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Surface dilation and helium bubbles corresponding to the fcc and bcc phases, (a) and (b), respectively. (c) Statistical data for the helium bubbles formed in the bcc and fcc phases-including the bubble size, bubble density, and bubble swelling.
Source publication
A cryo-quenched 70 wt % Fe-15 wt% Cr-15 wt% Ni single-crystal alloy with fcc (face centered cubic), bcc (body centered cubic), and hcp (hexagonal close packed) phases was implanted with 200 keV He+ ions up to 2 × 1017 ions·cm−2 at 773 K. Surface-relief features were observed subsequent to the He+ ion implantation, and transmission electron microsco...
Contexts in source publication
Context 1
... 9 shows the cross-section TEM sample from a surface relief area where a second phase is observed that is consistent with the surface "upheaval" or dilation feature. At higher magnification, Figure 10a,b show helium bubbles at the helium peak concentration region in the fcc and bcc phases. The helium bubbles are much larger but less dense in the bcc phase than in the fcc phase. ...
Context 2
... the peak helium concentration region, the helium bubbles exhibit the greatest variation in bubble size in the different phases. The plotted results are shown in Figure 10c. The average sizes of helium bubbles in the bcc phase and fcc phases are about ~18 and ~9 nm in diameter, respectively. ...
Context 3
... 9 shows the cross-section TEM sample from a surface relief area where a second phase is observed that is consistent with the surface "upheaval" or dilation feature. At higher magnification, Figure 10a,b show helium bubbles at the helium peak concentration region in the fcc and bcc phases. The helium bubbles are much larger but less dense in the bcc phase than in the fcc phase. ...
Context 4
... the peak helium concentration region, the helium bubbles exhibit the greatest variation in bubble size in the different phases. The plotted results are shown in Figure 10c. The average sizes of helium bubbles in the bcc phase and fcc phases are about ~18 and ~9 nm in diameter, respectively. ...
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Citations
... He bubbles in FeCrNi alloy subjected to He implantation seen as bright regions in TEM micrograph[41]. ...
The review starts by highlighting the significance of nuclear power plants in the contemporary world, especially its indispensable role in the global efforts to reduce CO 2 emissions. Then, it describes the impact of irradiation on the microstructure and mechanical properties of reactor structural materials. The main part provides the reader with a thorough overview of crystal plasticity models developed to address the irradiation effects so far. All three groups of the most important materials are included. Namely, the Zr alloys used for fuel cladding, austenitic stainless steels used for reactor internals, and ferritic steels used for reactor pressure vessels. Other materials, especially those considered for construction of future fission and fusion nuclear power plants, are also mentioned. The review also pays special attention to ion implantation and instrumented nanoindentation which are common ways to substitute costly and time-consuming neutron irradiation campaigns.
... ability to contain plasma by inverse or re-radiation due to excellent surface finish), mechanical and aforementioned properties. It is hypothesized that production of surface nanostructure (nanochannels) [83][84][85][86] will help escape of He [60, [87][88][89][90][91] inhibiting formation (nucleation and growth (crystallography) [90] or precipitation and coalescence (physical chemistry)) of bubble and restricts surface degradation. However, this is poorly understood and wrongfully approached. ...
... In physical chemistry sense, He atoms have strong repulsion to W atoms [80,92]. This ultra-low solubility forces He atoms to self-3 precipitate into small He bubbles [83] that become nucleation sites [90] for further void growth [93] under radiation induced vacancy supersaturations [94], resulting in material swelling [69,86,95] and high temperature He embrittlement [71,96,97], as well as surface blistering [75][76][77][78] under low energy and high flux He bombardment [54,98] at elevated temperatures [99]. This may be mitigated by engineering structures in material which help in outgassing of He. ...
Bubble (point defect) – a precursor of fuzz or under dense nanostructure formation is crystal lattice defect. Suitable selection of crystal lattice which inhibit Frenkel pair generation and intrinsically promotes selfinterstitial solid solution strengthening contributes effectively towards making plasma facing material. For this, interstitial sites, their size, amount / fraction, positions, tendency of occupation and diffusion parameters (e.g. activation energies (Q), activation volumes) are determined. Fcc iron carbon alloys (austenitic stainless steels AISI / SAE 321, fcc structure, Pearson code cF4, space group Fm3̅m) are proposed as suitable candidates. Along with their room temperature fcc structure having 12 interstitial positions (4 octahedral, 6 coordination sites and 8 tetrahedral, 4 coordination sites / unit cell) to allow insertion of self (iron) atoms, they have excellent corrosion resistance, thermal conductivity, and nonmagnetic properties. After their melting, casting, and machining to required dimensions and geometry, stabilizing heat treatment is applied to precipitate all carbon as TiC and prevent formation of Cr23C6 (sensitization). This resist heat and surface degradation and yield excellent architecture which not only inhibit Frankel pair generation but will also allow bulk assimilation or surface annihilation (loop punching) of this lattice point defect. A superior thermal, fluid, and structural design augment above
... 6 However, it is known to experience significant radiation induced swelling due to its FCC (Face-Centered Cubic) crystal lattice structure. 7 Also, significant activation occurs due to the Ni content compared to ferritic martensitic counterparts. 8 However, materials selection is always a compromise between different factors and one must consider multiple requirements and aspects to find the best materials compromise. ...
The effect of thermal oxide layer on He implanted 316L stainless steel was studied to evaluate experimentally how thermal oxidation affects the diffusion and distribution of He in the material. In the case of thermal oxidation of a He implanted sample, with an increase in oxidation time, the max swelling height increases logarithmically as a function of time and finally saturates for all samples except for the lowest dose of implanted He. Concerning TEM results, two void regions are identified. Similar to the calculation, the total irradiated depth was around 250 nm and the large void region was formed around 100–150 nm depth. On the other hand, the small void region was observed immediately under oxide layer from the thermal oxidation. In contrast, there were no voids in the altered zone near the metal/oxide interface in the non-thermal oxidized/He implanted sample. This description of the phenomena was justified using the Kirkendall effect and the Point Defect Model.
... Helium bubble formation has been found to be affected not only by chemical complexity but also by crystalline structures. 153 Single-crystal Ni 15 Fe 70 Cr 15 fcc was cycled through cryogenic quenching at 77 K by immersion in liquid nitrogen and then heated to room temperature. Partial martensitic transformation of the fcc to the bcc and hcp phases occurred, leading to three coexisting phases: fcc (γ-austenite), bcc (α'-martensite), and hcp (ε-martensite). ...
... The dual phases, or coexistence of multiple phases, is considered to be beneficial to the overcome strength−ductility trade-offductility loss resulting from metallurgical treatment to increase strength 196 or improve radiation performance. 150,153 A study of Ni 15 Fe 70 Cr 15 showed phase boundaries in this transformed alloy contained misfit dislocations, 153 which can serve as He traps that disperse the He bubble distribution and may be considered a desirable property because they increase the swelling resistance. Another study has shown that atoms in ductile fcc nanostructured NiFeCoCrCu rearranged into a hard bcc phase under room-temperature irradiation and formed alternating layers of the two phases. ...
... The dual phases, or coexistence of multiple phases, is considered to be beneficial to the overcome strength−ductility trade-offductility loss resulting from metallurgical treatment to increase strength 196 or improve radiation performance. 150,153 A study of Ni 15 Fe 70 Cr 15 showed phase boundaries in this transformed alloy contained misfit dislocations, 153 which can serve as He traps that disperse the He bubble distribution and may be considered a desirable property because they increase the swelling resistance. Another study has shown that atoms in ductile fcc nanostructured NiFeCoCrCu rearranged into a hard bcc phase under room-temperature irradiation and formed alternating layers of the two phases. ...
The development of advanced structural alloys with performance meeting the requirements of extreme environments in nuclear reactors has been long pursued. In the long history of alloy development, the search for metallic alloys with improved radiation tolerance or increased structural strength has relied on either incorporating alloying elements at low concentrations to synthesize so-called dilute alloys or incorporating nanoscale features to mitigate defects. In contrast to traditional approaches, recent success in synthesizing multicomponent concentrated solid-solution alloys (CSAs), including medium-entropy and high-entropy alloys, has vastly expanded the compositional space for new alloy discovery. Their wide variety of elemental diversity enables tunable chemical disorder and sets CSAs apart from traditional dilute alloys. The tunable electronic structure critically lowers the effectiveness of energy dissipation via the electronic subsystem. The tunable chemical complexity also modifies the scattering mechanisms in the atomic subsystem that control energy transport through phonons. The level of chemical disorder depends substantively on the specific alloying elements, rather than the number of alloying elements, as the disorder does not monotonically increase with a higher number of alloying elements. To go beyond our knowledge based on conventional alloys and take advantage of property enhancement by tuning chemical disorder, this review highlights synergistic effects involving valence electrons and atomic-level and nanoscale inhomogeneity in CSAs composed of multiple transition metals. Understanding of the energy dissipation pathways, deformation tolerance, and structural stability of CSAs can proceed by exploiting the equilibrium and non-equilibrium defect processes at the electronic and atomic levels, with or without microstructural inhomogeneities at multiple length scales. Knowledge of tunable chemical disorder in CSAs may advance the understanding of the substantial modifications in element-specific alloy properties that effectively mitigate radiation damage and control a material’s response in extreme environments, as well as overcome strength–ductility trade-offs and provide overarching design strategies for structural alloys.
... ability to contain plasma by inverse or re-radiation due to excellent surface finish), mechanical and aforementioned properties. It is hypothesized that production of surface nanostructure (nanochannels) [83][84][85][86] will help escape of He [60, [87][88][89][90][91] inhibiting formation (nucleation and growth (crystallography) [90] or precipitation and coalescence (physical chemistry)) of bubble and restricts surface degradation. However, this is poorly understood and wrongfully approached. ...
... In physical chemistry sense, He atoms have strong repulsion to W atoms [80,92]. This ultra-low solubility forces He atoms to self-precipitate into small He bubbles [83] that become nucleation sites [90] for further void growth [93] under radiation induced vacancy supersaturations [94], resulting in material swelling [69,86,95] and high temperature He embrittlement [71,96,97], as well as surface blistering [75][76][77][78] under low energy and high flux He bombardment [54,98] at elevated temperatures [99]. This may be mitigated by engineering structures in material which help in outgassing of He. ...
Bubble (point defect) – a precursor of fuzz or under dense nanostructure formation is crystal lattice defect. Suitable selection of crystal lattice which inhibit Frenkel pair generation and intrinsically promotes selfinterstitial solid solution strengthening contributes effectively towards making plasma facing material. For this, interstitial sites, their size, amount / fraction, positions, tendency of occupation and diffusion parameters (e.g. activation energies (Q), activation volumes) are determined. Fcc iron carbon alloys (austenitic stainless steels AISI / SAE 321, fcc structure, Pearson code cF4, space group Fm3̅m) are proposed as suitable candidates. Along with their room temperature fcc structure having 12 interstitial positions (4 octahedral, 6 coordination sites and 8 tetrahedral, 4 coordination sites / unit cell) to allow insertion of self (iron) atoms, they have excellent corrosion resistance, thermal conductivity, and nonmagnetic properties. After their melting, casting, and machining to required dimensions and geometry, stabilizing heat treatment is applied to precipitate all carbon as TiC and prevent formation of Cr23C6 (sensitization). This resist heat and surface degradation and yield excellent architecture which not only inhibit Frankel pair generation but will also allow bulk assimilation or surface annihilation (loop punching) of this lattice point defect. A superior thermal, fluid, and structural design augment above
... The current understanding of He evolution in solids is well presented in the review "Radiation-Induced Helium Bubbles in Metals" [12]. The nuanced importance of composition and nanostructure for both fusion and Generation IV fission relevant materials are highlighted in the papers by El Atwani et al. exploring equiaxed nanocrystalline W and ultrafine grained W-TiC Alloy [13], by Kim et al.'s study of dual-phase 12Cr oxide-dispersion-strengthened alloy [14], and by Zhang et al. examining swelling and He bubble morphology in a cryogenically treated FeCrNi alloy with martensitic transformation and reversion after He implantation [15]. Two studies in model metal systems (Cu and Pd) noted the impact of radiation environments by comparing various sequential and concurrent heavy ion irradiation and He implantation conditions and by controlling the irradiation temperature. ...
Despite its scarcity in terrestrial life, helium effects on microstructure evolution and thermo-mechanical properties can have a significant impact on the operation and lifetime of applications, including: advanced structural steels in fast fission reactors, plasma facing and structural materials in fusion devices, spallation neutron target designs, energetic alpha emissions in actinides, helium precipitation in tritium-containing materials, and nuclear waste materials. The small size of a helium atom combined with its near insolubility in almost every solid makes the helium–solid interaction extremely complex over multiple length and time scales. This Special Issue, “Radiation Damage in Materials—Helium Effects”, contains review articles and full-length papers on new irradiation material research activities and novel material ideas using experimental and/or modeling approaches. These studies elucidate the interactions of helium with various extreme environments and tailored nanostructures, as well as their impact on microstructural evolution and material properties.
The interfaces including grain boundaries and hetero‐phase interfaces could withstand the radiation damage via providing sites for trapping the defects. With the urgent demand for high radiation‐tolerance nanostructured materials, exploration of the structural properties, especially the role of the interfaces on the radiation response in the nanomaterials, is of significant essentialness in developing the design of new radiation‐resistant materials. Herein, the recent progress and development of two main kinds of interfaces, namely the grain boundaries in nanocrystalline and the hetero‐phase interfaces in nanolayered materials and nanocomposites, in nanostructured materials that are designed for radiation tolerance are summarized. In addition, the radiation response and resistant mechanism under irradiation conditions of the nanocrystalline materials like metals, single‐phase alloys, and ceramics, as well as the nanocomposites like nanolayered metals, nanolayered ceramics, metal‐carbon nanocomposite, and nanoparticle dispersed steels, are reviewed. Finally, challenges and perspectives are proposed to offer advice for the development of radiation resistance nanomaterials.
Helium bubble formation and swelling were systematically studied in Ni-based concentrated solid solution alloys containing different numbers and types of elements. Our microscopy analysis showed that although increasing the alloy chemical complexity helps suppress bubble formation in general, there is no monotonic relationship between the bubble growth rate and the number of alloying elements. Certain elements (e.g., Fe and Pd) are more effective in suppressing bubble growth than others (e.g., Cr and Mn). Atom probe tomography was applied to accurately measure elemental segregation around bubbles, revealing unique effects of certain alloying elements on vacancy migration towards bubbles. More specifically, the high vacancy mobility via Cr sites leads to a large vacancy flux and an increased bubble size, while the high degree of atomic size mismatch introduced by Pd helps deflect vacancy flow away from bubbles and decrease the amount of swelling. The effects identified in this study provide new strategies to design concentrated solid solutions with superior resistance to swelling.
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.
