Sinks For Radiation-induced Defects Research Articles

Oxides dispersion-strengthened (ODS) FeCoNiCuZn high entropy alloys through different rare earth elements: Synthesis, structural, physical, and experimental radiation transmission properties

The oxide dispersoids within ODS alloys can act as sinks for radiation-induced defects, such as vacancies and dislocation loops, effectively reducing their mobility and preventing their accumulation. This property is crucial for HEAs employed in radiation-intensive environments, such as nuclear reactors. The objective of this research is to examine the impact of rare earth elements (REE) such as Y2O3, Er2O3, Pr2O3, and Sm2O3, on Oxides Dispersion-Strengthened (ODS) FeCoNiCuZn High Entropy Alloys (HEAs). The mechanical alloying technique is employed to produce a high entropy alloy consisting of Fe, Co, Ni, Cu, and Zn in their raw form. Subsequently, the raw alloy powder is enriched with separate amounts of Y2O3, Er2O3, Pr2O3, and Sm2O3. The microstructural analysis of the samples obtained from the mechanical alloying process was performed utilizing the X-ray diffraction (XRD) technique. In addition, scanning electron microscopy (SEM) was employed to analyze the ODS-HEA samples encoded S1, S2, S3, S4, and S5. To investigate the transmission properties of gamma-ray and neutron radiation, experimental studies are carried out using two types of detectors: Ultra High Purity Germanium (HPGe) detector and Canberra NP-100B BF3 gas proportional detector, respectively. The X-ray diffraction (XRD) spectra of samples did not display any observable peaks that could be attributed to the presence of dispersed rare earth element (REE) oxides. The uniform distribution of the metallic constituents that make up the High Entropy Alloy (HEA) is observed in the samples. Additionally, it can be observed that the implementation of the ODS-HEA technique, incorporating a 3% (wt.) Er2O3 additive, results in the most advantageous results with respect to the characteristics of gamma ray absorption. The S3 sample demonstrated the greatest degree of neutron absorption, as demonstrated by a recorded value of 0.857 μSv/h, where the S1 sample demonstrated the minimum level of absorption, as evidenced by a recorded value of 0.452 μSv/h. Based on the observed effects on neutron and gamma-ray attenuation behaviors in ODS-HEAs, it can be concluded that Er2O3 exhibits characteristics of a monotonic oxide. This feature is particularly advantageous for applications that necessitate a dual enhancement in these behaviors. It can also be concluded that the S1 sample may be deemed appropriate for situations where the utmost consistency of chain reactions in nuclear reactor fuel rods is desired, due to its possession of the lowest neutron absorption properties.

The mechanism behind the high radiation tolerance of Fe–Cr alloys

Fe–Cr alloys are at the forefront for high radiation tolerant materials with long-standing validated performance. Yet, the detailed mechanism behind their high radiation resistance is in question and understanding the effect of varying Cr percentage is a grand challenge limiting further improvements. Here, we applied depth-resolved positron annihilation lifetime spectroscopy and Doppler broadening spectroscopy to study the effect of Cr alloying on the formation and evolution of atomic size clusters induced by ion-irradiation in Fe. We also used atom probe tomography to investigate the possible presence of Cr clusters or α′ phase precipitates with high Cr composition. The study reveals that the well-known resistance to radiation in Fe–Cr alloys may arise from the stabilization of vacancy clusters around Cr atoms, which act as sinks for radiation-induced defects. This implies that Cr atoms do not provide a direct sink for interstitials; rather defect complexes that consist of Cr atoms and vacancies, in turn, act as sinks for irradiation-induced vacancies and interstitials. we also find that lower amounts of Cr create smaller defect clusters that act as efficient sinks for radiation damage, but larger quantities of Cr form a defect structure that is less homogenous and larger in size, resulting in less efficient damage recombination. No evidence of α′ was found before or after irradiation, which indicates that it does not contribute to the observed radiation tolerance.

Radiation-resistant nanotwinned austenitic stainless steel

A key strategy to increase the radiation resistance of materials has been to introduce a high density of interfaces that can act as sinks for radiation-induced defects. Twin boundaries are a type of interface that can be introduced through deformation but are usually considered to be ineffective sinks. Using heavy ion irradiation and transmission electron microscopy, this study investigates the influence of a high area per unit volume of twin boundaries on the radiation-induced swelling response of an austenitic stainless steel. The study shows that swelling can be suppressed in regions containing a high density of closely-spaced deformation twin boundaries.

Self-healing mechanism of irradiation defects in nickel–graphene nanocomposite: An energetic and kinetic perspective

The self-healing mechanism of radiation-induced defects in nickel–graphene nanocomposite is investigated by atomistic simulations. Compared with pure nickel, nickel–graphene nanocomposite has less defects remained in the bulk region after collision cascades, illustrating self-healing performance. Nickel–graphene interfaces (NGIs) serve as sinks for radiation-induced defects and preferentially trap interstitials over vacancies. Energetic and kinetic calculations reveal that the defect formation energy and diffusion barrier are reduced in the vicinity of NGIs, and the reduction are pronounced for interstitials. When NGIs are loaded with interstitials, their segregation ability on radiation-induced defects improves significantly, and the radiation-induced defects near the NGIs diffuse more easily. Especially, the vacancies (or interstitials) near the NGIs tend to annihilate (or aggregate) with the interstitials trapped at the NGIs, which only happens at the interstitial-loaded side of NGIs. Therefore, nickel–graphene nanocomposite exhibits excellent radiation tolerance and shows promise as a structural material for advanced nuclear reactors due to its NGIs with the energetic and kinetic driving forces acting on radiation-induced defects.

Size effect in AlN/SiN multilayered films irradiated with helium and argon ions

The results of microstructure and phase composition of the AlN/SiNx multilayered films after the irradiation with helium and argon ions are presented. The multilayered films with alternating nanocrystalline (nc-AlN) and amorphous (a-SiNx) phases with thicknesses from 2 to 10 nm were obtained by reactive magnetron sputtering. X-ray diffraction results showed the dependence of the crystal lattice parameters of the nc-AlN phase on the thickness which is explained by size affect. After the irradiation with helium (30 keV) and argon (180 keV) ions the radiation-induced point defects as well as their clusters are produced in the films and mainly localized in the amorphous a-SiNx layers. It is of the consequence of enhanced implanted ions migration towards the a-SiNx layers and bubbles formation, as revealed by high-resolution transmission electron microscopy. The average size of the bubbles is 2.0–2.4 nm and grows up to 4–5 nm after the post-irradiation vacuum (800 °C) annealing. The amorphous a-SiNx layers are believed to serve as sinks for radiation-induced defects.

Radiation tolerance of Cu/W multilayered nanocomposites

Magnetron sputtered Cu/W multilayer samples with individual layer thicknesses from 2.5 to 50 nm were irradiated by 50 keV He + ions at ion fluences from 7 × 10 20 to 6 × 10 21 m −2 at room temperature. Evolution of the interfacial structure during irradiation is monitored by X-ray diffraction and cross-sectional transmission electron microscopy. Moreover, radiation responses on the individual layer thickness and He + ion irradiation fluence are revealed. The highly morphological stability of the multilayered structure suggests that the interfacial structure and grain boundary can serve as sinks for radiation-induced defects.

Efficient Annealing of Radiation Damage Near Grain Boundaries via Interstitial Emission

Although grain boundaries can serve as effective sinks for radiation-induced defects such as interstitials and vacancies, the atomistic mechanisms leading to this enhanced tolerance are still not well understood. With the use of three atomistic simulation methods, we investigated defect-grain boundary interaction mechanisms in copper from picosecond to microsecond time scales. We found that grain boundaries have a surprising "loading-unloading" effect. Upon irradiation, interstitials are loaded into the boundary, which then acts as a source, emitting interstitials to annihilate vacancies in the bulk. This unexpected recombination mechanism has a much lower energy barrier than conventional vacancy diffusion and is efficient for annihilating immobile vacancies in the nearby bulk, resulting in self-healing of the radiation-induced damage.

Design of Nano-Composites for Ultra-High Strengths and Radiation Damage Tolerance

AbstractThe combination of high strength and high radiation damage tolerance in nanolaminate composites can be achieved when the individual layers in these composites are only a few nanometers thick and therefore these materials contain a large volume fraction associated with interfaces. These interfaces act both as obstacles to slip, as well as sinks for radiation-induced defects. The morphological and phase stabilities of these nano-composites under ion irradiation are explored as a function of layer thickness, temperature and interface structure. Using results on model systems such as Cu-Nb, we highlight the critical role of the atomic structure of the incoherent interfaces that exhibit multiple states with nearly degenerate energies in acting as sinks for radiation-induced point defects. Reduced radiation damage also leads to a reduction in the irradiation hardening, particularly at layer thickness of approximately 5 nm and below. The strategies for design of radiation-tolerant structural materials based on the knowledge gained from this work will be discussed.