Interstitial Loop Growth Research Articles

Revealing the critical role of vanadium in radiation damage of tungsten-based alloys

Refractory tungsten-based medium- and high-entropy alloys form a promising class of strong, heat- and radiation-resistant materials for nuclear applications. Here, we systematically explore the radiation tolerance of Mo–Nb–Ta–V–W alloys using molecular dynamics simulations and a machine-learned interatomic potential. Going from pure W to W-based equiatomic binaries, ternaries, quaternary, and the quinary high-entropy alloy, we simulate the radiation damage accumulation up to 0.4–0.8 dpa through cumulative collision cascades. We find that the radiation damage and evolution are controlled by a combination of cascade-induced clustering, the balance in migration of vacancies and interstitials, and defect cluster binding energies. These mechanisms are strongly influenced by atom size, among which vanadium as the smallest atom is decisive. In pure W and alloys without V, the microstructure is characterised by large and growing interstitial dislocation loops. In stark contrast, in alloys containing V, vacancy dislocation loops are formed directly in collision cascades and the balanced migration rate of vacancies and interstitials promotes recombination and prevents growth of interstitial loops. The atom size differences also lead to clear segregation, with small atoms (V) in interstitial clusters and larger atoms (Ta, Nb) segregating around vacancy clusters. Furthermore, the small cluster sizes, formation of vacancy loops, segregation, and balanced migration in V-containing alloys lead to low swelling and an exceptionally efficient defect annealing compared to pure W and V-free alloys. Our results highlight WTaV as a promising low-activation material, where almost 90% of defects are annealed at 2000 K during 5 ns compared to <20% in pure W.

A Monte Carlo model of irradiation-induced recrystallization in polycrystalline UMo fuels

Experiments show that irradiation-induced recrystallization speeds up the swelling kinetics in U-10 wt% Mo fuels. However, recrystallization mechanisms and the effect of initial grain microstructures on recrystallization kinetics are still unclear. In this work a Monte Carlo model coupling the rate theory of defect evolution has been developed to study the irradiation-induced recrystallization. The rate theory is used to describe the spatial evolution of gas bubbles, interstitials and interstitial loops; First-Passage Kinetic Monte Carlo (FPKMC) approach is used to describe the fast and strongly anisotropic migration of interstitials, and a Cellular Automata method is used to model the formation of recrystallized grains. With the assumption that recrystallization may occur when the local interstitial loop density is larger than a given critical value, simulation results reveal that 1) recrystallized grains first nucleate on grain boundaries and the recrystallization zone front moves to the center of original coarse grains in the UMo matrix, and 2) recrystallization starts earlier in coarse polycrystalline structures, while the overall recrystallization kinetics decreases with increasing grain size. These results agree with experimental observations. The comparison of recrystallization kinetics obtained from experiments and modeling suggests that the interstitial loop accumulation leads to the recrystallization and the interstitial loop growth is suppressed inside coarse grains due to the over-pressured intra-granular gas bubbles.

A multiscale method for the analysis of defect behavior in Mo during electron irradiation

In order to overcome a lack of experimental information on values for key materials properties and kinetic coefficients, a multiscale modeling approach is applied to defect behavior in irradiated Mo where key materials properties, such as point defect (vacancy and interstitial) migration enthalpies as well as kinetic factors such as dimer formation, defect recombination, and self interstitial–interstitial loop interaction coefficients, are obtained by molecular dynamics calculations and implemented into rate-theory simulations of defect behavior. The multiscale methodology is validated against interstitial loop growth data obtained from electron irradiation of pure Mo. It is shown that the observed linear behavior of the loop diameter vs. the square root of irradiation time is a direct consequence of the 1D migration of self-interstitial atoms.

In situ study of self-ion irradiation damage in W and W–5Re at 500 °C

In situ self-ion irradiations (150 keV W+) have been carried out on W and W–5Re at 500 °C, with doses ranging from 1016 to 1018 W+m−2 (∼1.0 dpa). Early damage formation (1016W+m−2) was observed in both materials. Black–white contrast experiments and image simulations using the TEMACI software suggested that vacancy loops were formed within individual cascades, and thus, the loop nucleation mechanism is likely to be ‘cascade collapse’. Dynamic observations showed the nucleation and growth of interstitial loops at higher doses, and that elastic loop interactions may involve changes in loop Burgers vector. Elastic interactions may also promote loop reactions such as absorption or coalescence or loop string formation. Loops in both W and W–5Re remained stable after annealing at 500 °C. One-dimensional hopping of loops (b = 1/2 ⟨111>) was only seen in W. At the final dose (1018W+m−2), a slightly denser damage microstructure was seen in W–5Re. Both materials had about 3–4 × 1015 loops m−2. Detailed post-irradiation analyses were carried out for loops of size ⩾ 4 nm. Both b = 1/2 ⟨111⟩ (∼75%) and b = ⟨100> (∼25%) loops were present. Inside–outside contrast experiments were performed under safe orientations to determine the nature of loops. The interstitial-to-vacancy loop ratio turned out close to unity for 1/2 ⟨111⟩ loops in W, and for both 1/2 ⟨111⟩ and ⟨100⟩ loops in W–5Re. However, interstitial loops were dominant for ⟨100⟩ loops in W. Re seemed to restrict loop mobility, leading to a smaller average loop size and a higher number density in the W-Re alloy.

Computer simulations of interstitial loop growth kinetics in irradiated bcc Fe

The growth kinetics of (001) [001] interstitial loops in bcc Fe is studied by phase-field modeling. The effect of defect (vacancy/interstitial) concentration, generation, recombination, sink strength, and elastic interaction on the growth kinetics of interstitial loops is systematically simulated. Results show that the elastic interaction between the defects and interstitial loops speeds up the growth kinetics and affects the morphology of the interstitial loops. Linear growth rate, i.e., the loop average radius is linear to time, under both aging and irradiation are predicted, which is in agreement with experimental observation. The results also show that the interstitial loop growth rate, which is directly related to the sink strength of the interstitial loop for interstitials, increases linearly with the initial interstitial concentration during aging while changing logarithmically with the interstitial generation rate under irradiation.

Capture efficiency for clustering reaction between charged defects in β-SiC

In order to investigate the formation kinetics of interstitial loops in β-SiC during irradiation, a capture efficiency of a charged point defect to a charged sink was derived taking into account electric interaction between the defects. The derived capture efficiency depends much on effective charges of defects. For defect accumulation in β-SiC under high-energy electron irradiation, the numerical analysis using the rate theory based model with the capture efficiency was performed, and the calculation results indicate that an interstitial loop grows while keeping its charge to be as neutral as possible, leading to the suppression of the growth of interstitial loops. Roughly speaking, the calculated concentration of clustered interstitials with effective charges approaches to the experimental data. It implies that the capture efficiency derived in the present work may play an important role for better reproducing of the experimental result.

Irradiation-Induced Recrystallization of Cellular Dislocation Networks in Uranium-Molybdenum Alloys

ABSTRACTWe developed a rate-theory-based model to investigate the nucleation and growth of interstitial loops and cavities during low-temperature in-reactor irradiation of uranium-molybdenum alloys. Consolidation of the dislocation structure takes into account the generation of forest dislocations and capture of interstitial dislocation loops. The theoretical description includes stress-induced glide of dislocation loops and accumulation of dislocations on cell walls. The loops accumulate and ultimately evolve into a low-energy cellular dislocation structure. Calculations indicate that nanometer-size bubbles are associated with the walls of the cellular dislocation structure. The accumulation of interstitial loops within the cells and of dislocations on the cell walls leads to increasing values for the rotation (misfit) of the cell wall into a subgrain boundary and a change in the lattice parameter as a function of dose. Subsequently, increasing values for the stored energy in the material are shown to be sufficient for the material to undergo recrystallization. Results of the calculations are compared with SEM photomicrographs of irradiated U- 10Mo, as well as with data from irradiated UO2.

Damage structures in fission-neutron irradiated Ni-based alloys at high temperatures

The defects formed in Ni based (Ni–Si, Ni–Cu and Ni–Fe) alloys which were irradiated with fission-neutrons were examined by electron microscopy. Irradiations were carried out at 473 K and 573 K. In the 473 K irradiated specimens, a high density of large interstitial loops and small vacancy clusters with stacking fault tetrahedra (SFT) were observed. The number densities of these two types of defects did not strongly depend on the amount of solute atoms in each alloy. The density of the loops in Ni–Si alloys was much higher than those in Ni–Cu and Ni–Fe alloys, while the density of SFT only slightly depended on the kind of solute. Also, the size of the loops depended on the kinds and amounts of solute. In 573 K irradiated Ni-Cu specimens, a high density of dislocation lines developed during the growth of interstitial loops. In Ni–Si alloys, the number density and size of the interstitial loops changed as a function of the amount of solute. Voids were formed in Ni–Cu alloys but scarcely formed in Ni–Si alloys. The number density of voids was one hundredth of that of SFT observed in 473 K irradiated Ni–Cu alloys. Possible formation processes of interstitial loops, SFT dislocation lines and voids are discussed.

Microstructure of copper and nickel irradiated with fission neutrons near 230°C

The microstructures of pure copper and nickel specimens were examined by transmission electron microscopy (TEM) following irradiation at ∼ 230°C in a mixed spectrum fission reactor to damage levels between 0.01 and 0.25 displacements per atom (dpa). A high density of small stacking fault tetrahedra (SFT) and dislocation loops was observed in both materials, and a moderate density of small voids was observed in the irradiated copper specimens. From a comparison with published studies, the proportion of SFTs and dislocation loops in copper was observed to be nearly constant over a wide range of damage levels at temperatures between 200 and 250°C. This suggests that nucleation and growth of interstitial loops to visible sizes is rather difficult in copper in this temperature regime. On the other hand, the loop size and density in nickel increased steadily with increasing dose. The defect clusters in nickel formed a {001} planar wall pattern which became visible after damage levels of ∼ 0.1 dpa. Defect cluster patterning was not observed in the irradiated copper specimens.

Dislocation climb and interstitial loop growth under cascade damage irradiation

The effects of intracascade clustering and recombination in radiation damage have been considered previously in semiquantitative calculations involving vacancy accumulation at voids, within the concept of production bias. To model void swelling and microstructural evolution quantitatively, similar effects on dislocation climb and interstitial loop growth have to be considered. In this regard, at elevated temperatures (such as in the peak-swelling temperature regime), the concentration of freely migrating vacancies is much higher than that of the interstitials, owing to the evaporation from the primary vacancy clusters (i.e. those produced by intracascade clustering). It is not immediately obvious how the dislocations can be net interstitial sinks, and hence that the observed nucleation and growth of the interstitial loops at elevated temperatures can be correctly predicted as in the conventional theory.

Accumulation and annihilation processes of cascades in metallic and nonmetallic crystals under irradiation with ions and/or electrons

In situ observation has been performed under dual-beam irradiation with heavy ions and fast electrons in the HVEM-accelerator facility at Kyushu University. The objectives of the present study are to understand the structure and the accumulation process of cascade damage and the synergistic effect of electronic excitation and/or free point defects on the behavior of cascade damage in metallic and nonmetallic crystals, such as Cu, graphite, SiC, Si, Ge, Ge-20at%Si, MgO, α-Al 2O 3, MgAl 2O 4 and others. The HVEM-accelerator facility has been confirmed to be extremely useful for observing the behavior of cascade damage: (1) Cascade damage is directly produced without the help from other collision cascades in Cu. (2) Most collisions themselves in Si, Ge and Ge-20at%Si induce invisible structural change through electron microscopy, but they are converted into visible ones by the impact from other collision cascades. (3) No visible cascade damage is formed in covalent crystals with light masses such as graphite and SiC and in ionic crystals such as MgO, α-Al 2O 3 and MgAl 2O 4. (4) Free interstitials suppress the evolution of visible cascade damage and assist the nucleation and the growth of interstitial loops around collision cascades in metallic and nonmetallic crystals.

Effect of external tensile stress on the growth of interstitial loops

The behaviour of the climb of Frank loops of the interstitial type formed in Ag and Cu during electron irradiation at high temperature in the presence of an external tensile stress was studied using a combined in-situ irradiation/deformation experiment in a high-voltage electron microscope. It was found that an external tensile stress has a profound effect on climb rates of Frank loops of interstitial type, and that the climb rates of otherwise crystallographically equivalent edge dislocations become anisotropic in the presence of the external tensile stress. The results were compared with theoretical predictions made by Bullough using stress-induced preferential absorption (SIPA) mechanism of point defects, and it was found that the present experimental results can not be explained easily by the Bullough's model.

Effects of Systematic Modification with Oversized Elements on Void Formation in 316L Austenitic Stainless Steel under Electron Irradiation

Solution annealed type 316L austenitic stainless steels modified with IV or V group elements were electron irradiated up to a dose of 10.8 dpa at temperatures of 623 to 843 K in an HVEM. X-ray diffraction analysis was also performed to determine the quantitative size factors of the additional elements in the 316L steel. All the additives are shown to be oversized atoms with the size factors in the order Hf>Zr>Ta>Nb>Ti>V. Each addition retards void swelling through a prolongation of the incubation period for void nucleation and a suppression of the void growth, and decreases the peak swelling and shifts it to lower temperature. The larger the size factor is, the more emphasized those tendencies are. The size, number density and resultant swelling of voids clearly shows a systematic dependence on the size factor. The increase in the apparent vacancy migration energy of the steel due to the presence of the additives and the interaction of vacancies with them are confirmed in terms of the interstitial loop growth. The above findings indicate that the enhancement of the vacancy-interstitial recombination and reduction of the vacancy mobility are attained through vacancy trapping at the oversized additional atoms, and furthermore, the atoms act as nucleation sites for voids.

Characteristics of point defects and their clustering in pure ferritic steels

Abstract HVEM 1 MeV-electron irradiation of Fe-10% Cr and Fe-10% Cr-1% Ni alloys and subsequent EDS analysis were carried out. A remarkable feature of the microstructural evolution is that nucleation and growth of interstitial loops is very slow. The analysis of the experimental results based on the rate theory of defect processes showed that the stable nuclei of interstitial loops were tri-interstitials and that absorption of free interstitials by the loop was restricted. These could be the reasons for the slow accumulation of damage. Addition of Ni in the alloy reduces the mobility of interstitials and thus suppresses the evolution of microstructure.

Calculated strength of irradiated ferritics

Effects of irradiation on dislocation recovery kinetics and strengthening have been calculated and compared to reported responses in Fe-2 1 4Cr-lMo steels. The analysis indicates that predicted interstitial loop growth causes increases in strength at 673 K that are comparable to measure increases in strength at 10 dpa. With increasing temperature, radiation softening was predicted as expected from experimental observations. The transition from hardening to softening was predicted to occur at temperatures about 50 K above the measured transition. A strong hardening effect was predicted from depleted zone accumulation at 673 K, but only when depleted zones were assumed to be unbiased microvoids. Radiation enhanced growth of Mo 2C precipitates were predicted to have an influence on hardening that was less than predicted for hardening caused by loop growth. At later stages of carbide evolution, irradiation softening may occur because of reduction in solid solution hardening and enhancement in carbide coarsening.

Interstitial loop growth in HVEM-irradiated nickel foils

A recently presented model for the development of a dislocation density in HVEM-irradiated thin foils has been applied to nickel. From matching the results to previously reported experimental measurements of loop radius growth as a function of position in the foil, the dislocation bias and vacancy migration energy in nickel have been determined. A comparison has been carried out of the effect on the results of using two distinct sink strengths to model the flow of defects to the dislocation loops. Equally good fits to the experimental data have been obtained with either sink strength and consequently it has not proved possible to select a preferable representation for the loops.

Interstitial loop growth in irradiated thin foils

In order to understand the behaviour of materials in an irradiation environment, a knowledge of the interaction between point defects and dislocations is essential. It has been shown previously that using the rate theory continuum method, the loss of defects to the foil surface being accounted for through a smeared-out foil surface sink strength, measurements of interstitial loop growth rate in irradiated thin foils can yield such information. The present work describes a scheme for loop growth in a thin foil, based on the rate theory continuum model, which accounts directly for defect loss to the foil surface. A set of simultaneous time-dependent equations for defect concentrations and loop growth are solved numerically on a finite number of equally-spaced mesh points in the foil. Account is taken of loop overlap onto neighbouring mesh points and of loop intersection with the foil edge. Loop growth measurements in copper have been analysed with this scheme and values deduced for vacancy migration energy and dislocation bias.

Defect aggregation in irradiated Ni3Al and NiAl

Defect aggregation has been investigated by high-voltage electron irradiation at temperatures from 300 K to 810 K in Ni3Al and NiAl, which are Ll2 and B2 ordered alloys respectively. For both alloys, the number densities of radiation-induced dislocation loops decrease but the size of the loops increase with increasing irradiation temperature. In Ni3Al, both extrinsic 13<111/111 Frank loops and interstitial type 12<011/011 anti-phase loops are created. Transformation from Frank loops to anti-phase loops have been observed in Ni3Al during irradiation. In NiAl, only interstitial type perfect <001/001 prismatic loops are created. Under the same irradiation conditions, interstitial loop growth rates in NiAl are much smaller than those in Ni3Al, implying that the migration energy of vacancies in NiAl must be much higher than that in Ni3Al.

Evidence for vacancy migration at 220 k in iron, given by high voltage electron microscopy

Abstract The growth of interstitial loops has been studied by 1 MeV electron microscopy in pure iron samples, irradiated and observed between 180 K and 800 K. It is shown that, below 220 K, the increase of the interstitial loop diameter obeys a tn law, with n<1 (t being the observation time). Above 220 K, a linear t law is observed. Moreover, the evolution of the loop density is strongly modified below and above 220 K. Both results lead to the conclusion that vacancies migrate at 220 K in iron. The discrepancy with previous results obtained by other physicists is explained by a model in which mobile vacancies interact with residual impurities.

Investigation of pulsed irradiation effects on interstitial loop growth in metals

A computer program for the solution of non steady-state diffusion equations describing the evolution of point defects and interstitial dislocation loops during pulsed and continuous irradiation is developed. The equations take into account mutual recombination of point defects, defect migration to dislocation loops and line dislocations, and the existence of equilibrium thermal vacancies. It is shown that interstitial loops grow from 2 to 9 run in diameter due to the surplus flux of interstitials in the non steady-state regime (dynamic preference) at 573 K. At 873 K the dislocation loops begin to shrink owing to line tension forces. Comparison of interstitial loop and vacancy behaviour for pulsed and continuous irradiation at 573 and 873 K is performed. It is shown that at pulse duration 2 × 10 −6 s and repetition rate 100 pulses/s, pulsing does not affect the interstitial loop behaviour.