied to cross-sectional samples prepared by the focused ion beam technique. A band-like arrangement of loops is observed in the depth range close to the peak of injected interstitials. Two levels of displacement damage, 0.1 and 1 dpa (displacements per atom), as well as post-irradiation annealed conditions, are included for both RPV materials. Compared with neutron irradiation, ion irradiation creates a similar average size but a higher number density of loops presumably due to the higher dose rate during ion irradiation.

Investigation of vacancy trapping by solutes during quenching in aluminum alloys

In aluminum alloys, excess vacancies often form pairs or clusters with solute atoms. This is commonly referred to as vacancy trapping by solutes. Such trapping results in a change in the number of mobile or free vacancies, which in turn affects solute diffusion and precipitate formation during heat treatment. In the current work, such trapping is investigated indirectly in the Al-Sn binary model alloy via studying the dependence of Frank loop formation on the availability of free vacancies. Current experimental results show that the addition of 100 at. ppm. Sn to aluminum leads to a significant reduction in the number density of quenched-in Frank loops. Modeling of mono- and divacancy trapping by solutes during quenching demonstrates the influences of the vacancy-solute binding energy, solute concentration, and temperature on the trapping process. The results reveal that the binding between vacancies and solutes significantly reduces the concentration of free vacancies after quenching in Al-Sn.

Microstructure Evolution and Effect on Deuterium Retention in TiC- and ZrC-Doped Tungsten under He+ Ion Irradiation

Combining the advantages of a wet chemical method and spark plasma sintering, carbide-doped materials W-1wt%TiC and W-1wt%ZrC were prepared. Microstructural evolution in W-1wt%TiC and W-1wt%ZrC under irradiation of 5 keV He+ at 600 °C to fluences up to 5.0 × 1021 ions/m2 with ion flux of about 8.8 × 1017 ions/m2s was investigated by transmission electron microscopy (TEM). The dislocation loop number density of W-1wt%TiC was higher than that of W-1wt%ZrC, but the average loop size of the W-1wt%TiC was in average smaller. There were no observable helium bubbles in W-1wt%TiC and W-1wt%ZrC, exhibiting higher radiation resistance to He+ compared to pure W. He+ pre-damaged and undamaged W-1wt%TiC and W-1wt%ZrC samples were irradiated by 5 keV D2+ to estimate the D retention in doped W materials. The irradiation damage impact of He+ on deuterium retention was examined by a method of thermal desorption spectroscopy (TDS). Compared with the undamaged samples, it was illustrated that D2 retention of W-1wt%TiC and W-1wt%ZrC increased after He+ pre-irradiation.

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.

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.

Gamma-ray irradiation induced dislocation loops in hexagonal zirconium

ABSTRACT With doses simulating those received by spent nuclear fuel (SNF) facilities, gamma-ray (γ-ray) irradiation experiments were conducted on hexagonal zirconium. γ-ray irradiation–induced crystalline defects were observed in the hexagonal zirconium and characterized. Monte Carlo calculations combined with Transmission Electron Microscopy characterizations revealed that the energy of primary knock-on atoms could reach the threshold energy () of zirconium atoms and result in dislocation loops within the α-zirconium grains. The influence of γ dose on dislocation loop size and number density was also evaluated.

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.