Formation Of Interstitial Loops Research Articles

Exploring radiation damage in (Hf0.2Zr0.2Ta0.2Ti0.2Nb0.2)C high-entropy carbide ceramic: Integrating experimental and atomistic investigations

This study investigates the intricate mechanisms that govern irradiation damage in high-entropy ceramic materials. Specifically, we synthesized (Hf0.2Zr0.2Ta0.2Nb0.2Ti0.2)C high-entropy carbide ceramics (HECC) with a single-phase rock-salt structure using spark plasma sintering. These ceramics were then subjected to irradiation with 1.08 MeV C ions, resulting in a dose of 7.2 dpa (dpa: displacements per atom) at both room temperature (RT) and 500 °C. To understand the resulting damage structure, we analyzed bulk irradiated HECC samples using Grazing Incidence X-ray Diffraction (GIXRD) and Transmission Electron Microscope (TEM) at both irradiation temperatures. GIXRD analysis revealed an average tensile strain out-of-plane of 0.16% for RT irradiation and 0.14% for irradiation at 500 °C. In addition, TEM analysis identified a buried damaged band, approximately 970 nm thick, under both irradiation temperatures. By employing the bright field TEM imaging technique under kinematic two-beam conditions, dislocation loops of both a/3 〈111〉{111} and a/2 〈110〉{110} types within the damaged band were observed. Furthermore, our analysis indicated an increase in the average size of the total dislocation loops within the band from 1.2 nm to 1.4 nm as the density decreased. Importantly, no amorphization, precipitates, or voids were detected in the damaged band under both irradiation temperatures. Density functional theory (DFT) simulations indicated that carbon predominantly resides in 〈110〉 split interstitial sites causing lattice expansion, while vacancies, particularly Nb, induced compression along the c-axis. Carbon atoms tend to bond when collectively present in the <110> split interstitial sites, contributing to the formation of interstitial loops.

Shockwave generates < 100 > dislocation loops in bcc iron

The formation mechanism of < 100 > interstitial dislocation loops in ferritic steels stemming from irradiation remains elusive, as their formations are either too short for experiments, or too long for molecular dynamics simulations. Here, we report on the formation of both interstitial and vacancy dislocation loops in high energy displacement cascades using large-scale molecular dynamics simulations with up to 220 million atoms. Riding the supersonic shockwave generated in the cascade, self-interstitial atoms are punched out to form < 100 > dislocation loops in only a few picoseconds during one single cascade event, which is several orders of magnitude faster than any existing mechanisms. The energy analysis suggests that the formation of the interstitial loops depends on kinetic energy redistribution, where higher incidence energy or larger atom mass could improve the probability of the direct nucleation of interstitial dislocation loops.

Helium bubbles aggravated defects production in self-irradiated copper

Under the environment of high radiation, materials used in fission and fusion reactors will internally accumulate numerous lattice defects and bubbles. With extensive studies focused on bubble resolution under irradiation, the mutually effects between helium bubbles and displacement cascades in irradiated materials remain unaddressed. Therefore, the defects production and microstructure evolution under self-irradiation events in vicinity of helium bubbles are investigated by preforming large scale molecular dynamics simulations in single-crystal copper. When subjected to displacement cascades, distinguished bubble resolution categories dependent on bubble size are observed. With the existence of bubbles, radiation damage is aggravated with the increasing bubble size, represented as the promotion of point defects and dislocations. The atomic mechanisms of heterogeneous dislocation structures are attributed to different helium-vacancy cluster modes, transforming from the resolved gas trapped with vacancies to the biased absorption of vacancies by the over-pressured bubble. In both cases, helium impedes the recombination of point defects, leading to the accelerated formation of interstitial loops. The results and insight obtained here might contribute to understand the underlying mechanism of transmutant solute on the long-term evolution of irradiated materials.

Fermi-Level Effects on Extended Defect Evolution in Si+ and P+ Implanted In0.53Ga0.47As

The evolution of implant damage in InGaAs is studied for electrically active Si+ and isoelectronic P+ implants. Extrinsic loops formed by excess interstitials are shown to be less stable upon annealing for n-type Si+ implants relative to isoelectronic P+ implants. Damage created by P+ implants into heavily n-doped InGaAs is also shown to be less stable than damage created by P+ implants into unintentionally doped InGaAs indicating that the background doping concentration can significantly effect the evolution of implant damage upon annealing. Previous results have suggested that the electrical activation and diffusion behavior of n-type dopants, like Si in InGaAs, may be strongly influenced by vacancy concentration. TEM results in this study also suggest that heavy n-type doping in InGaAs results in the formation of a large population of vacancy defects that enhance the dissolution or inhibit formation of interstitial loops.

Characterization of swift heavy ion irradiation damage in ceria

Abstract

Microstructural stability of an HT-9 fuel assembly duct irradiated in FFTF

To explore whether the known resistance of fully tempered HT-9 to neutron-induced phase instability and void swelling are maintained under realistic time-dependent reactor operating conditions, the radiation-induced microstructure of an HT-9 ferritic/martensitic hexagonal duct was examined following a 6-year irradiation campaign of a fuel assembly in the Fast Flux Test Reactor Facility (FFTF). Microscopy examination was conducted on specimens irradiated to 4 dpa at 505 °C, 28 dpa at 384 °C and 155 dpa at 443 °C where quoted temperatures are the average operating temperatures over the lifetime of the duct. The dislocation and phase microstructure were observed to remain relatively unchanged at 4 dpa at 505 °C, but significant microstructural changes were observed to have occurred at 28 and 155 dpa and 384 and 443 °C respectively. At these doses the microstructures have experienced precipitation and formation of interstitial loops. In addition, void swelling had occurred at 155 dpa with an average swelling of ∼0.3%, although some local areas swelled as much as 1.2%. In general it appears that this alloy retains its swelling resistance under typical reactor operation conditions up to 155 dpa.

Formation of interstitial loops in tungsten under helium ion irradiation: Rate theory modeling and experiment

Formation of interstitial loops in tungsten under helium ion irradiation has been evaluated by microstructural observation and numerical model calculation. In the model, two types of mechanisms of interstitial loop nucleation are employed in addition to introducing an effect of trap mutation. When the binding energy of an SIA to a helium-vacancy complex is around 0.7 eV, the irradiation-time dependence of calculated interstitial loop density shows good agreement with the experiment. The energy value (0.7 eV) is consistent with molecular dynamics evaluations in the literature.

Effect of neutron irradiation on the microstructure and hardness in particle dispersed ultra-fine grained V–Y alloys

Effects of neutron irradiation on the microstructure and hardness in V–1.6 and 2.6Y (mass%) alloys were studied. The alloys consisted of both fine grains of a few hundreds of nanometers in diameter with finely dispersed particles and small amounts of coarse grains of a few micrometers in diameter without dispersed particles. The microstructures after neutron irradiation at 290 and 800 °C to 0.25 and 0.7 dpa, respectively, were examined with transmission electron microscopy. Irradiation did not change the grain size. Interstitial loops and voids were formed in the coarse grains irradiated at 290 °C, while voids were formed only in the fine grains. The number density of voids decreased with decreasing grain size. Interstitial loops and voids were not observed in V–1.6Y alloy irradiated at 800 °C, while thin plates were formed only in the coarse grains. The formation of interstitial loops and voids was efficiently suppressed in the fine grains.

Physical property change of heavily neutron-irradiated Si 3N 4 and SiC by thermal annealing

Changes in macroscopic length, lattice parameter and thermal diffusivity of neutron-irradiated Si 3N 4 and SiC ceramics up to a fluence of 4.2×10 26 n/ m 2 were measured. Macroscopic length increase of Si 3N 4 was almost one half of that of SiC. Thermal diffusivity of both ceramics was reduced severely by the irradiation at 390–540°C. Slight increase in the a-axis and slight decrease in the c-axis lattice parameter were detected for Si 3N 4. The amount of lattice parameter change of Si 3N 4 was very small compared with the macroscopic length change. Changes in these properties due to post-irradiation thermal annealing up to 1500°C were measured. Large part of thermal diffusivity of Si 3N 4 was recovered by annealing, with small step at ∼1100°C, but macroscopic length did not significantly change by annealing. Change in lattice parameter showed a complicated trend. It is supposed that formation of interstitial loops on the planes parallel to the c-axis, formation of voids during annealing or difficulty of recovery of points defects/loops, or solid solution formation due to glassy grain boundary phase may influence the recovery behavior of Si 3N 4 ceramics. Changes in macroscopic length, lattice parameter or thermal diffusivity of SiC by annealing coincided with the results of previous works. The critical irradiation conditions for loop formation/XRD line broadening for SiC is discussed based on the present and previous results.

Investigation on modification of plasma facing surface under long duration discharges by means of a collector probe in TRIAM-1M

Metallic specimens mounted on a collector probe head were inserted in the scrape-off layer and exposed to long pulse tokamak plasma in TRIAM-1M. Strong radiation damage (formation of interstitial loops) was caused by high energy (>2 keV) charge-exchange hydrogen neutrals. The surfaces of the probe specimens were covered by new material made of co-deposited in-vessel elements such as Fe, Cr, Mo, O and C. The deposited layer was often exfoliated by blistering. These results indicate that energetic hydrogen neutrals and the co-deposition of impurities considerably change the surface properties of plasma facing components and even introduce new impurity sources.

Irradiated Microstructures of Magnesium Aluminate Spinel and their Controlling Factors

AbstractThis paper reviews our recent progress in study on the strong resistance of magnesium aluminate spinel to void swelling during irradiation, along with the related characteristic features of its radiation damage. Comparative experimental results on irradiated microstructures and mechanical properties in magnesium aluminate spinel and alpha-alumina are shown in terms of controlling factors of radiation resistance of the former crystal. It is experimentally shown that structural vacancies due to non-stoichiometry provide effective recombination sites for displaced cations to suppress the formation of interstitial loops. Decreased formation of interstitial loops enhances the further recombination of interstitials and vacancies and thereby the formation of voids.

Effect of Formation and Growth of Dislocation Loops and Cavities on Low-Temperature Swelling of Irradiated Uranium-Molybdenum Alloys

AbstractScanning electron photomicrographs of U–10 wt.% Mo irradiated at low temperature in the Advanced Test Reactor (ATR) to about 40 at.% burnup show the presence of cavities. We have used a rate-theory-based model to investigate the nucleation and growth of cavities during low-temperature irradiation of uranium-molybdenum alloys in the presence of irradiation-induced interstitial-loop formation and growth. Our calculations indicate that the swelling mechanism in the U–10 wt.% Mo alloy at low irradiation temperatures is fission-gas driven. The calculations also indicate that the observed bubbles must be associated with a subgrain structure. Calculated bubble-size-distributions are compared with irradiation data.

On density effects in point defect solutions under irradiation. II. Formation of interstitial dislocation loops

The chemical potential approach, making it possible for one to account for density effects in supersaturated solution of point defects, is applied to consideration of interstitial loop formation kinetics in metals under homogeneous irradiation. Quasi-chemical rate equations are reformulated in terms of chemical potential, rather than in terms of concentrations of species involved in reactions, following a procedure known in the theory of irreversible processes. The extension of rate equations is tested on a simpler model of bimolecular reactions. Then the density effect, such as radiation induced spontaneous clustering of interstitials, is investigated in irradiated metals on the background of interstitial loop formation as described by conventional rate equation solution. The phenomenon, that reveals itself by spontaneous formation of di-interstitials, would be expected to become noticeable at irradiation temperatures about 0.25 to 0.3 of the absolute melting temperature if the rate of defect generation is high enough. For a steady-state solution analyzed, the density effect brings about the higher number density of forming interstitial loops with reduced mean dimensions if compared with conventional based consideration. Possibilities of experimental investigations of the proposed mechanism of interstitial loop formation are discussed.

Interstitial cluster formation in metals under intense irradiation

A new mechanism of point defect clustering in irradiation environment is proposed. The process considered reflects a possibility for spontaneous (nondiffusional) clustering to exist in a highly enough supersaturated solution of point defects under certain conditions. The influence of the production of di-interstitial clusters by this mechanism on the kinetics of interstitial dislocation loop nucleation and growth is analyzed. It is shown that the intensity of irradiation must be high enough and the temperature of the sample must be correspondingly low enough for the effect of the spontaneous small interstitial clustering to prevail over the usual mechanism of diffusion driven nucleation and growth of interstitial clusters. Both analytical and numerical solutions of an extended version of conventional rate equations demonstrate the possibility of this phenomenon. This version takes into account the terms of higher order in an expansion of the chemical potential over the interstitial concentration. The new form of equations naturally follows from a canonical approach to chemical-type rate equations that is known in nonlinear nonequilibrium thermodynamics. When density effects are negligible, one gets the conventional kinetics of diffusion driven interstitial loop formation. A simple model of interstitial loop nucleation and growth through di-interstitial is used in the analysis of the clustering kinetics. A noticeable enhancement of interstitial loop number density with smaller radii is expected for temperatures below one third of the melting point, where the loop growth process due to diffusional processes is weak enough. For this novel phenomenon to be noticeable, a prolonged irradiation is necessary. Conditions and possibilities for experimental investigations of the predicted effect as well as the possible competitive processes are considered.

Effect of cyclic temperature change on microstructural evolution in austenitic stainless steels under fission neutron irradiation

To understand mechanistically the influence of cyclic temperature change on microstructural evolution in Fe-Cr-Ni austenitic alloys, temperature-cycle neutron irradiations of 473 K/673 K and 573 K/723 K were carried out at JMTR (0.9−1.4 × 10 24 n/m 2 (> 1.0 MeV)). Suppression of interstitial loop formation was remarkable. Only a few large voids were formed in Fe-Cr-Ni and Fe-Cr-Ni-P alloys, but unexpectedly large void swelling (1.5–1.6%, 9–11%/dpa) was observed in Fe-Cr-Ni-Ti alloy. Rate theory of defect clustering for temperature-variant irradiation indicates that vacancy-rich condition appears temporarily at the beginning of the high-temperature period due to the reclustering of the vacancies formed in the low-temperature period through the reactions with radiation-induced vacancies and interstitials.

Radiation Enhanced Diffusion in Mgo

AbstractMeasurements of marker spreading following 2.0 MeV Kr+irradiation at 25 to 1300°C have been performed on MgO samples containing O18, Ca and Zn buried tracer atoms. Ion beam mixing at room temperature on both sublattices was approximately 2.0 to 3.0 A5/eV. From 600 to 1000°C, the apparent activation enthalpy for diffusion on the anion sublattice (O18) was 0.35 eV and the diffusion coefficient was linear in the irradiation flux. From 1150 to 1300°C the measured activation enthalpy was 4.1 eV and the diffusion coefficient was proportional to the square root of flux. The measured activation enthalpy on the cation sublattice was roughly 0.30 eV up to 700°C for both Ca or Zn doped samples. Measurable extrinsic thermal diffusion from vacancies present for trivalent impurity charge compensation occurred above this temperature, complicating the analysis at higher temperatures. The observed kinetics in the lower temperature range are most likely controlled by interstitial loop formation. In the higher temperature range, vacancy traps with a binding energy of approximately 2 eV could account for the high activation enthalpy.

Swelling and microstructure of A1N irradiated in a fast reactor

Abstract Aluminum nitride (A1N) ceramics were neutron-irradiated in the fast-breeder reactor, JOYO, up to 9.1×10 25 n / m 2 ( E >0.1 MeV ) at about 500°C. The macroscopic length expanded 1.74%, whereas the lattice parameter of the c-axis expanded 2.11% compared with 0.05% that of the a-axis. Microstructural observation revealed the formations of high density interstitial-type dislocation loops in grains and microcracks at grain boundaries. The expansion of the c -axis should be related to the introduction of stacking faults by interstitial loop formation on the basal plane. The change in macroscopic volume calculated from macroscopic length became larger than that in a unit cell volume calculated from the lattice parameter. The discrepancy started at around 1% in volume change, corresponding to a fluence of 8×10 24 −2× 10 25 n / m 2 at irradiation temperatures around 500°C. The bending strength of the specimen in this study reduced significantly, which was caused by the microcrack formation along the grain boundary due to the significant anisotropy of lattice expansion.

Formation of interstitial loops on transmutation products in aluminum

Effects of the transmutation products on the nucleation of interstitial loops in aluminum have been investigated by electron microscopy. Zone refined aluminum was pre-irradiated with neutrons, helium, and hydrogen ions at elevated temperature and further irradiated with 200 keV electrons at temperatures between 100–323 K to investigate the formation of the interstitial loops. The loop density showed remarkable enhancement by pre-irradiation and dependence on the electron-irradiation temperature. It was suggested that isolated silicon atoms, helium-vacancy complexes, a hydrogen-vacancy complexes act as nucleation sites for the interstitial loops during the subsequent electron irradiation.

Effect of hydrogen on the formation of interstitial loops in hydrogen-irradiated aluminum

The role of hydrogen atoms in the formation of interstitial type dislocation loops in aluminum irradiated with 15 keV hydrogen ions was investigated by electron microscopy. It was found that the loop density satisfied and Arrhenius relation with the pre-irradiation temperature, yielding an apparent activation energy of about 0.3 eV. Experimental results are interpreted in terms of the mechanism that small hydrogen-vacancy complexes formed by the pre-irradiation of hydrogen ions act as nucleation sites of dislocation loops in the subsequent electron irradiation. A basically identical mechanism seems to be acting in the irradiation with hydrogen ions at room temperature.

A hrem study of neutron-irradiation-induced dislocations in aluminium nitride

Abstract Aluminium nitride polycrystals have been irradiated to fluences up to 5·2 × 1024 neutrons m-2 (E > 1·0 MeV) in a thermal-fast mixed reactor, the JMTR. Dislocations induced by the neutron irradiation were studied by high-resolution electron microscopy coupled with a multi-slice simulation. Very small interstitial loops, a few nanometres in diameter, lying on the (0001) plane of AIN were observed directly for the specimens irradiated to fluences higher than 2·4 × 1024 neutrons m-2, and it was determined that those loops have a Burgers vector of (c/2)[0001]. The dilation of the c axis by neutron irradiation was greater than that of the a axis, and this anisotropy of the lattice dilation is ascribed to the formation of interstitial loops lying on the basal plane.