Vacancy-interstitial Recombination Research Articles

Defect production in tungsten under sub-threshold energy irradiation: role of hydrogen and surface effects

Despite the low solubility of hydrogen isotopes (HIs) in tungsten (W), their concentration can reach up to ∼10 at.% after low-energy plasma irradiation. This is generally attributed to the vacancies that may accommodate excessive HIs. However, the kinetic energy of incident HIs transferred to W is far below the energy threshold to create a Frenkel pair, the underlying mechanism of defect production is still unclear. Here, we investigate the influence of H on the defect production in W using the molecular dynamic (MD) simulations. It is found that the threshold displacement energy (TDE) in bulk W slight decreases with the increasing of H concentration. This is due to the formation of H-vacancy complexes, which prevents the vacancy-interstitial recombination. More importantly, the H effects are significantly magnified in the surface region. On the one hand, the maximum kinetic energy transferred from 400 eV H to W can reach up to ∼21 eV due to the double-hit process, which is two times higher than that predicted by elastic collision model. On the other hand, the momentum transferred to W is completely random, including both the recoil direction upward and downward from the surface. Accordingly, the lowest TDE in W surface is only 15–21 eV at sub-surface layers with the depth of 6.7–11.1 Å, which is lower than the maximum kinetic energy transferred to W. Therefore, the low-energy HIs irradiation can create the defects in W surface directly. Our findings provide deep insight into defect production in W at sub-threshold energy and have wider implications for materials performance under low-energy ions irradiation.

Radiation-resistant binary solid solutions via vacancy trapping on solute clusters

Additions of solute that trap vacancies slow down vacancy diffusion and promote point-defect recombination in alloys subjected to irradiation. Such selective alloying can thus help to minimize the detrimental consequences resulting from point defect fluxes. The current work investigates the effect of solute additions on the recombination rate using kinetic Monte Carlo simulations for a model alloy system, which was parametrized to Cu-Ag in the dilute limit, but with an increased solubility limit, ≈ 0.86 at.% at 300 K. As the solute concentration was increased above 0.1 at.%, solute clustering was observed and led to a strong increase in recombination rate. The beneficial effects of solute clustering on reducing vacancy mobility, and reducing solute drag, were analyzed by calculating relevant transport coefficients using the KineCluE code (Schuler et al., Computational Materials Science (2020) 172,109,191). Moreover, it was observed in the KMC simulations that large recombination rates resulted in a shift of steady-state distributions of solute cluster sizes to smaller clusters compared to equilibrium distributions in the solid solution. This shift is rationalized as resulting from the irreversible character of the interstitial-vacancy recombination reaction. These results suggest a novel irradiation effect on phase stability where a high recombination rate increases the solubility limit of a solute at steady state over its equilibrium value.

Collaborative motion of helium and self-interstitial atoms enhanced self-healing efficiency of irradiation-induced defects in tungsten

Helium (He) is a typical impurity element and plays a crucial role in the microstructural evolution in nuclear materials under irradiation. Here, we systematically investigate the interactions between He and self-interstitial atoms (SIAs) as well as their influences on the kinetic behaviors of SIAs in tungsten (W), using both first-principles and object kinetic Monte Carlo methods. It is found that there are attractive interactions between He and SIAs, which become stronger with the increasing of SIA numbers. Specifically, the He-SIA1 and He-SIA2 complexes adopt a three-dimensional (3D) migration pattern with an effective energy barrier of 0.38 and 0.61 eV, respectively, which is completely different from the 1D migration of SIAs in W (⩽0.033 eV) without He. Such an unexpected collaborative 3D motion of He-SIA complexes increases the probability of vacancy-interstitial recombination and reduces the number of surviving defects. Consequently, our calculations reveal the enhanced effect of He on the self-healing efficiency in W, which is originated from the collaborative 3D motion of He-SIA complexes. The current results can improve our fundamental understanding of the influence of He on the evolution of irradiation defects and have great implications to estimate the performance of W-PFMs in fusion environment.

Reduced interstitial mobility through multicomponent alloying in bcc W

We report the first principles calculations of the defect energetics and dynamics in W ternary alloys. A solute pair of different kinds strongly attracts a W self-interstitial atom, forming a solute-interstitial triple complex. The attractive solute-interstitial interaction can be stronger in ternary than in binary alloys, which works as a primary dissociation barrier to reduce the interstitial mobility. This, in turn, means that a single solute-interstitial pair also attracts a solute atom of a different kind to form a triple complex. The dissociation energy of a triple complex slows the interstitial migration, and increases the odds of vacancy-interstitial recombination in collision cascades. Therefore the reduced interstitial mobility and increased interstitial retention ability of multicomponent alloys may change the fundamental irradiated-damage mechanism in fusion environment.

The effect of flux on ion irradiation-enhanced precipitation in AISI-316L: An in-situ TEM study

Thin foils of AISI 316L stainless steel were irradiated in-situ in a transmission electron microscope (TEM) with 325 keV Xe ions at 550 °C at three different fluxes to study flux effects. The kinetics of the radiation-enhanced precipitation (REP) and the evolution of the precipitates were found to be correlated with the irradiation flux. At lower fluxes (1 and 2 × 1012 ions∙cm−2∙s−1), cascade mixing played an important role in the accumulation of point defects within the austenite matrix, facilitating the formation of clusters which act as sinks for heterogeneous nucleation of precipitates with high areal density. At the highest flux (4 × 1012 ions∙cm−2∙s−1) the cascade mixing favours the recombination of vacancies and interstitials which supresses the growth of existing precipitates beyond a certain total damage level. The results agree with a previous radiation-enhanced precipitation model proposed by Wiedersich, Okamoto and Lam and further studied by Bruemmer, but a small modification is proposed when the flux is close to the vacancy-interstitial recombination limit.

The Effect of Flux on Ion Irradiation-Induced Precipitation in AISI-316L: An In-Situ TEM Study

Thin foils of AISI 316L stainless steel were irradiated in-situ in a transmission electron microscope (TEM) with 325 keV Xe ions at 550°C at three different fluxes to study flux effects. The kinetics of the radiation-induced precipitation and the evolution of the precipitates were found to be correlated with the irradiation flux. At lower fluxes (1 and 2 × 1012 ions·cm-2·s-1), cascade mixing played an important role in the accumulation of point defects within the austenite matrix, facilitating the formation of clusters which act as sinks for heterogeneous nucleation of precipitates with high areal density. At the highest flux (4 × 1012 ions·cm-2·s-1) the cascade mixing favours the recombination of vacancies and interstitials which supresses the growth of existing precipitates beyond a certain total damage level. The results agree with a previous radiation-induced precipitation (RIP) model proposed by Wiedersich, Okamoto and Lam and further studied by Bruemmer, but a small modification is proposed when the flux is closer to the vacancy-interstitial recombination limit.

Modelling the influence of strain fields around precipitates on defect equilibria and kinetics under irradiation in ODS steels: A multi scale approach

Resistance to radiation damage is a crucial property for structural materials in fission or fusion reactors. The particle-matrix interface in oxide dispersion strengthened (ODS) steels is considered a sink for point defects created during irradiation. In this work we address the question, if elastic strain fields around oxide particles cause a long-range interaction between the precipitates and point defects. We use kinetic Monte Carlo simulations to simulate the diffusion of point defects under the influence of elastic strain fields caused by Y2O3 and Y2Ti2O7 precipitates. We show that there is essentially no interaction between vacancies and the strain fields, while the sink strength of precipitates for interstitials increases with misfit strain between precipitate and matrix. Consequently, the interstitial concentration decreases with misfit strain, while due to reduced interstitial-vacancy recombination the vacancy concentration increases with misfit strain. The total change of point defect concentration with misfit strain is, however, rather limited.

In-situ TEM study of microstructural evolution in NFA and Cr3C2@SiC-NFA composite during ion irradiation

In this work, the ion irradiation responses of a Fe-based nanostructured ferritic alloy or ‘NFA’ (Fe–9Cr–2W–0.2V–0.4Ti–0.3Y2O3) and a Cr3C2@SiC-NFA composite were assessed. In-situ ion irradiation with TEM observation was carried out by using 1 MeV Kr++ ions at doses of 0, 1, 3, 5, 10 dpa and temperatures of 300 °C and 450 °C. Both the NFA and Cr3C2@SiC-NFA samples showed significant dislocation density after 10 dpa at 300 °C. However, the Cr3C2@SiC-NFA composite showed a significantly lower dislocation loop density and a smaller average loop size during the irradiation at 450 °C as opposed to the NFA. At 300 °C, 1/2<111> type dislocation loops were observed in both the NFA and Cr3C2@SiC-NFA samples. Interestingly, at 450 °C, <100> type loops were dominant in the NFA sample while 1/2<111> type loops were still dominant in the Cr3C2@SiC-NFA sample. The results were discussed based on the large surface sink effects and enhanced interstitial-vacancy recombination at higher temperatures. The additional Si element in the Cr3C2@SiC-NFA sample might have played a significant role in determining the dominant loop types.

Investigation of kinetic recovery process in low dose neutron-irradiated nuclear graphite by thermal annealing

ABSTRACT To investigate the kinetic recovery process of low dose neutron-irradiated graphite, nuclear-grade isotropic graphite IG-110U and ETP-10 were neutron irradiated using the JMTR up to 1.38 × 1023 n/m2 (En > 1 MeV) at ~473 K. In-situ measurement of macroscopic length was conducted during the isothermal and isochronal annealing process from room temperature up to 1673 K. From room temperature to 773 K for IG-110U, and to 1023 K for ETP-10, macroscopic lengths, lattice parameters, and unit cell volumes of both specimens recovered to their pre-irradiation values, and this recovery process subdivided into two stages. The activation energies of macroscopic volume recovery at 523–673 K and 673–773 K were determined to be ~0.22 eV and ~0.57 eV for IG-110U, respectively; ~0.13 eV and ~2.59 eV at 523–923 K and 923–1023 K for ETP-10, respectively. The migration of not only single interstitials but also interstitials dissociated from submicroscopic interstitial groups along basal planes followed by vacancy-interstitial recombination play a dominant role in the first stage. The second stage is suggested to proceed via the motion of carbon groups along basal planes for IG-110U, and migration of single interstitials along the c-axis for ETP-10. During 773 K or 1023 K up to 1673 K, macroscopic length continuously shrank with decreasing shrinking rate, even with a turnaround to swell at 1173 K for IG-110U.

Modeling the Influence of Strain Fields Around Precipitates on Defect Equilibria and Kinetics Under Irradiation in ODS Steels: A Multi Scale Approach

Resistance to radiation damage is a crucial property for structural materials in fission or fusion reactors. The particle-matrix interface in oxide dispersion strengthened (ODS) steels is considered a sink for point defects created during irradiation. In this work we address thequestion, if elastic strain fields around oxide particles cause a long-range interaction between the precipitates and point defects. We use kinetic Monte Carlo simulations to simulate the diffusion of point defects under the influence of elastic strain fields caused by Y2O3 and Y2Ti2O7 precipitates. We show that there is essentially no interaction between vacancies and the strain fields, while the sink strength of precipitates for interstitials increases with misfit strain between precipitate and matrix. Consequently, the interstitial concentration decreases with misfit strain, while due to reduced interstitial-vacancy recombination the vacancy concentration increases with misfit strain. The total change of point defect concentration with misfit strain is, however, rather limited.

First-principles study of solvent-solute mixed dumbbells in body-centered-cubic tungsten crystals

Tungsten (W) is considered as a promising candidate for plasma-facing materials for future nuclear fusion devices, and selecting optimal alloying constituents is a critical issue to improve radiation resistance of the W alloys as well as to improve their mechanical properties. We conducted in the current study a series of first-principles calculations for investigating solvent-solute mixed dumbbells in W crystals. The results suggested that titanium (Ti), vanadium (V), and chromium (Cr) are favorable as solutes for W alloys from irradiation-effect perspectives because these elements are expected to promote vacancy-interstitial recombination without causing radiation-induced precipitation that reduces ductility of irradiated materials.

Three-dimensional phase-field simulations of intragranular gas bubble evolution in irradiated U-Mo fuel

The evolution of fission gas bubbles in irradiated materials plays a critical role in the microstructural processes that leads to dimensional changes of U-Mo alloy fuels, e.g., fuel swelling. Although the intergranular bubbles-induced fuel swelling has been long-discussed for U-Mo fuel, there are very few computational studies of the formation of intragranular gas bubbles and its impact on fuel swelling. To this end, we develop a three-dimensional phase-field model to investigate the evolution of intragranular gas bubbles in U-Mo fuel. Fission induced defect formation and annihilation processes, such as vacancy-interstitial recombination, fission gas atom resolution, and interactions with dislocations and grain boundaries are incorporated in the model. Simulations show that the intragranular gas bubbles can be stabilized to certain sizes due to the balance between the generation and annihilation of defects. The intragranular gas bubbles induced fuel swelling is predicted to be comparable to experimental measurements. The effects of the irradiation and fuel fabrication conditions (i.e., fission rate, fuel grain size, and mechanical work induced deformation) on the bubble evolution and the resultant swelling are investigated. The current simulations provide a better understanding of intragranular gas bubble-induced swelling and a solid foundation for the future study of the nucleation and growth of intergranular gas bubbles and recrystallization in U-Mo fuel.

Radiophotoluminescence and thermoluminescence characteristics of undoped and Mg doped LiF phosphor in the high dose region

Radiophotoluminescence (RPL) and thermoluminescence (TL) from undoped and Mg-doped LiF powder samples were investigated for high dose dosimetric applications. F-aggregate centres in LiF are created by radiolysis. The RPL emission at 530 nm and 650 nm from F3+ and F2 centres respectively (on excitation at 450 nm) in undoped commercial LiF powder enhanced with increasing sintering temperature. Its TL glow curve showed peaks at 145, 290 and 365 °C. The RPL-TL correlation studies carried out for the first time showed that the thermal annihilation of F3+ and F2 centes following Fi0 - Va (interstitial-vacancy) recombination cause 290 °C and 365° TL peaks respectively. This showed that F2 centres are thermally more stable than F3+ centres. Its integrated TL response kept increasing nearly linearly up to 14 kGy, in contrast to the response of the TL peaks in dosimetry grade LiF:Mg, Ti (TLD-100). However, the TL from LiF suffers from fading problems. A 100 °C, 15 min pre-read annealing treatment erased most of the 145 °C TL peak without disturbing the other high temperature peaks. A 750° C, 22 h sintering treatment in carbon atmosphere shifted the major TL glow peak of undoped LiF to 200 °C but reduced its TL sensitivity by more than a factor of 10. Mg-doped LiF samples exhibit a TL glow curve peaking at ∼240 °C with satellite peaks typical of Mg doped LiF whose shape did not change significantly with gamma dose. However it showed lower TL and RPL sensitivities as compared to those of undoped LiF. All LiF except LiF alfa (750 °C, 1 h, CB) showed sublinear increase in TL with dose and therefore will cause inaccuracy in dose measurements especially at high doses due to reduced sensitivity. LiF alfa 750 °C, 22 h and 1000 °C, 6 min sintered powder samples are better than others because their RPL sensitivity increases nearly linearly with dose. However, a low fading observed in the latter sample supports its application. The bright yellow RPL emission seen through a 550 nm cut on filter illuminated with a 450 nm blue LED or He-Cd laser light could quickly confirm an irradiated consignment on which LiF samples are affixed. The increase in green to red emission ratio with dose and post-irradiation storage period might be related to slow migration of anion vacancies created by one secondary electron track to F centres created by another track and consequent conversion of F2 → F3+ centres.

On the possibility of universal interstitial emission induced annihilation in metallic nanostructures

Nano-structured metals have been explored as self-healing materials for radiation damage due to vacancy-interstitial recombination at interfaces, such as grain boundaries (GBs) in nanocrystals (NCs) and free surfaces in nanoporous metals (NPs). The annihilation in copper was recently proposed to be via the interstitial-emission (IE) from the GB. However, whether the IE is applied as a possibly universal mechanism in other NPs and NCs deserves further investigation. The interstitial is often considered to overcome a large binding energy with the GB/surface to migrate out of the GB/surface and induce annihilation. In this work, with emphasis on the IE in tungsten where the interstitial is exceptionally tightly bound to the tungsten GB/surface, we found that the interstitial only needed to migrate over a small distance from the GB/surface to recombine with the vacancy nearby at a small energy barrier. During the process, the annihilation region around a static interstitial at the GB/surface propagated or extended greatly to the adjacent bulk area. The annihilation was found to be via the IE from the GB/surface or coupled vacancy hop with the IE. The ratio of the interstitial-GB/surface binding energy to the trapping-radius of the GB/surface for the interstitial was proposed to be a better qualitative indicator for the difficulty in the IE than the binding energy. Given that the IE works in the investigated NPs and NCs tungsten, iron and copper where the interstitials of distinct configurations bind with the GBs/surfaces at diverse levels, the IE mechanism may work universally in various nano-structured metals.

Effect of Displacement Damage on Tantalum Oxide Resistive Memory

ABSTRACTThe radiation environment of space poses a challenge for electronic systems, in particular flash memory, which contains multiple radiation-sensitive parts. Resistive memory (RRAM) devices have the potential to replace flash memory, functioning as an inherently radiation resistant memory device. Several studies indicate significant radiation resistance in RRAM devices to a broad range of radiation types and doses. In this study, we focus on the effect of displacement damage on tantalum oxide-based RRAM devices, as this form of damage is likely a worst-case scenario. An Ar+ (170 keV) ion beam was used to minimize any contribution from ionization damage, maximizing the effect of displacement damage. Fluence levels were chosen to generate enough oxygen vacancies such that devices in the high resistance state (HRS) would likely switch to the low resistance state (LRS). More than half of devices tested at the highest fluence level (1.43E13 ions/cm2) switched from HRS to LRS. The devices were then switched for 50 set/reset cycles, after which the radiation-induced resistance shift disappeared. These results suggest that device switching may mitigate radiation damage by accelerating oxygen vacancy-interstitial recombination.

Long term simulation of point defect cluster size distributions from atomic displacement cascades in Fe70Cr20Ni10

We have used an Object Kinetic Monte Carlo (OKMC) model to simulate the long term evolution of the primary damage in Fe70Cr20Ni10 alloys. The mean number of Frenkel pairs created by different Primary Knocked on Atoms (PKA) was estimated by Molecular Dynamics using a ternary EAM potential developed in the framework of the PERFORM-60 European project. This number was then used to obtain the vacancy–interstitial recombination distance required in the calculation of displacement cascades in the Binary Collision Approximation (BCA) with code MARLOWE (Robinson, 1989). The BCA cascades have been generated in the 10–100keV range with the MARLOWE code and two different screened Coulomb potentials, namely, the Molière approximation to the Thomas–Fermi potential and the so-called “Universal” potential by Ziegler, Biersack and Littmark (ZBL). These cascades have been used as input to the OKMC code LAKIMOCA (Domain et al., 2004), with a set of parameters for describing the mobility of point defect clusters based on ab initio calculations and experimental data. The cluster size distributions have been estimated for irradiation doses of 0.1 and 1dpa, and a dose rate of 10−7dpa/s at 600K. We demonstrate that, like in the case of BCC iron, cluster size distributions in the long term are independent of the cascade energy and that the recursive cascade model suggested for BCC iron in Souidi et al. (2011) also applies to FCC Fe70Cr20Ni10. The results also show that the influence of the BCA potential is sizeable but the qualitative correspondence in the predicted long term evolution is excellent.

Kinetic modeling of Fluorine vacancy/F center creation in LiF:Mg,Ti including vacancy-interstitial recombination: Evaluating the factors leading to the lack of supralinearity in the optical absorption F center concentration dose response

Kinetic model simulations of charge carrier transport following irradiation of LiF:Mg,Ti (TLD-100) including Fluorine vacancy/F center creation by the radiation and dose-dependent vacancy-interstitial recombination are presented which describe the experimentally measured linear/exponentially saturating optical absorption dose response of the electron trapping centers at 4.0eV, 4.77eV, 5.08eV (F band) and 5.45eV. Linear/exponentially saturating dose response is commonly observed for centers which are not created by the radiation. The creation of Fluorine vacancies by the radiation could therefore be expected to lead to a supralinear dose response of the F center before the onset of saturation. Nonetheless, the dose response is linear from 10Gy to 500Gy and can be fitted with a dose-filling constant β=6.1·10−5Gy−1 corresponding to a 5% and 25% decrease from linearity at 103Gy and 5·103Gy respectively. The model attempts to resolve a central question concerning the mechanisms leading to the linear/exponentially saturating dose response of the F band even though Fluorine vacancies are being continuously created during the irradiation. The electron-trapping characteristics of the created vacancies are assumed to differ somewhat from the vacancies originally present in un-irradiated samples due to differences in their immediate environment. Vacancy-interstitial recombination for separation distances less than a critical distance, dc is demonstrated to be significant for D>500Gy (dc=36Å) and is an important mechanism contributing to the F center saturation at high dose-levels. The kinetic model accurately simulates the experimentally observed F center dose response over the entire investigated dose range of 10–105Gy under the following conditions: (i) The concentration of vacancies initially present is unexpectedly high at ∼1023m−3, possibly due to the highly doped, non-crystalline and hot-pressed nature of the LiF:Mg,Ti samples. (ii) The transition probability, An4o, for electron capture into the initially-present vacancies is ∼40 times greater than An4, the transition probability for the radiation-created vacancies. These two factors marginalize the effect of the created vacancies at low dose resulting in a linear dose response.

An energetic and kinetic perspective of the grain-boundary role in healing radiation damage in tungsten

Radiation-induced damage in tungsten (W) and W alloys has been considered as one of the most important issues in fusion research, because radiation-produced defects not only degrade the mechanical property but also change the behaviours of H and He in W significantly, such as the retention of H. Nano-structured W has been developed to reduce accumulation of defects within grains and further mitigate radiation-induced damage. However, the fundamental role of a grain boundary (GB) in healing radiation damage in W is not yet well understood. Using molecular dynamics and statics, we evaluate energetically and kinetically the role of a GB in defect evolution (vacancy and interstitial segregation and their annihilation) near the GB in W, by calculating the vacancy (interstitial) formation energy, segregation energy, diffusion barrier, vacancy–interstitial annihilation barrier near the GB and the corresponding influence range of the GB. We find that, as reported and expected, interstitials are preferentially trapped into GBs over vacancies during irradiation, with vacancies dominant near the GB and interstitials highly localized at the GB. On the one hand, the GB serves as a sink both for vacancies and interstitials near itself by reducing their formation energy and diffusion barrier. The formation energy of the vacancy decreases only by ∼0.86 eV, but 7.5 eV is reduced for the formation energy of the interstitial in the GB core, indicating that the sink is unexpectedly stronger for interstitials than vacancies. The average barrier of vacancy diffusion is 0.98 eV much less than 1.8 eV in the bulk; the interstitial migrates into the GB via a barrier-free process. On the other hand, the GB acts as a catalyst for the vacancy–interstitial recombination at the GB by lowering the annihilation barrier. The annihilation with the average barrier as low as 0.31 eV works even at a low temperature of 121 K; besides, the annihilation of a close vacancy–interstitial pair is spontaneous. Meanwhile, the annihilation mechanism near the GB is modified due to the exceptionally large reduction in the interstitial formation energy. The influence range of the GB is small (1–1.5 nm), leading to a small volume fraction of the GB region working as the sink and the catalyst. This indicates that GBs in fine W grains may play a limited role in improving radiation performance.

The role of thermal spike compactness in radiation-induced disordering and Frenkel pair production in Ni3Al

We show that the shape of the kinetic energy distribution in radiation-induced thermal spikes may be described using a dimensionless number, proportional to (volume) 2/3 /(surface area), known as compactness. The disorder produced in thermal spikes in Ni 3 Al increases with compactness because the thermal spike cooling rate, which determines the time available for thermal disordering, decreases with compactness. On the other hand, Frenkel pair production is inversely correlated to compactness because longer thermal spike lifetimes enhance vacancy–interstitial recombination.

The influence of dipole polarisation on threshold displacement energies in UO2

Molecular dynamics simulations of radiation damage in nuclear materials have traditionally been performed using rigid ion empirical potential models. The inclusion of dipole polarisability via the shell model offers the possibility to incorporate more of the underlying physics into simulations of radiation damage via cascade simulations. Here we assess the impact of dipole polarisability on the evolution of radiation damage in UO2 by performing molecular dynamics simulations to determine the threshold displacement energies using both a standard rigid ion potential and the equivalent model including dipole polarisability. The inclusion of dipole polarisability reduces the threshold displacement energies by 20–30% relative to the rigid ion model, suggesting a significant increase in the number of defects created during the collisional phase of the cascade. However, a reduction in the activation energies for migration due to inclusion of polarisability may increase the rate of subsequent interstitial–vacancy recombination. The overall level of radiation damage will, therefore, be determined by the relative effectiveness of these two opposing processes.