Radiation-induced Defects Research Articles

Bridging microscale to macroscale mechanical property measurements of FeCrAl alloys by crystal plasticity modeling

FeCrAl alloys are candidates for accident tolerant fuel cladding of light water reactors. In this work, a microstructure- and temperature-dependent crystal plasticity model is employed to bridge microscale to macroscale mechanical property measurements of FeCrAl alloys. With the visco-plastic self-consistent (VPSC) polycrystal plasticity framework, a mechanism-based single crystal plasticity (MSCP) model adopts the Arrhenius type rate equation to describe the dependence of the critical resolved shear stress for dislocation slips on their temperature-dependent intrinsic frictional resistance and the microstructure-dependent irradiation hardening. The intrinsic frictional resistance associated with {110}<111> and {112}<111> slip systems were measured by in-situ micromechanical testing on unirradiated/irradiated samples at 25-500 °C. The irradiation hardening is estimated by the Bacon-Kocks-Scattergood (BKS) model with density and size of radiation-induced defects measured from microstructural characterization. Several features associated with thermo-mechanical behavior of unirradiated/irradiated polycrystalline FeCrAl alloys are captured. High density of deformation-induced dislocations and radiation-induced defects results in obvious hardening at room temperature, which is weakened at high temperature, and facilitates damage evolution during deformation. Moreover, both high temperature and radiation-induced defects, which facilitate dislocation multiplication, trigger large hardening rate. The proposed method together with application of accelerator-based ion irradiation technique is a surrogate approach to simulate neutron damage, improving the efficiency associated with evaluation of mechanical properties of FeCrAl alloys exposed to temperature, stress and radiation conditions.

Effect of local chemical order on the irradiation-induced defect evolution in CrCoNi medium-entropy alloy

High- (and medium-) entropy alloys have emerged as potentially suitable structural materials for nuclear applications, particularly as they appear to show promising irradiation resistance. Recent studies have provided evidence of the presence of local chemical order (LCO) as a salient feature of these complex concentrated solid-solution alloys. However, the influence of such LCO on their irradiation response has remained uncertain thus far. In this work, we combine ion irradiation experiments with large-scale atomistic simulations to reveal that the presence of chemical short-range order, developed as an early stage of LCO, slows down the formation and evolution of point defects in the equiatomic medium-entropy alloy CrCoNi during irradiation. In particular, the irradiation-induced vacancies and interstitials exhibit a smaller difference in their mobility, arising from a stronger effect of LCO in localizing interstitial diffusion. This effect promotes their recombination as the LCO serves to tune the migration energy barriers of these point defects, thereby delaying the initiation of damage. These findings imply that local chemical ordering may provide a variable in the design space to enhance the resistance of multi-principal element alloys to irradiation damage.

Crystalline–Amorphous Nanostructures: Microstructure, Property and Modelling

Crystalline metals generally exhibit good deformability but low strength and poor irradiation tolerance. Amorphous materials in general display poor deformability but high strength and good irradiation tolerance. Interestingly, refining characteristic size can enhance the flow strength of crystalline metals and the deformability of amorphous materials. Thus, crystalline-amorphous nanostructures can exhibit an enhanced strength and an improved plastic flow stability. In addition, high-density interfaces can trap radiation-induced defects and accommodate free volume fluctuation. In this article, we review crystalline-amorphous nanocomposites with characteristic microstructures including nanolaminates, core-shell microstructures, and crystalline/amorphous-based dual-phase nanocomposites. The focus is put on synthesis of characteristic microstructures, deformation behaviors, and multiscale materials modelling.

Materials swelling revealed through automated semantic segmentation of cavities in electron microscopy images

Accurately quantifying swelling of alloys that have undergone irradiation is essential for understanding alloy performance in a nuclear reactor and critical for the safe and reliable operation of reactor facilities. However, typical practice is for radiation-induced defects in electron microscopy images of alloys to be manually quantified by domain-expert researchers. Here, we employ an end-to-end deep learning approach using the Mask Regional Convolutional Neural Network (Mask R-CNN) model to detect and quantify nanoscale cavities in irradiated alloys. We have assembled a database of labeled cavity images which includes 400 images, > 34 k discrete cavities, and numerous alloy compositions and irradiation conditions. We have evaluated both statistical (precision, recall, and F1 scores) and materials property-centric (cavity size, density, and swelling) metrics of model performance, and performed targeted analysis of materials swelling assessments. We find our model gives assessments of material swelling with an average (standard deviation) swelling mean absolute error based on random leave-out cross-validation of 0.30 (0.03) percent swelling. This result demonstrates our approach can accurately provide swelling metrics on a per-image and per-condition basis, which can provide helpful insight into material design (e.g., alloy refinement) and impact of service conditions (e.g., temperature, irradiation dose) on swelling. Finally, we find there are cases of test images with poor statistical metrics, but small errors in swelling, pointing to the need for moving beyond traditional classification-based metrics to evaluate object detection models in the context of materials domain applications.

Multi–length scale characterization of point defects in thermally oxidized, proton irradiated iron oxides

A key for the success of safe nuclear power generation system is to consider structural materials that are economical, meet mechanical property needs, possess good corrosion resistance, and are radiation tolerant. Nevertheless, fundamental insights that elucidate the details of radiation damage on materials corrosion performance are lacking. This includes the behavior of surface oxides which often regulate corrosion. For example, it is unclear how non-equilibrium point defects, oxide structure, mass transport in oxides, and subsequent oxidation behavior are altered by the radiation. In this work, some of the effects of proton irradiation on the attributes of point defects, iron oxide microstructures, and the physical nature of the oxidation product were correlated with corrosion behavior. Iron oxides, fabricated by thermal oxidation in air at 400°C and 800°C for 1 h, were subjected to 200 keV, 0.03 dpa (displacements per atom) of proton irradiation, and subjected to corrosion reactivity assessment using AC and DC electrochemical methods. Experimental methods that target materials information at different length scales, such as positron annihilation spectroscopy (atomistic), transmission electron microscopy (mesoscopic), and electrochemical methods (macroscopic), were coupled to shed light on the impact of radiation-induced defect modifications and structural changes in oxides on corrosion reactivity which ultimately affects durability in harsh environments.

Thermal and Radiation Stability in Nanocrystalline Cu

Nanocrystalline metals have presented intriguing possibilities for use in radiation environments due to their high grain boundary volume, serving as enhanced irradiation-induced defect sinks. Their promise has been lessened due to the propensity for nanocrystalline metals to suffer deleterious grain growth from combinations of irradiation and/or elevated homologous temperature. While approaches for stabilizing such materials against grain growth are the subject of current research, there is still a lack of central knowledge on the irradiation-grain boundary interactions in pure metals despite many studies on the same. Due to the breadth of available reports, we have critically reviewed studies on irradiation and thermal stability in pure, nanocrystalline copper (Cu) as a model FCC material, and on a few dilute Cu-based alloys. Our study has shown that, viewed collectively, there are large differences in interpretation of irradiation-grain boundary interactions, primarily due to a wide range of irradiation environments and variability in materials processing. We discuss the sources of these differences and analyses herein. Then, with the goal of gaining a more overarching mechanistic understanding of grain size stability in pure materials under irradiation, we provide several key recommendations for making meaningful evaluations across materials with different processing and under variable irradiation conditions.

Extraordinary radiation tolerance of a Ni nanocrystal-decorated carbon nanotube network encapsulated in amorphous carbon

Nanostructured materials with abundant defect sinks show good radiation tolerance due to their efficient absorption of irradiation-induced interstitials and vacancies. However, the poor thermal stability and limited size of such nanomaterials severely limit their practical applications. Herein, we report a novel flexible free-standing network-structured hybrid consisting of amorphous carbon encapsulated nickel nanocrystals anchored on a single-wall carbon nanotube scaffold with excellent radiation tolerance up to 5 dpa at 673 K and exceptional thermal stability up to 1073 K. The nano-scale Ni-SWCNT network with abundant Ni-SWCNT interfaces and grain boundaries provides effective sinks and fast transportation channels for defects, which effectively absorb irradiation-induced defects and improved the irradiation tolerance. Furthermore, the formation of a low-energy Ni-C interface and surface thermal grooves significantly reduces the system free energy and increased thermal stability. The amorphous carbon layer produces an external compressive radial stress that inhibits Ni grain boundaries from migrating, which greatly improves the thermal stability of the hybrid by pinning GBs at grooves between grains and facilitates the annihilation of irradiation-induced defects at the sinks. This work provides a new strategy to improve the thermal stability and radiation tolerance of nano-materials used in an irradiation environment.

Probing the Cr3+ luminescence sensitization in β-Ga2O3 with ion-beam-induced luminescence and thermoluminescence

Ion-beam-induced luminescence (IBIL) measurements were performed in Cr-doped β-Ga2O3 using both protons and helium ions, showing a strong enhancement of the Cr3+ luminescence upon ion irradiation. Theoretical modelling of the IBIL intensity curves as a function of the fluence allowed estimating the effective cross-sections associated with the defect-induced IBIL enhancement and quenching processes. The results suggest that sensitizing the Cr3+ luminescence is more efficient for H+ than for He+ irradiation. Thermoluminescence (TL) studies were performed in the pristine sample, with no TL signal being observed in the spectral region corresponding to the Cr3+ emission. In agreement with the IBIL study, upon ion irradiation (with either protons or helium ions), this TL emission is activated. Moreover, it can be quenched by annealing at 923 K for 10 s, thus revealing the role played by the defects induced by the irradiation. These results show that the irradiation-induced defects play a major role in the activation of the Cr3+ luminescence, a fact that can be exploited for radiation sensing and dosimetry.

Comparison of the dosimetric response of two Sr salts irradiated with 60Co γ-rays and synchrotron X-rays at ultra-high dose rate

As part of the development of innovative radiotherapy techniques using X-ray synchrotron sources, namely FLASH-radiotherapy and microbeam radiation therapy (MRT), two rod dosimeters based on strontium sulfate (SrSO4) and strontium carbonate (SrCO3) were developed and irradiated at the ESRF-ID17 Biomedical beamline to investigate new solid-state dosimeters.The radiation-induced defect in Sr-based dosimeters was detected by means of an electron paramagnetic resonance (EPR) spectrometer. For the study, the EPR response of the dosimeters irradiated with synchrotron X-rays in the 30–600 keV range and dose rate of 11.6 kGy/s was compared with the response after irradiation with a 60Co γ-ray source with a dose rate of 0.288 Gy/s.Both Sr dosimeters show an increased intensity of the EPR signal in the case of irradiation at the synchrotron with respect to the case of 60Co γ-ray irradiation. The measured enhancement of the delivered dose is explained by the fact that Sr shows an increased photoelectric interaction at orthovoltage X-ray energies, producing an increased number of free radicals.The experimental relative response rESRF of the EPR signals between samples irradiated at the ID17 beamline and the 60Co beam quality was found to be 9.66 for SrSO4 and 11.19 for SrCO3. Meanwhile, the theoritical relative responses rμen/ρESRF calculated as the ratio of mass energy absorption coefficients (μen/ρ) of the dosimeters to that for water at the ID17 with respect to the 60Co beam was found to be 11.85 for SrSO4 and 13.31 for SrCO3.The increased EPR response of Sr salt dosimeters irradiated with photons in the keV range proved that these detectors type could be a good candidate for the development of very sensible detectors for radiotherapy applications.

Radiation-induced segregation at grain boundaries of alloy 800H: Experimentally-informed atomistic simulations

The irradiation of alloys can alter the spatial distribution of solute and impurity elements. The redistribution of alloying elements can cause local chemical changes at surfaces and interfaces, which may degrade the mechanical and chemical properties of alloys. In this paper, a combination of microscopic experiments and atomistic simulations are employed to study the effect of irradiation damage on radiation-induced segregation (RIS) at grain boundaries (GBs) of the alloy 800H. The grain boundary character distribution in alloy 800H was determined by electron backscatter diffraction, and proton-irradiation-induced defects were identified by transmission electron microscopy. The experimentally-observed GBs were recreated in atomistic simulations to further study their atomistic structure, grain boundary energies, and local element concentration. The effects of radiation-induced defects on the redistribution of alloying elements at various GBs were investigated. A mechanism is proposed to explain the local chemical changes at GBs after the interaction with radiation-induced dislocation loops, which is of importance to nuclear reactor performance.

Microstructural Effects on Irradiation Creep of Reactor Core Materials

The processes that control irradiation creep are dependent on the temperature and the rate of production of freely migrating point defects, affecting both the microstructure and the mechanisms of mass transport. Because of the experimental difficulties in studying irradiation creep, many different hypothetical models have been developed that either favour a dislocation slip or a mass transport mechanism. Irradiation creep mechanisms and models that are dependent on the microstructure, which are either fully or partially mechanistic in nature, are described and discussed in terms of their ability to account for the in-reactor creep behaviour of various nuclear reactor core materials. A rate theory model for creep of Zr-2.5Nb pressure tubing in CANDU reactors incorporating the as-fabricated microstructure has been developed that gives good agreement with measurements for tubes manufactured by different fabrication routes having very different microstructures. One can therefore conclude that for Zr-alloys at temperatures < 300 °C and stresses < 150 MPa, diffusional mass transport is the dominant creep mechanism. The most important microstructural parameter controlling irradiation creep for these conditions is the grain structure. Austenitic alloys follow similar microstructural dependencies as Zr-alloys, but up to higher temperature and stress ranges. The exception is that dislocation slip is dominant in austenitic alloys at temperatures < 100 °C because there are few barriers to dislocation slip at these low temperatures, which is linked to the enhanced recombination of irradiation-induced point defects.

Effects of proton irradiation-induced point defects on Shockley–Read–Hall recombination lifetimes in homoepitaxial GaN p–n junctions

This work examined the intentional generation of recombination centers in GaN p–n junctions on freestanding GaN substrates. Irradiation with a 4.2 MeV proton beam was used to create a uniform distribution of vacancies and interstitials across GaN p+/n− and p−/n+ junctions through anode electrodes. With increasing proton dose, the effective doping concentrations were found to be reduced. Because the reduction in the doping concentration was much higher than the hydrogen atom concentration, this decrease could not be attributed solely to carrier compensation resulting from interstitial hydrogen atoms. In fact, more than half of the electron and hole compensation was caused by the presence of point defects. These defects evidently served as Shockley–Read–Hall (SRH) recombination centers such that the SRH lifetimes were reduced to several picoseconds from several hundred picoseconds prior to irradiation. The compensation for holes in the p−/n+ junctions was almost double that for electrons in the p+/n− junctions. Furthermore, the SRH lifetimes associated with p−/n+ junctions were shorter than those for p+/n− junctions for a given proton dose. These differences can be explained by variations in the charge state and/or the formation energy of intrinsic point defects in the p-type and n-type GaN layers. The results of the present work indicate the asymmetry of defect formation in GaN based on the fact that intrinsic point defects in p-type GaN readily compensate for holes.

Structural Properties of He-Irradiated Zr/Nb Multilayer Investigated by Grazing Incidence X-ray Diffraction

Zr/Nb nanoscale multilayers are regarded as one of the important candidate materials used in next-generation reactors. Understanding structural evolution induced by ion bombardment is crucial for the evaluation of lifetime performance. Magnetron sputter-deposited Zr/Nb multilayers with a periodicity of 7 nm were subjected to 300 keV He ion irradiation with three different fluences at room temperature. The depth-resolved strain and damage profiles in the Zr/Nb multilayers were investigated by grazing incidence X-ray diffraction. The tensile strain was found in the deposited Zr/Nb films. After He ion irradiation, the intensity of diffraction peaks increased. The change in diffraction peaks depends on He fluence and incident angle. Irradiation-induced pre-existing defect annealing was observed and the ability to recover the microstructure was more significant in the Zr films compared to the Nb films. Furthermore, the efficiency of defect annealing depends on the concentration of pre-existing defects and He fluence. When the He fluence exceeds the one for pre-existing defect annealing, residual defects will be formed, such as 1/3&lt;12¯10&gt; and 1/3&lt;11¯00&gt; dislocation loops in the Zr films and 1/2&lt;111&gt; dislocation loops in the Nb films. Finally, introducing deposited defects and interfaces can improve the radiation resistance of Zr/Nb nanoscale multilayers. These findings can be extended to other multilayers in order to develop candidate materials for fusion and fission systems with high radiation resistance.

Role of low-level void swelling on plasticity and failure in a 33 dpa neutron-irradiated 304 stainless steel

Radiation damage introduces a wide range of defects in structural materials depending on the irradiation conditions. Irradiation-induced loops are known to lead to dislocation-channeling, which reduces the work hardenability and increases the susceptibility to intergranular stress corrosion cracking. In a higher temperature regime, void formation is an integral radiation-induced defect type in addition to radiation-induced loops. Previous bulk mechanical tests suggest that a high level of void swelling can lead to severe embrittlement while a low level of void swelling does not negatively affect intergranular cracking and ductility. This study utilized microscale bicrystal testing to evaluate the role of low-level void swelling on dislocation-channel suppression and intergranular cracking in a 33 dpa neutron-irradiated 304 stainless steels with 2% and 3.7% void swelling. The results provided direct observation on the suppression of dislocation-channel and intergranular cracking in the presence of low-level void swelling.

Irradiation hardening and deuterium retention behaviour of tungsten under synergistic irradiations of 3.5 MeV Fe13+ ions and deuterium plasma

This work investigates the irradiation hardening and deuterium (D) retention behaviour of tungsten (W) under synergistic irradiations of heavy ions and D plasma. 3.5 MeV iron (Fe13+) ion irradiation was performed on recrystallized W samples (RW) to produce displacement damage of 1 dpa. Then, low-energy (38 eV) D plasma exposure was conducted at 500 K. It is found that Fe ion irradiation creates substantial vacancy-type defects and dislocation loops/networks in RW. These irradiation-induced defects not only function as nucleation sites for dislocations that increase the activation probability of new dislocations and suppress the pop-in events, but also act as barriers for dislocations that result in irradiation hardening. Nano-indentation results show that the average hardness of RW-D, RW-Fe and RW-Fe-D increases from 5.26 GPa for RW to 5.28, 6.23 and 6.56 GPa, respectively. The strong interaction between dislocations and high density of damage-induced defects is suggested to be the chief source for the obvious irradiation hardening observed in the synergistic irradiation case. Besides, compared with RW-D, the total D retention in RW-Fe-D is increased by a factor of 2.65. NRA and TDS results suggest that Fe pre-irradiation not only increases D retention within the damage layer (within the first 1.3 μm), but also enhances that beyond the damage layer (> 1.3 μm) before surface blisters are formed. This work further improves the fundamental understanding on the microstructure evolution, D retention and nano-mechanical behaviour of W under synergistic irradiation effect of heavy ions and D plasma.

The Effect of Black-Dot Defects on FeCrAl Radiation Hardening

FeCrAl is regarded as one of the most promising cladding materials for accident-tolerant fuel at nuclear fission reactors due to its comprehensive properties of inherent corrosion resistance, excellent irradiation resistance, high-temperature oxidation resistance, and stress corrosion cracking resistance. In this work, the irradiation response of FeCrAl irradiated by 2.4 MeV He2+ ions with a fluence of 1.1 × 1016 cm−2 at room temperature was studied using X-ray diffraction, transmission electron microscopy, and nanoindentation. The characterization results of structural and mechanical properties showed that only black-dot defects exist in irradiated FeCrAl samples, and that the hardness of the irradiated samples was 11.5% higher than that of the unirradiated samples. Similar to other types of radiation defects, black-dot defects acted as fixed defect obstacles and hindered the movement of slip dislocations moving under the applied load, resulting in a significant increase in the hardness of FeCrAl. Importantly, this work points out that irradiation-induced black-dot defects can significantly affect the mechanical properties of materials, and that their contribution to radiation hardening cannot be ignored.

Raman and optical spectroscopy study of iron-bearing bio-relevant phosphate glasses: Assessment of γ-ray irradiation effects

Glasses with 50P2O5-30CaO-(20 – x)Na2O-xFe2O3 (x = 0.1, 1, 5, 10 mol%) compositions were prepared by melting, and characterized by X-ray diffraction, Raman scattering, and optical spectroscopy. The glasses were subjected to γ-ray irradiation and structural and optical properties evaluated via Raman, absorption, and photoluminescence spectroscopy. The pristine glasses were X-ray amorphous, whereas Raman spectroscopy indicated Fe2O3 depolymerizes glass structure. Absorption spectroscopy showed that iron occurred largely as Fe3+ with Fe2+ revealed at high iron contents. Following γ-ray irradiation, the Raman features were preserved supporting lack of structural alteration. Radiation-induced defects were detected optically for 0.1 mol% Fe2O3 but prevented at higher concentrations. Bluish luminescence ascribed to electron centers was detected under 265 nm excitation for 0.1 mol% Fe2O3 similar to the irradiated host exhibiting ms-scale decay. The optical properties of γ-irradiated glasses at higher iron contents were like the pristine, supporting an increase in glass network resistance against gamma radiation.

X-ray diffraction line profile analysis of defects in neutron-irradiated austenitic stainless steels at low displacement damage levels

Neutron irradiation-induced microstructural changes in variants of austenitic stainless steels were characterized using the Angle Dispersive X-Ray Diffraction (ADXRD) technique with a synchrotron X-ray source of energy 14.968 keV (wavelength ∼0.82833 Å). SS 304 L(N), SS 316 L(N), and SS 316 materials were neutron-irradiated to low displacement damage levels (1–6.75 dpa) in the Fast Breeder Test Reactor at Kalpakkam at temperatures in the range ∼380 °C-400 °C. The effect of the irradiation-induced defects on X-ray diffraction peak profile parameters such as peak broadening, peak shift, and asymmetry were quantified and related to dislocation density. Peak broadening is found to increase as a function of dpa in most of the cases. Considering the highest dpa sample, lattice contraction has been shown by SS 304 L(N) and SS 316 whereas lattice expansion is exhibited by SS 316 L(N). Asymmetry in peak broadening is found to decrease in case of the highest dpa sample. Ferrite volume percent is seen to have increased as a function of dpa in SS 304 L(N) and SS 316 L(N) but not observed in SS 316. SS 304 L(N) has shown a greater propensity towards the austenite to ferrite transformation and 5 dpa sample has about 18 vol% of α-ferrite. However, in case of 5 dpa SS 316 L(N), ferrite content has increased to ∼4 vol% . The increase in yield strength was also related to the dislocation density amongst all three grades of stainless steel. SS 304 L(N) showed saturation in dislocation density and yield strength around 2 dpa whereas SS 316 L(N) showed linear behaviour up to the highest dpa.

The Total Ionizing Dose Effects on Perovskite CsPbBr3 Semiconductor Detector.

Perovskite CsPbBr3 semiconductors exhibit unusually high defect tolerance leading to outstanding and unique optoelectronic properties, demonstrating strong potential for γ-radiation and X-ray detection at room temperature. However, the total dose effects of the perovskite CsPbBr3 must be considered when working in a long-term radiation environment. In this work, the Schottky type of perovskite CsPbBr3 detector was fabricated. Their electrical characteristics and γ-ray response were investigated before and after 60Co γ ray irradiation with 100 and 200 krad (Si) doses. The γ-ray response of the Schottky-type planar CsPbBr3 detector degrades significantly with the increase in total dose. At the total dose of 200 krad(Si), the spectral resolving ability to γ-ray response of the CsPbBr3 detector has disappeared. However, with annealing at room temperature for one week, the device's performance was partially recovered. Therefore, these results indicate that the total dose effects strongly influence the detector performance of the perovskite CsPbBr3 semiconductor. Notably, it is concluded that the radiation-induced defects are not permanent, which could be mitigated even at room temperature. We believe this work could guide the development of perovskite detectors, especially under harsh radiation conditions.

In-Situ TEM Investigation of Helium Implantation in Ni-SiOC Nanocomposites

Ni-SiOC nanocomposites maintain crystal-amorphous dual-phase nanostructures after high-temperature annealing at different temperatures (600 °C, 800 °C and 1000 °C), while the feature sizes of crystal Ni and amorphous SiOC increase with the annealing temperature. Corresponding to the dual-phase nanostructures, Ni-SiOC nanocomposites exhibit a high strength and good plastic flow stability. In this study, we conducted a He implantation in Ni-SiOC nanocomposites at 300 °C by in-situ transmission electron microscope (TEM) irradiation test. In-situ TEM irradiation revealed that both crystal Ni and amorphous SiOC maintain stability under He irradiation. The 600 °C annealed sample presents a better He irradiation resistance, as manifested by a smaller He-bubble size and lower density. Both the grain boundary and crystal-amorphous phase boundary act as a sink to absorb He and irradiation-induced defects in the Ni matrix. More importantly, amorphous SiOC ceramic is immune to He irradiation damage, contributing to the He irradiation resistance of Ni alloy.