Evolution Of He Bubbles Research Articles

In-situ study on the differential evolution of He bubbles in the multilayer oxide of Fe9Cr1.5W0.4Si F/M steel corroded in lead-bismuth eutectic

The combined effects of corrosion and irradiation on nuclear components have been an important but not yet fully revealed topic. Here, the irradiation behavior of the oxide scale formed on F/M steel after lead-bismuth corrosion was in-situ investigated during He+ irradiation. The results showed that the oxide scale included a Fe3O4 outer oxide layer, a nanograin Fe(FexCr2-x)O4 spinel inner oxide layer, and an internal oxide layer. He bubbles formed in Fe3O4, Fe-Cr spinel and F/M steel were polygon, irregular elongated pores and small spheres, respectively. These differences were attributed to variations in defect generation, migration, and corrosion-induced crystal defects in different oxides. Numerous corrosion-induced nanograin boundaries and vacancies in Fe-Cr spinel exhibited more effective absorption of irradiation-induced defects. Moreover, rhombic perfect dislocation loops were detected in Fe3O4 at the late stage of irradiation, their relative positional relationship with He bubbles indicated a potential interaction between bubbles and loops.

Exploring the suppression methods of helium-induced damage in tungsten by investigating the interaction between beryllium and helium

Abstract Helium (He)-induced damage is a sensitive concern for the performance of tungsten plasma facing materials (W-PFMs). Recent experiments have revealed that trace impurities in He plasma can effectively prevent the formation of He bubbles and fuzz on W surfaces. To explore its plausibility and underlying mechanism, we performed a multiscale computational study that combines density functional theory calculations and object kinetic Monte Carlo simulations to investigate the effects of a small quantity of beryllium (Be) on the evolution of He bubbles. It is found that there is a strong attractive interaction between He and Be, which can be attributed to the decrease in electron density and the lattice distortion induced by embedded Be atoms. Therefore, the co-implantation of Be continuously introduces trapping centers for He. Due to the low implantation depth and high migration energy of Be, the Be atoms are located close to the surface, leading to the trapping of the majority of He within the near-surface region and the development of a shielding layer for He permeation. The presence of Be facilitates the dispersion of the trapped He, skewing the He clusters into smaller sizes. More importantly, the Be trapping centers bring the He clusters closer to the surface, significantly increasing the probability of bubble bursting and the release of He back to the vacuum. This ultimately leads to a lower retention of He in the case of He + Be co-irradiation, compared with the case of He-only irradiation. Consequently, our findings elucidate the suppressive effect of a low flux of Be atoms on the growth of He bubbles, highlighting the need to focus on synergetic effects between plasma species.

In-situ investigation of helium effect on diverse precipitate/matrix interface areas in irradiated 5052 aluminum alloy

Due to its low neutron cross-section and high irradiation swelling resistance, 5052 aluminum alloy is extensively utilized in the core structure and irradiation shielding of research nuclear reactors. Indeed, irradiation-induced defects may accumulate preferentially in the precipitates of aluminum alloys during service, leading to prior impairment of these precipitates, thereby inducing the degradation of mechanical properties. The interaction between irradiation defects and precipitates under irradiation, however, remains inadequately understood. In this work, the microstructure of 5052 aluminum alloy irradiated in-situ with He+ at different temperatures and subsequent annealing was investigated by transmission electron microscopy. After 1.0 dpa irradiation at different temperatures, the evolution of He bubbles within the precipitate was greater than in the matrix, and manifested a pronounced growth threshold between 373 K and 473 K. During post-irradiation annealing, compared to the matrix, also the dynamic evolution of bubbles was observed on the inner side of the precipitate interface, where bubbles coalesced and rapidly dissolved. The dynamic behavior of He bubbles within different precipitates manifested a certain discrepancy, and sink strength of two typical Al-containing precipitates was estimated. The difference in bubble evolution can be attributed to the different sink efficiency and vacancy concentration gradient around the precipitate.

The behavior of helium bubble evolution under neutron irradiation in different tungsten surfaces

Tungsten, as a kind of material for the plasma-facing wall in fusion reactor and tokamak, is subjected to irradiation by the high-energy neutron and the high fluxes of hydrogen and helium plasmas. Especially, in the presence of helium (He) bubbles, neutron irradiation can significantly exacerbate the deterioration of the material's mechanical properties. However, due to experimental limitations, the evolution of He bubbles in tungsten irradiated by neutrons and the underlying mechanisms remain unclear. Here, molecular dynamics (MD) simulation is used to investigate the behavior of He bubbles in tungsten with different crystal surfaces irradiated by neutrons. We observed the complete evolutionary process of the He bubble, i.e. standing, expanding, and bursting. During this process, high-energy neutron transfers energy to atoms within the He bubble through cascade collision, leading to a rapid increase in energy and pressure of the He bubble, and then the rapid expansion of the He bubble makes the pressure rapidly released, ultimately leading to the rupture of the helium bubble. Furthermore, we also explore the evolution of the He bubble under varying conditions (such as temperature, neutron energy, and He/vacancy ratio) within the same crystal surface. We found that at lower temperatures and neutron energy with a higher He/vacancy ratio, it tends to form pinhole-like channels to release helium atoms, resulting in the He bubble bursting. On the contrary, it tends to form a hill-like expansion to spurt helium atoms out, leading to the bubble bursting. Additionally, it was found that the direction and time of He bubble rupture varied under different crystal surfaces. He bubbles under the (112) surface burst the slowest with excellent corrosion resistance. These findings provide new insights into the behavior of irradiation resistance of fusion reactor materials under extreme conditions and offer significant guidance for the improvement and selection of optimal materials for plasma-facing walls.

Inhibition of Ti-rich precipitates and improvement of the irradiation resistance in V-4Cr-4Ti by adding La element

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+/V2+ 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 Mechanisms of Inhibition Effects on Bubble Growth in He-Irradiated 316L Stainless Steel Fabricated by Selective Laser Melting.

The AISI 316L austenitic stainless steel fabricated by selective laser melting (SLM) is considered to have great prospects for applications in nuclear systems. This study investigated the He-irradiation response of SLM 316L, and several possible reasons for the improved He-irradiation resistance of SLM 316L were systematically revealed and evaluated by using TEM and related techniques. The results show that the effects of unique sub-grain boundaries have primary contributions to the decreased bubble diameter in SLM 316L compared to that in the conventional 316L counterpart, while the effects of oxide particles on bubble growth are not the dominant factor in this study. Moreover, the He densities inside the bubbles were carefully measured using electron energy loss spectroscopy (EELS). The mechanism of stress-dominated He densities in bubbles was validated, and the corresponding reasons for the decrease in bubble diameter were freshly proposed in SLM 316L. These insights help to shed light on the evolution of He bubbles and contribute to the ongoing development of the steels fabricated by SLM for advanced nuclear applications.

Helium diffusion and bubble evolution in single-phase tungsten-based W-Ta-Cr-V complex concentrated alloy

Tungsten-based complex concentrated alloy (CCA) W-Ta-Cr-V films with bcc single-phase have been investigated for its exceptional radiation tolerance. This work aims to investigate the formation of helium (He) bubble and He retention behavior in the CCA film under 100 eV He plasma at a fluence of 2.02 × 1025 He/m2. The results show that the size of He bubbles in the CCA film is smaller than that in the pure W film, and the depth distribution is shallower, which indicates that the evolution of He bubbles is inhibited. As for the underlying mechanism of resistance to the formation of He bubbles in CCAs, it is believed that the characteristic of low He mobility reduces the diffusion depth of He, and the sluggish diffusion of He inhibits the aggregation of He atoms or He clusters to a certain extent. In other words, the unique energy and microstructure characteristics also inhibit the diffusion of He in the CCA film, which affects the nucleation and growth of He bubbles. Furthermore, the release of He is mainly from He adsorption of the near surface, and the desorption peaks shift to lower temperatures in the CCA film within the desorption temperature of 1273 K.

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

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

Effect of He content on the evolution of He bubbles and He thermal desorption behaviors in FeCr based nanocrystalline film

He-charged FeCr based nanocrystalline films were fabricated at room temperature in a mixed atmosphere of He and Ar by radio frequency magnetron sputtering. The effect of the He/Ar ratio on the crystalline structure, surface morphology, He desorption behaviors, and the size and distribution of He bubbles had been investigated by X-ray diffraction (XRD), scanning electron microscopy (SEM), thermal desorption spectrum (TDS), and transmission electron microscopy (TEM), respectively. The results of XRD revealed that He-charged FeCr based films had a typical bcc structure. Surface SEM images showed that the grain size of FeCr based films decreased with the He/Ar ratio increasing. The results of TDS profiles exhibited that these desorption peaks corresponded to the release of He atoms dissociated from surface, HenV (2≤n≤6), HenVm (1≤n, 2≤m), and He bubbles, respectively. Based on the results of TEM observation, the density of He bubbles was proportional to the He/Ar ratio, while the size of He bubbles could be scarcely influenced by the He/Ar ratio.

Panorama of “fuzz” growth on tungsten surface under He irradiation

Nanostructure formation on tungsten (W) surfaces under helium (He) irradiation is a unique and interesting phenomenon that directly impacts the performance and lifetime of W-based materials as plasma-facing materials (PFMs) in nuclear fusion reactors. The evolution of He bubbles beneath the W surface and its effect on surface morphology are essential for understanding the deterioration of the surface properties of PFMs. In this work, the morphology of the W surface under different irradiation conditions was investigated using molecular dynamics in order to extract essential factors that determine the nanostructure of the W surface. The design concerned with investigating the W surface morphology encompassed: (1) the morphology of the W-free surface without He bubbles vs. recoil energy; (2) the morphology of the W surface in the presence of He bubbles vs. recoil energy; and (3) sputtering vs. surface roughness in the absence of He bubbles. A mass of adatoms generated primarily by loop punching of sub-surface He bubbles slightly roughens the surface. Bubble bursting further heightens surface protrusions. In this case, the fiber-like structure of the surface is attributable to two processes involved in He bubble evolution: He diffusion plus aggregation and bubble expansion plus bursting. A high temperature window expedites He diffusion and aggregation beneath the protruded surface. Once the He bubbles burst regardless of spontaneous bursting (when the bubble pressure is beyond the critical value), or bursting due to external irradiation, the protrusions will be further heightened, thus serving as precursors of the fiber-like structure. Continuous aggregation, followed by bursting, should constitute the recursive process of protrusion upgrowth. The panoramic view of the fiber-like structure growth described here is expected to provide a clue for designing tungsten-based materials with high radiation-resistance performance.

Simulation study of helium bubble coalescence in tungsten at various temperatures relevant to fusion conditions

Molecular dynamics (MD) methods are used to study nanosized helium (He) bubble coalescence process in tungsten (W) at various temperatures relevant to fusion conditions, on an atomistic scale. Bubble coalescence in W is observed at a higher temperature and He/V ratio, while the calculated internal bubble pressure due to virial stress increases with the increase in the He/V ratio; bubble coalescence is significantly dependent on the bubble distance. In these MD simulations, coalescence occurs, only when the surface distance between the two bubbles is equal to 1a0, where a0 denotes the lattice constant and is approximately 0.317 nm at 2100 K. On the other hand, a bubble diameter between 1a0 and 3a0 may have relatively limited effect on the coalescence, although larger-sized bubbles may have higher migration energy. Local stress profile calculated indicates that initial bubbles can interact with each other, which may enhance the He atoms diffusion between bubbles and their coalescence. Physical contact at the initial stage of coalescence may occur between two nearby bubbles accompanied by W lattice distortion because of the limited displacement of W atoms near the bubbles and rapid migration of He atoms within the two bubbles. These results are beneficial for understanding the evolution of He bubbles in bulk W.

Diffusion controlled helium bubble formation resistance of FeCoNiCr high-entropy alloy in the half-melting temperature regime

For the structure materials applied in advanced nuclear energy system, helium bubble formation is always a big concern which will severely degrade the performance of materials around or above the half-melting temperature regime (∼0.5 Tm). To explore the He bubble formation resistance in the FeCoNiCr alloy, that is a novel face-centered cubic (fcc) high-entropy alloy (HEA) showing excellent radiation damage tolerance, we conducted a series of 2 MeV He ions irradiation experiments on them at three different temperatures (0.46, 0.51 and 0.57 Tm). For reference purpose, a model fcc metallic system of pure Ni was irradiated simultaneously. Through transmission electron microscopy (TEM), He bubble formation in the irradiated samples was systematically investigated. The results show that in any designated temperature, He bubbles have a smaller size, higher number density, and denser distribution in the HEA when comparing to that of pure Ni. The volume fraction of He bubbles is also less in the HEA, suggesting a suppressed bubble evolution. For the underlying mechanism of the He bubble formation resistance of HEA, we suggest that the featured energy barriers for point defects migration in the HEA will promote the recombination of defects and somewhat reduce the vacancy concentration during irradiation. Such unique effect could suppress the He diffusion through vacancy mechanism, it will finally influence the evolution of He bubbles in the HEA.

Radiation-Induced Helium Bubbles in Metals.

Helium (He) bubbles are typical radiation defects in structural materials in nuclear reactors after high dose energetic particle irradiation. In the past decades, extensive studies have been conducted to explore the dynamic evolution of He bubbles under various conditions and to investigate He-induced hardening and embrittlement. In this review, we summarize the current understanding of the behavior of He bubbles in metals; overview the mechanisms of He bubble nucleation, growth, and coarsening; introduce the latest methods of He control by using interfaces in nanocrystalline metals and metallic multilayers; analyze the effects of He bubbles on strength and ductility of metals; and point out some remaining questions related to He bubbles that are crucial for design of advanced radiation-tolerant materials.

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.

Evolution of helium bubbles below different tungsten surfaces under neutron irradiation and non-irradiation conditions

Studies of the evolution of embedded helium (He) bubbles in tungsten are very important for understanding the failure mechanism of tungsten as a plasma-facing material. In this work, the processes of He bubble evolution under neutron irradiation and non-irradiation conditions have been investigated by molecular dynamics simulations. The influences of temperature, helium/vacancy ratio, and surface crystal orientations on the stability and evolution of He bubbles were discussed. Four evolution cases (standing, expanding, bursting, and oversaturation expanding) of He bubbles under {1 0 0}, {1 1 0}, and {1 1 1} surfaces were found, and the oversaturation expanding case was discussed in detail. In particular, the differences between the stable bubbles bursting from {1 0 0}, {1 1 0}, and {1 1 1} surfaces under neutron irradiation condition were investigated. It was also found that He bubbles can burst from the surfaces (especially from a surface with low crystal face density) more easily under neutron irradiation condition.

Formation of bubbles and extended defects in He implanted (1 0 0) Si at elevated temperatures

The formation and thermal evolution of He bubbles and extended defects in Si are studied as a function of the implantation ( T i) and annealing ( T a) temperatures. (1 0 0) Si wafers were implanted with 40 keV He + ions to a fluence of 1 × 10 16 cm −2 at temperatures T i from 300 to 623 K. After the implantations the samples were submitted to thermal treatments for 600 s at temperatures T a from 673 to 1073 K. The samples were analyzed using transmission electron microscopy (TEM), Rutherford backscattering/channeling (RBS/C) and elastic recoil detection (ERD) techniques. The results obtained show that the microstructure characteristics of the bubble and the extended defect systems, as well as their thermal evolution behavior, depend significantly on the implantation temperature. A correlation between the dynamic annealing behavior and the bubble and defect formation process is proposed.

Radiation damage–He interaction in He implanted Si during bubble formation and their evolution in voids

He atoms were implanted in crystalline and pre-amorphized silicon wafers at doses in the 2×10 161×10 17 cm −2 range. Using transmission electron microscopy (TEM) we monitored the evolution of He bubbles into voids upon thermal annealing. Bubbles are formed in both crystalline and amorphous silicon. However, in amorphous material bubble interaction with the moving crystalline–amorphous interface during the epitaxial regrowth prevents their evolution into voids. By implanting He at different target temperatures in crystalline Si, thus by changing the structure of radiation damage, we found that the interaction between point defects and He atoms is essential for the generation of He bubbles and for their subsequent evolution into voids.

Response of helium bubbles in gold to displacement-cascade damage.

Evolution of He bubbles in Au has been followed with in situ electron microscopy during 400 keV Ar ion irradiation at 500 K. He bubbles were produced by implantation at 500 K of 3 keV He ions into prethinned Au samples that were then annealed to 670 K. During Ar irradiation, bubbles undergo athermal migration, coalescence, and disintegration. The bubbles after annealing were underpressurized, but Ar irradiation brought the bubbles to equilibrium pressure after a dose of 0.6 displacements per atom (dpa). Bubble behavior can be understood on the basis of their interactions with adjacent cascades. We propose that a discrete bubble jump is induced by the melt zone formed during the thermal spike phase of a contiguous cascade. The rate of bubble migration is modeled with the assumption that the jump distance is determined by the bubble and cascade volumes.