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Gadolinium titanate (Gd2Ti2O7, or GTO) and other lanthanide pyrochlores are interesting candidates for actinide waste disposal and fast ion conduction because the relevant material properties are intimately dependent on local cation structure. Therefore, a deep understanding of the kinetics associated with cation ordering and disordering is required if such material properties are to be tuned for specific device designs. To this end, single crystals of GTO were irradiated with 190 keV helium ions to a total fluence of 1 × 1017 ions/cm2, amorphizing the sample surface to a depth of ~1 μm and resulting in significant He bubble accumulation. FIB lamellae lifted out from the irradiated sections were examined during heat treatment in the (scanning) transmission electron microscope. Two distinct stages of the re-crystallization of the amorphized material were observed. The material near the end of the ions’ range transformed first and with the same orientation as the pristine material. This was due to the close proximity of the pristine material and the presence of small defect fluorite seeds, but the propagation of this growth front was frustrated by the large pores (formerly bubbles) in the He accumulation layer. This was followed by heterogeneous nucleation of new crystallites at random orientations at the top of the He accumulation layer, which is attributed to the high surface area associated with the many small He bubbles in that region. It is inferred that the kinetics of grain growth in this material are significantly faster than the kinetics of grain nucleation.

Refining grain size and adding alloying elements are two complementary approaches for enhancing the radiation tolerance of existing nuclear materials. Here, we present detailed in-situ irradiation research on defect evolution behavior and irradiation tolerance of ultrafine W-TiC alloys (thin foils) irradiated with 1 MeV Kr+2 at RT and 1073 K, and compare their overall performance to pure coarse grained tungsten. Loop Burgers vector was studied confirming the presence of loops whose population increased at high temperature. Loop density, average loop area, and overall damage are reported as a function of irradiation dose revealing distinct defect evolution behavior from pure materials. The overall damage generally followed the average loop size trend, which decreased with time for both temperatures, but was higher at 1073 K and attributed to biased vacancy sink behavior of the TiC dispersoids evidenced by large vacancy clusters on their interfaces. By comparison, the overall loop and void damage in pure tungsten was larger by a factor of six and two, respectively. The improved irradiation damage resistance in the alloys is thus attributed to the effect of dispersoids in 1) the enhancement in annihilating defects and mutual defect recombination due to both dispersoids and a higher grain boundary density; 2) decreasing the loop mobility, causing shrinkage and annihilation of loop density, which was confirmed via in-situ video. Several mechanisms are illustrated to describe the performance of the complex alloy system. The results motivate further experimental and modeling research that aims to understand the many different phenomena occurring at different time scales.

Nanocrystalline metals are often postulated as irradiation tolerant materials due to higher grain boundary densities. The efficiency of these materials in mitigating irradiation damage is still under investigation. Here, we present an in-situ transmission electron microscopy with ion irradiation study on equiaxed 35 nm grained tungsten (NCW-35 nm) and compare its radiation tolerance, in terms of dislocation loop damage, to several other grades of tungsten with different grain sizes at two temperatures (RT and 1073 K). The NCW-35 nm was shown to possess significant higher radiation tolerance in terms of loop damage. As demonstrated by Kinetic Monte Carlo simulations, at least part of the higher radiation tolerance of the small grains is due to higher interstitial storage (at the grain boundaries) and defect recombination (in the grain interiors) in the small grain material. In addition, experimental observations reveal rapid and efficient dislocation loop absorption by the grain boundaries and this is considered the dominant factor for mass transport to the boundaries during irradiation, enabling the remarkable radiation tolerance of 35 nm grained tungsten. This study demonstrates the possibility of attaining high radiation tolerant materials, in terms of dislocation loop damage, by minimizing grain sizes in the nanocrystalline regime.

Understanding the structure-property relationship for materials destined for irradiation extremes is a key step in developing materials with reliable, long-term performance. One crucial relationship is the ability of a material to withstand or accommodate amorphization, as this dictates its potential use as a nuclear waste form. Pyrochlores are one such class of materials for consideration as waste forms and there has been significant work examining how both the crystal structure and chemistry impacts amorphization resistance, leading to the important conclusion that the amorphization resistance of pyrochlores (A2B2O7) is very sensitive to the nature of the B cations. For example, pyrochlores with B Ti amorphize much more readily than B Zr compounds. However, there are still questions regarding how these types of materials respond to prolonged or high-dose irradiation conditions. In this work, Gd2Ti2O7 (GTO) and Gd2Zr2O7 (GZO) pyrochlores were implanted with 400 keV Kr++ ions at room temperature to calculated peak damages of 119 and 135 displacements per atom (dpa), respectively. As expected, GTO amorphized completely under irradiation. However, discrete bubbles of Kr coalesced within the amorphous matrix without micro-cracking or spallation. In contrast, GZO transforms to a disordered fluorite structure under irradiation with no indications of localized amorphization. But, the accumulation of Kr within the host material leads to sub-grain structures, extended defects, and the development of micro-cracks. Thus, while GTO readily amorphizes even under low dose irradiations, the resistance of the amorphous GTO matrix to micro cracking and gas release, even in the presence of large bubble formation, suggests an enhanced propensity to retain gaseous species. Consideration of long-term dose accumulation effects in nuclear waste forms would suggest reconsideration of amorphization processes in pyrochlores and related materials as a potential beneficial effect for immobilization and long-term storage of actinide materials.

The nucleation and subsequent growth of helium bubbles in bulk tungsten is investigated using molecular dynamics simulations. By considering a setting that includes the diffusion process of helium clusters, we study their attachment to existing bubbles and their interaction with tungsten crowdion structures generated in the bubble growth process. We find that incoming helium atoms, and especially small helium clusters, can become trapped in the crowdion structures, providing nucleation sites for new helium bubbles, and leading to a distributed network of bubbles rather than a single, growing bubble. The nature of this network depends on both the temperature and the implantation flux of helium. Our results indicate that the kinetic interaction of He with generated dislocations is a key factor dictating the evolution of bubble distributions in plasma-exposed tungsten.

A potentially enhanced radiation resistance of nanocrystalline materials, as a consequence of the high density of interfaces and surfaces, has attracted much attention both to understand the fundamental role of these defect sinks and to develop them for high-radiation environments. Here, irradiation response of nanocrystalline A2Ti2O7 (A = Gd, Ho and Lu) pyrochlore powders with grain sizes of 20–30 nm was investigated by 1-MeV Kr2+ ion bombardment. In situ transmission electron microscopy (TEM) revealed that the critical amorphization fluence for each nanocrystalline compound at room temperature was greater than that for their coarse-grained counterparts, indicating an enhanced amorphization resistance. The effect of temperature on the irradiation response of one of these compounds, nanocrystalline Lu2Ti2O7, was further examined by performing ion irradiation at an elevated temperature range of 480–600 K. The critical amorphization temperature (Tc) was found to be noticeably higher in nanocrystalline Lu2Ti2O7 (610 K) than its coarse-grained counterpart (480 K), revealing that nanocrystalline Lu2Ti2O7 is less resistant to amorphization compared to its coarse-grained phase under high temperatures. We interpret these results with the aid of atomistic simulations. Molecular statics calculations find that cation antisite defects are less energetically costly to form near surfaces than in the bulk, suggesting that the nanocrystalline form of these materials is generally less susceptible to amorphization than coarse-grained counterparts at low temperatures where defect kinetics are negligible. In contrast, at high temperatures, the annealing efficiency of antisite defects by cation interstitials is significantly reduced due to the sink properties of the surfaces in the nanocrystalline pyrochlore, which contributes to the observed higher amorphization temperature in the nano-grained phase than in coarse-grained counterpart. Together, these results provide new insight into the behavior of nanocrystalline materials under irradiation.

Nickel superalloys with cubic L12 structured γ′ (Ni3(Al, Ti)) precipitates exhibit high strength at high temperatures and excellent corrosion resistance when exposed to water. Unlike prior studies on irradiation damage of other Ni-based superalloys, our study on Rene N4 involves much larger γ′ precipitates, ∼450 nm in size, a size regime where the irradiation-induced disordering and dissolution kinetics and the corresponding mechanical property evolution are unknown. We report that under heavy ion irradiation at room temperature, the submicron-sized γ′ precipitates were fully disordered at ∼0.3 dpa and only later partially dissolved after 75 dpa irradiation. Nanoindentation experiments indicate that the mechanical properties of the alloy change significantly, with a dramatic decrease in hardness, with irradiation dose. Three contributions to the change in hardness were examined: defect clusters, disordering and dissolution. The generation of defect clusters in the matrix and precipitates slightly increased the indentation hardness, while disordering of the submicron-sized γ′ precipitates resulted in a dramatic decrease in the total hardness, which decreased further during the early stages of the intermixing between γ′ precipitates and matrix (

ZnO nanowires (NWs) have been extensively studied for various device applications. Although these nanowires are often suspected to be impractical and highly unstable under hostile radiation environments, to date little is known on their radiation tolerance. Here, we show outstanding resilience of ZnO NWs by using in situ Kr ion irradiation at room temperature inside a transmission electron microscope. Our studies show that ZnO nanowires with certain diameters become nearly immune to radiation damage due to the existence of dislocation loop denuded zones. A remarkable size effect also holds: the smaller the nanowire diameter, the lower the defect density. Rate theory modeling suggests that the size effect arises from fast interstitial migration and a limit in size to which interstitial loops can grow. In situ studies also revealed a surprising phenomenon: the pristine prismatic loops can prevail over the strongest known defect sinks, free surfaces, to trap radiation-induced defect clusters. This study comprises the first critical step toward in-depth understanding of radiation response of functional oxide nanowires for electronic device applications in extreme environments.

Oxide dispersion strengthened ferritic/martensitic alloys have attracted significant attention for their potential uses in future nuclear reactors and storage vessels, as the metal/oxide interfaces act as stable high-strength sinks for point defects while also dispersing helium. Here, in order to unravel the evolution and interplay of interface structure and chemistry upon irradiation in these types of materials, an atomically sharp FeCr/MgO interface was synthesized at 500 °C and separately annealed and irradiated with Ni3+ ions at 500 °C. After annealing, a slight enrichment of Cr atoms was observed at the interface, but no other structural changes were found. However, under irradiation, sufficient Cr diffuses across the interface into the MgO to form a Cr-enriched transition layer that contains spinel precipitates. First-principles calculations indicate that it is energetically favorable to incorporate Cr, but not Fe, substitutionally into MgO. Furthermore, our results indicate that irradiation can be used to form new phases and complexions at interfaces, which may have different radiation tolerance than the pristine structures.

Cr diffusion in trilayer thin films of 100 nm Fe–18Cr/125 nm TiO2−x/100 nm Fe–18Cr deposited on MgO substrates at 500 °C was studied by either annealing at 500 °C or Ni3+ ion irradiation at 500 °C. Microchemistry and microstructure evolution at the metal/oxide interfaces were investigated using (high-resolution) transmission electron microscopy, energy-dispersive X-ray spectroscopy and electron energy loss spectroscopy. Diffusion of Cr into the O-deficient TiO2 layer, with negligible segregation to the FeCr/TiO2−x interface itself, was observed under both annealing and irradiation. Cr diffusion into TiO2−x was enhanced in ion-irradiated samples as compared to annealed. Irradiation-induced voids and amorphization of TiO2−x was also observed. The experimental results are rationalized using first-principles calculations that suggest an energetic preference for substituting Ti with Cr in sub-stoichiometric TiO2. The implications of these results on the irradiation stability of oxide-dispersed ferritic alloys are discussed.

The point defect structure of cementite (Fe 3 C) is investigated using a combination of the statistical mechanical Wagner–Schottky model and first-principles calculations within the generalized gradient approximation. Large 128-atom supercells are employed to obtain fully converged point defect formation energies. The present study unambiguously shows that carbon vacancies and octahedral carbon interstitials are the structural defects in C-depleted and C-rich cementite, respectively. The dominant thermal defects in C-depleted and stoichiometric cementite are found to be carbon Frenkel pairs. In C-rich cementite, however, the primary thermal excitations are strongly temperature-dependent: interbranch, Schottky and Frenkel defects dominate successively with increasing temperature. Using the nudged elastic band technique, the migration barriers of major point defects in cementite are also determined and compared with available experiments in the literature.