This study investigates the pivotal role of rare earth lanthanum (La) in mitigating helium (He)-induced degradation in V-4Cr-4Ti alloys, a leading candidate for future fusion reactor structural materials. Through systematic helium ion irradiation experiments at 550°C across multiple doses, we reveal how La micro-alloying reshapes helium bubble evolution and enhances radiation resistance. By integrating aberration-corrected transmission electron microscopy (AC-TEM), geometric phase analysis (GPA), and atomic-resolution electron energy loss spectroscopy (EELS), we achieve unprecedented insights into helium bubble morphology, stress fields, and helium-to-vacancy (He/V) ratios validated against classical equations of state (EOS). Key findings include: 1. Swelling Suppression: La addition reduces swelling rates to 53% in V-4Cr-4Ti-1.0La compared to the V-4Cr-4Ti alloy at 10,000 appm He, attributed to elevated He/V ratios in bubbles that limit volumetric expansion. 2. Hidden Microstructural Drivers: Less than 29% of implanted He resides in TEM-visible bubbles, and the He in the invisible clusters dominate the alloy’s exceptional swelling resistance while contributing minimally (<25%) of He bubbles to irradiation hardening via the dispersion barrier hardening (DBH) model. 3. Stress Field Modulation: GPA quantifies compressive stress gradients around faceted bubbles, linking bubble geometry to localized strain accommodation mechanisms. This work challenges conventional paradigms by demonstrating that macroscopic swelling and hardening are governed by atomic-scale He-vacancy interactions rather than visible bubble populations. The findings provide a transformative strategy-rare earth micro-alloying in vanadium alloys for the development of durable structural materials for commercial fusion reactors.

Irradiation damage can substantially modify the corrosion response of nuclear structural materials, particularly when transmutation helium is involved. In this research, alumina-forming austenitic stainless steel (AFA steel) was irradiated at 600 °C using either single Fe-ion irradiation or Fe+He dual-beam irradiation, reaching a damage level of 15 dpa and a He concentration of 750 appm, and was then exposed to supercritical CO2 (S-CO2) at 600 °C and 15 MPa for 50 h. The results showed that Fe+He dual-beam irradiation accelerated corrosion much more strongly than single Fe-ion irradiation. Compared with the unirradiated AFA steel, the oxide thickness in grain interiors of the AFA steel increased by a factor of 2.6 after Fe-ion irradiation and by a factor of 28 after Fe+He dual-beam irradiation. Meanwhile, the oxide scale evolved from a simple dual-layer structure to a complex five-layer structure. This transition was attributed to the different irradiation-induced defect structure: single Fe-ion irradiation primarily increased the dislocation density, whereas Fe+He dual-beam irradiation produced a high density of cavities that promoted fast diffusion and enhanced oxidation. In contrast to the grain interiors, the examined grain boundaries (GBs) region exhibited better irradiation-corrosion tolerance after Fe+He dual-beam irradiation. This improved corrosion resistance was associated with defect absorption at GBs and Al supplied from NiAl precipitates located at GBs, which promoted the formation of a protective Al2O3 scale. These findings clarify the role of He-assisted defect evolution in controlling corrosion of AFA steel in S-CO2.

An analytical model is presented that characterizes the evolution of irradiation-induced dislocation loops in ferritic/martensitic (F/M) steels. The model establishes quantitative relationships among microstructural changes, irradiation conditions, and mechanical properties, predicting both dislocation loop behavior and irradiation hardening. Analysis indicates that the density of a/2<111> loops decreases with increasing temperature over 250 degrees C, whereas a<100> loops peak at approximately 450 degrees C before declining. Both loop types exhibit saturation at high irradiation doses, and their size distributions are well described by a log-normal function. An L-shaped phase diagram identifies a transition region between 350-400 degrees C, reflecting a shift in the dominant dislocation loop type. The model's explicit functional relationships provide a basis for optimizing the design and performance of F/M steels in advanced nuclear systems, potentially contributing to enhanced irradiation resistance and material reliability.

For advanced liquid metal-cooled nuclear reactors, one of the primary materials challenges is the degradation of structural component performance caused by high-energy neutrons. This degradation is substantially exacerbated by helium generation via nuclear transmutation reactions. While silicon addition to martensitic steels represents an established strategy for mitigating liquid metal corrosion, its multifaceted influence on the fundamental mechanisms of irradiation damage-encompassing defect evolution, microstructural stability, and helium accumulation-necessitates systematic investigation to ensure long-term reliability and safety. In light of these factors, four series of Fe9Cr1.5 W martensitic steels containing silicon at 0, 0.4, 0.7, and 1.0 wt.% was designed, and were performed dual-beam irradiation experiments with helium injection concentrations of 0,5 and 100 appm/dpa at 550 degrees C. After irradiation, transmission electron microscopy was employed to characterize the dislocation loops and cavities in the irradiated materials, while nanoindentation was used to evaluate the irradiation-induced hardening behavior. Under various dual beam irradiation, the size of irradiation-induced dislocation loops increases with rising silicon content, while the numerical density exhibits the opposite trend. Additionally, at high helium concentrations, silicon addition significantly reduced irradiation swelling. Specimens containing 0.4 wt.% silicon exhibited the lowest average cavity size and number density. As silicon content further increased, changes in irradiation cavities and number density tended toward saturation. Simultaneously, specimens with 0.4 wt.% silicon demonstrated superior resistance to irradiation hardening under all irradiation conditions.

The disparity in dose rates between ion irradiation and neutron irradiation results in distinct microscopic defect structures in irradiated materials. Moreover, the hydrogen-helium synergistic effect further complicates the evolution of the defects. Therefore, these two effects are the two most of concerning issues in simulating neutron irradiation using ion irradiation. In order investigate the synergistic effect between dose rate and hydrogen-helium gas atom on the evolution of dislocation loops and material hardening, both single- and multi-beam irradiation schemes with a dose rate difference of 10 times were implemented on China Low-Activation Martensitic (CLAM) steel. It was found that as the dose rate increased, the average size of the dislocation loops became smaller while the number density elevated, and the irradiation-induced hardening was more pronounced, under both single- and multi- beam irradiations. Interestingly, it was found that the presence of hydrogen and helium significantly amplified the difference in dislocation loops and hardening caused by different dose rates. A mechanism based on rate theory and dislocation loop growth equations was proposed to elaborate this dose rate effect. Moreover, a clear physical picture of how dose rate, sink strength, and hydrogen-helium gas atoms collectively affect the evolution of dislocation loops has been raised. The present study provides a valuable reference for the evaluation of candidate structural materials of future fusion reactors using ion irradiation.

China Low-Activation Ferrite (CLF-1) steel, renowned for its excellent thermomechanical properties and irradiation resistance, plays a key role in the development of the R&D of the Chinese Helium-Cooled Ceramic Breeding Test Blanket Module. Cold-worked CLF-1 steels were irradiated with sequential dual ion beams of (Fe2+ and H+), followed by single He+ irradiation at 723 K, with a dose rate of 1.09 dpa/h, to explore the complex relationship between cold work, defect evolution, and irradiation hardening. Samples with cold-working deformations of 0%, 10%, and 50% (denoted as CW 0%, CW 10%, and CW 50%, respectively) were examined. The results based on nanoindentation, TEM, and EBSD reveal that moderate cold work (10%) introduces dense dislocations, acting as effective sinks to suppress irradiation-induced defect accumulation and hardening, while excessive cold work (50%) triggers partial recrystallization under relatively long-time multi-beam irradiation, reducing dislocation density, which leads to the comparable hardening with CW 10%. In contrast, non-deformed samples (0% cold work) exhibit severe irradiation hardening (38.46%). He bubbles and dislocation loops follow non-monotonic trends in number density (CW 50% < CW 0% < CW 10%) and size (CW 50% > CW 0% > CW 10%), governed by the interplay of sink efficiency, thermal diffusion, and recrystallization. These findings highlight that a moderate level of cold-working deformation contributes to enhancing the sink strength, thereby offering a viable approach for designing radiation-tolerant RAFM steels.

After being self-ion implanted at 400 °C and 550 °C, the microstructure and irradiation hardness of kilogram-scale V-4Cr-4Ti alloys were studied using a transmission electron microscope (TEM) and nano-indentation test technology. Irradiation was performed using self-ions (V2+) at 2.5 MeV with design influences of 1.15 × 1015 ions cm−2, 4.59 × 1015 ions cm−2, and 9.17 × 1015 ions cm−2, so that the peak damages of V-4Cr-4Ti alloys are 1, 4, and 8 dpa, respectively. Compared with the 400 °C samples, the 550 °C samples exhibited a higher-density number of dislocation loops and increased hardness and reached saturation at lower irradiation doses. The irradiation temperature was mainly responsible for these differences, and the potential mechanism for its effect on the irradiation behavior was discussed.

High energy neutron irradiation in fusion reactors compared to fission reactors necessitates more attention to the effects of hydrogen (H) and helium (He) generated by transmutation on structural materials. Due to the intricate synergistic effect between H and He, the experimental results have not exhibited satisfactory consistency and, sometimes were contradictory. Via summarizing the previous results, it is suggested that the mechanism of the HHe synergistic effect is highly correlated with the H/He ratios. Based on this perspective, multi-beam ion irradiation experiments with four distinct H/He ratios were implemented on China Low Activation Martensitic (CLAM) steel, a candidate structural material for fusion reactors. The results showed that a low H/He ratio contributed to the dispersion of cluster nucleation, resulting in high-density small dislocation loops, whereas a high H/He ratio promoted the growth of defect clusters. Irradiation-induced hardening was observed to increase at low H concentrations and decrease at high H concentrations. The present study proposed that the H-He synergistic effect can be viewed from the perspective of the formation of H2 molecules. The disparate responses of materials to the H-He synergistic effect may originate from the different H critical concentrations for the formation of H2 molecules. This critical concentration should be determined by the factors such as material composition and irradiation conditions.

Due to its excellent high-temperature stability and radiation resistance, high-entropy alloys (HEAs) have attracted wide attention as candidate materials for advanced nuclear reactor systems recently. To understand their microstructure evolution and mechanical performance at high irradiation temperatures, NbMoTaW coatings were irradiated with 2.7 MeV Si2+ ions to a high peak damage dose of 50 displacements per atom at 450, 550 and 600 degrees C. Transmission Electron Microscopy and nanoindentation were used to characterize the defects and irradiation-induced hardening. No voids were observed and the grain structure kept stable after irradiated. As the temperature increases, dislocation lines become more distorted and harder to slip, contributing to hardening in high-temperature irradiated samples. The average size of dislocation loops increases slightly but remains approximately 9 nm, and the density of dislocation loops exhibits a fluctuating tendency. The pre-existed microcracks near grain boundaries became narrower with the elevation of temperature and disappeared at 600 degrees C. It can be attributed to the larger swelling at higher temperature, where the invisible vacancies could squeeze these microcracks. These vacancies should be one of the key factor leading to the highest irradiation-induced hardening occurred at 600 degrees C. In addition, irradiation segregations were observed at grain boundaries at 600 degrees C.

Worldwide efforts were concentrated to obtain accident tolerant fuel (ATF) cladding to replace conventional zirconium alloy cladding after the Fukushima Daiichi accident, and iron-chromium-aluminum (FeCrAl) alloys were considered to be one of the most promising materials for future applications. The effect of thermal aging on the irradiation behavior of FeCrAl alloys was investigated to further enhance performance in present study. FeCrAl samples that had been aged at 400 degrees C for 0, 1000, 3000, and 5000 hours, respectively, were irradiated with 2.5 MeV Fe2+to a dose of 10 dpa at 360 degrees C. Microstructural changes were analyzed with transmission electron microscopy (TEM), and hardness and irradiation hardening were measured by nano-indentation. Experimental results showed that dislocation loops were induced in all irradiated samples, with thermally aged samples having larger average loop size and lower number density compared to the non-aged sample. The average loop size gradually increased and number density decreased with extended aging, though the density almost kept unchanging between the 3000 h and 5000 h samples. Thermal aging increased hardness before irradiation, with the 0 h sample showing the lowest initial hardness. The hardness of 0 h sample showed a significantly huge increase after irradiation, while thermally aged samples showed smaller changes, suggesting that thermal aging can effectively mitigate irradiation hardening. In the present work, Fe-13Cr-4.5Al exhibited optimum properties after 1000 h of thermal aging. These results indicate that thermal aging positively affects the properties of FeCrAl alloys, and mechanical properties and irradiation resistance can be improved by appropriately optimizing thermal aging time.

Si2+ irradiation with the dose of 33.8–100 dpa was used to study the dislocation loops and hardening in W and W-3Re alloy at 400 °C and 550 °C using transmission electron microscope (TEM) and nano-indentation tests. When irradiated at 550 °C, larger dislocation loops and numerous dislocation network formed with the dose increased from 33.8 dpa to 100 dpa for both materials. The density of dislocation loops might be saturated between 33.8–100 dpa for W and saturated at ∼53.8 dpa for W-3Re alloy. Compared with pure W, the mean size of dislocation loop was smaller in W-3Re alloy when irradiated to 33.8 dpa for both temperatures. Meanwhile, it was found that the irradiation hardening rate of W was larger than that of W-3Re alloy under the same irradiation conditions, indicating that the addition of 3 wt.% Re element could indeed decrease the irradiation hardening rate of W. The hardening rate might be saturated after ∼53.8 dpa irradiation for W-3Re alloy, which might be related to the saturated loop densities.

V-4Cr-4Ti alloy is one of the candidate structural materials for future fusion reactors due to its desirable characteristics. In our previous research, MAX-phase-dispersion-strengthened vanadium alloy (V-4Cr-4Ti-1.5Y-0.3Ti3SiC2), prepared through mechanical alloying, showed excellent thermal stability and creep resistance and was expected to have good radiation resistance. This study investigates the effects of 2.5 MeV V2+ ion irradiation on V-4Cr-4Ti-1.5Y-0.3Ti3SiC2 and V-4Cr-4Ti alloys at 500 °C, with peak damage of 0.8, 3.5, and 6.1 dpa. Transmission electron microscopy and nanoindentation were used to examine the changes in microstructure and hardness before and after irradiation. The microscopic analysis reveals that dispersed nanoparticles maintained good stability under irradiation. Defect clusters grow with increasing irradiation doses in both materials. The nanoindentation results show that V-4Cr-4Ti-1.5Y-0.3Ti3SiC2 has higher initial hardness and lower irradiation hardening, indicating better resistance to radiation hardening than V-4Cr-4Ti. This research serves as a valuable reference for the assessment of the irradiation resistance of Ti3SiC2-dispersion-strengthened V-4Cr-4Ti alloy.

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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.

Four kinds of V-4Cr-4Ti-xLa (nominal composition x = 0, 0.1, 0.5, 1.0 wt.%) alloys with different La contents were prepared by vacuum arc melting. Three of them (x = 0, 0.1, 1.0 wt.%) were selected for single He+ (2000 appm for peak concentration) irradiation and sequential He+/V2+ (10 dpa for peak damage) irradiation at 550 C. The results showed that the addition of La element could refine the grains by forming the second phase La2O3 particles. Moreover, the addition of La element reduced the sum of impurity elements concentration in the alloy, which inhibited the formation of Ti-rich precipitates and affected the vacancy concentration, thus affecting the evolution of He bubbles under different irradiation types. In addition, the V-4Cr-4Ti-1.0La alloy had a lower irradiation hardening after sequential He+/V(2+ )irradiation, which might be related to the removal of impurity elements. This work provided a reference for the performance optimization of vanadium alloy for fusion reactors.

The increasing complexity and high-dimensional nature of real-world optimization problems necessitate the development of advanced optimization algorithms. Traditional Particle Swarm Optimization (PSO) often faces challenges such as local optima entrapment and slow convergence, limiting its effectiveness in complex tasks. This paper introduces a novel Hybrid Strategy Particle Swarm Optimization (HSPSO) algorithm, which integrates adaptive weight adjustment, reverse learning, Cauchy mutation, and the Hook-Jeeves strategy to enhance both global and local search capabilities. HSPSO is evaluated using CEC-2005 and CEC-2014 benchmark functions, demonstrating superior performance over standard PSO, Dynamic Adaptive Inertia Weight PSO (DAIW-PSO), Hummingbird Flight patterns PSO (HBF-PSO), Butterfly Optimization Algorithm (BOA), Ant Colony Optimization (ACO), and Firefly Algorithm (FA). Experimental results show that HSPSO achieves optimal results in terms of best fitness, average fitness, and stability. Additionally, HSPSO is applied to feature selection for the UCI Arrhythmia dataset, resulting in a high-accuracy classification model that outperforms traditional methods. These findings establish HSPSO as an effective solution for complex optimization and feature selection tasks.

Qxide-dispersion-strengthened (ODS) reduced-activation ferritic/martensitic (RAFM) steels are considered as most promising structural materials for fusion reactors and other advanced nuclear systems due to its excellent irradiation resistance under high dose irradiations. However, the extreme complex synergistic effect between transmuted helium/hydrogen and displacement damage which most seriously affects the structure and mechanical properties of ODS-RAFM steels is very difficult to understand, especially the role of hydrogen still remains unclear. Taking advantage of very sensitive to small vacancy clusters of Positron annihilation spectroscopy (PAS) method, we conducted irradiation defect study on ODS-RAFM, RAFM and its corresponding model alloy Fe-9Cr and α-Fe to investigate their anti-swelling properties at the early stage of irradiation with hydrogen-helium synergism and the mechanism. Three distinct irradiation schemes including (1) single Fe ion beam, (2) simultaneous Fe and He (named Fe+He) ion beam, (3) subsequent H ion injection after Fe+He (named Fe+He/H) ion beam irradiation were performed. Hydrogen-helium synergistic effects on vacancy evolution within these steels were explored combining the first-principles calculation. The implantation of He was observed to significantly increase defect concentration among all four steels, while H exhibited interestingly complexity on affecting the evolution of helium bubbles. The H post-injection increased defect concentration obviously in the materials with abundant sinks, but improved poorly in ones with rare sinks. Through the systematic study of the interactions between H, He, and vacancies, it was revealed that even at the early irradiation stage, ODS-RAFM steels has already exhibited excellent irradiation resistance compared to other alloys across three irradiation schemes.

The development of human society has significantly impacted the natural habitats of certain marine animals, pushing many marine mammals to the brink of extinction. To protect and observe endangered animal populations, this article proposes a method for identifying marine mammal dorsal fins. The method utilizes a variable convolutional neural network to recognize dorsal fins, enabling the identification and detection of exposed fin marine mammals, the practicality of this method has been validated in this article.The proposed method employs attention mechanisms and feature fusion to learn distinctive features in different regions. Additionally, modifications are made to the loss function and feature extraction methods. These adjustments facilitate better learning of dorsal fin characteristics, ultimately achieving recognition of the white dolphins.By leveraging this advanced technology, we hope to enhance the conservation of endangered marine species, creating better conditions for their survival and reproduction.

ODS ferritic/martensitic (ODS-FM) steel was irradiated up to 500 dpa at 550°C with 2.5 MeV Fe2+ to investigate the microstructure evolution under low to very high dose. Transmission electron microscopy (TEM) and Energy-dispersive X-ray spectroscopy (EDS) were conducted to characterize the microstructure and chemical composition. Only a few a0<100> dislocation loops were observed under low dose, which evolved into dislocation networks with the increase of dose. Voids were preferentially formed at oxides/matrix interface and some voids gradually accumulated on grain boundaries and dislocation lines with the increase of dose. The oxide particles were dissolved, re-precipitated and coarsened, forming high-density nano-oxide particles which greatly inhibit swelling under irradiation. Ti-rich and (Y, O)-rich shells of Y-Ti-O particles were formed successively during dissolution. And a large amount of Cr, Mn and O were segregated to the sample surface at high dose irradiation.

An amount of 100 dpa Si2+ irradiation was used to study the effect of transmutation rhenium content on irradiated microscopic defects and hardening in W-xRe (x = 0, 1, 3, 5 and 10 wt.%) alloys at 550 °C. The increase in Re content could significantly refine the grain in the W-xRe alloys, and no obvious surface topography change could be found after high-dose irradiation via the scanning electron microscope (SEM). The micro defects induced by high-dose irradiation in W and W-3Re alloys were observed using a transmission electron microscope (TEM). Dislocation loops with a size larger than 10 nm could be found in both W and W-3Re alloy, but the distribution of them was different. The distribution of the dislocation loops was more uniform in pure W, while they seemed to be clustered around some locations in W-3Re alloy. Voids (~2.4 nm) were observed in W-3Re alloy, while no void was investigated in W. High-dose irradiation induced obvious hardening with the hardening rate between 75% and 155% in all W-xRe alloys, but W-3Re alloy had the lowest hardening rate (75%). The main reasons might be related to the smallest grain size in W-3Re alloy, which suppressed the formation of defect clusters and induced smaller hardening than that in other samples.