Evolution Of Radiation-induced Defects Research Articles

Properties of radiation-induced point defects in austenitic steels: a molecular dynamics study

Abstract Austenitic steels are recognized as excellent structural materials for pressurized water reactors (PWRs) due to their outstanding mechanical properties and radiation resistance. However, compared to the widely studied FeCrNi series of steels, little is known about the radiation resistance of FeCrNiMn steel. In this study, the generation and evolution of radiation-induced defects in FeCrNiMn steel were investigated by molecular dynamics (MD) simulations. The results showed that more defect atoms were produced in the thermal spike stage, but fewer defects survived at the end of the cascades in FeCrNiMn compared to pure Fe. Point defect properties were analyzed by molecular static (MS), and the formation energies of defects in FeCrNiMn were lower than those of pure Fe, while the migration energies were higher. Compared to FeCrNi, FeCrNiMn had smaller migration energies and a larger overlap of vacancy and interstitial migration energies. The low vacancy formation energies and widely overlapping migration energies suggested that the number of point defects in the thermal spike stage was higher, but the possibility of recombination was greater. Additionally, Mn exhibited the smallest interstitial formation energy and migration energy. The difference in defect migration energies revealed that vacancy and interstitial defects migrate through different alloy constituent elements. This study revealed the underlying mechanism for the excellent irradiation resistance of FeCrNiMn.

Development of machine learning and empirical interatomic potentials for the binary Zr-Sn system

Zirconium alloys are pivotal structural materials in nuclear reactors. Enhancing their properties and performance necessitates a profound understanding of the interactions between alloying elements and lattice defects. However, the atomic-level mechanisms that account for the effects of tin (Sn) on irradiation-induced defect evolution in zirconium (Zr) are not well understood yet. To bridge this gap, we conducted extensive first-principles calculations and utilized the obtained data to develop two interatomic potentials for the Zr-Sn system: a Zr-Sn moment tensor machine learning interatomic potential (MTP) and a modified embedded atom potential (MEAM) built upon existing unary ones. The reliability of the constructed potentials was validated through systematically comparing relevant physical properties of pure Zr, Sn metals, and Zr-Sn solid solutions, calculated using these potentials, to values obtained from density functional theory calculations and/or experiments. The developed potentials exhibit high fidelity and accuracy in capturing the behavior of the Zr-Sn system. Their applicability in terms of both reliability and computational efficiency was also discussed. These potentials provide a solid foundation for further exploration of the atomic-scale behavior of the Zr-Sn alloy systems, thereby facilitating the continuous optimization and application of the zirconium alloys.

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.

Effect of Fe11+ ion combined with helium and deuterium plasmas irradiation on the transient thermal shock behaviors of pure and potassium-doped tungsten

The large-scale warm-rolled pure tungsten (PW) and hot-rolled potassium-doped tungsten (W-K) plates were prepared by powder metallurgy technology to meet the demands of International Thermonuclear Experimental Reactor (ITER) in engineering application. In order to evaluate the performance of plasma-facing materials (PFM) in a more realistic service environment as much as possible, the influence of combined irradiation effects of high-energy iron (Fe) ion, deuterium (D) and helium (He) plasmas on the transient thermal shock behavior in the PW and W-K was investigated. The results indicate that the PW can generate more vacancy- and dislocation-type defects than that of W-K after irradiation. The cracking thresholds for un-irradiated PW and W-K are >0.66 and 0.44–0.55 GW/m2, respectively. However, the cracking threshold for both samples decreased to 0.22–0.33 GW/m2 after combined irradiation, indicating that the combined irradiation can significantly reduce the thermal shock resistance of PW and W-K. In addition, according to the statistical results of the crack density and width on the surface of irradiated samples, irradiated W-K owns better thermal shock resistance than irradiated PW under the same thermal shock condition. These results were also verified by the determination of residual stresses. Besides, the relationship between the evolution of irradiation-induced defects and K bubbles is also discussed in detail. These studies can provide an important reference for the identification of future PFM and numerical simulation under combined irradiation plus high thermal load.

Comparison of formation and evolution of radiation-induced defects in pure Ni and Ni–Co–Fe medium-entropy alloy

High-entropy alloys (HEAs) and medium-entropy alloys (MEAs) have attracted a great deal of attention for developing nuclear materials because of their excellent irradiation tolerance. Herein, formation and evolution of radiation-induced defects in NiCoFe MEA and pure Ni are investigated and compared using molecular dynamics simulation. It is observed that the defect recombination rate of ternary NiCoFe MEA is higher than that of pure Ni, which is mainly because, in the process of cascade collision, the energy dissipated through atom displacement decreases with increasing the chemical disorder. Consequently, the heat peak phase lasts longer, and the recombination time of the radiation defects (interstitial atoms and vacancies) is likewise longer, with fewer deleterious defects. Moreover, by studying the formation and evolution of dislocation loops in Ni–Co–Fe alloys and Ni, it is found that the stacking fault energy in Ni–Co–Fe decreases as the elemental composition increases, facilitating the formation of ideal stacking fault tetrahedron structures. Hence, these findings shed new light on studying the formation and evolution of radiation-induced defects in MEAs.

Molecular dynamics simulations of displacement cascade near precipitate in zirconium alloys

Precipitates play an important role in the evolution of irradiation-induced defects and mechanical property of irradiated metals. In this work, the effects of a Zr2Cu precipitate on the production and subsequent evolution of cascade-induced point defects (vacancies and interstitials) in ZrCu alloy were investigated by molecular dynamics simulations at room temperature. The simulation results show that the precipitate increases the number of residual point defects at the end of cascade. However, most of the residual defects reside in the precipitate and near precipitate boundary. In the matrix, more interstitials survive than vacancies. In addition, a defect-free region is seen in matrix and near precipitate boundary, indicating that the precipitate boundary traps vacancies/interstitials in matrix and retains them in the precipitate. To explain these observations, the formation energy and migration energy of point defects are calculated. The formation energies of both vacancy and interstitial decrease as they are moving to the precipitate boundary from matrix and precipitate, indicating that they can be absorbed by precipitate boundary and result in the defect-free region. However, the migration barrier of vacancy is lower than that of interstitial. Therefore, more vacancies in matrix are absorbed. The current work provides new insights into understanding the irradiation effects in zirconium alloys.

On the strain effect in dislocation loops produced by self-ion irradiation to emulate in-service reactor condition

A series of experiments combining deformation, irradiation damage, and high temperature have been carried out. First, small tensile samples (less than 500 μm thick) will be subjected to constant deformation (supported by a suitably instrumented mini tensile machine) during irradiation of Fe ions at different temperatures (room temperature, RT, and 450 °C). Subsequently, FIB (Focused Ion Beam) samples will be extracted from the irradiated specimen, and an EBSD (Electron Backscatter Diffraction) and TEM (Transmission Electron Microscopy) characterisation will be carried out. The material studied is a high purity Fe- BCC model alloy. The objective is to determine the possible synergy between the effect of deformation and temperature during irradiation on the evolution of radiation-induced defects, whereby the chosen dose allows us to determine a potential bias for the generation of such defects.

Influence of helium on the evolution of irradiation-induced defects in tungsten: An object kinetic Monte Carlo simulation* *Project supported by the Science Challenge Project (Grant No. TZ2018002), the National Natural Science Foundation of China (Grant No. 11905135), and the National MCF Energy R&D Program of China (Grant No. 2018YFE0308103).

Understanding the evolution of irradiation-induced defects is of critical importance for the performance estimation of nuclear materials under irradiation. Hereby, we systematically investigate the influence of He on the evolution of Frenkel pairs and collision cascades in tungsten (W) via using the object kinetic Monte Carlo (OKMC) method. Our findings suggest that the presence of He has significant effect on the evolution of irradiation-induced defects. On the one hand, the presence of He can facilitate the recombination of vacancies and self-interstitial atoms (SIAs) in W. This can be attributed to the formation of immobile He-SIA complexes, which increases the annihilation probability of vacancies and SIAs. On the other hand, due to the high stability and low mobility of He-vacancy complexes, the growth of large vacancy clusters in W is kinetically suppressed by He addition. Specially, in comparison with the injection of collision cascades and He in sequential way at 1223 K, the average sizes of surviving vacancy clusters in W via simultaneous way are smaller, which is in good agreement with previous experimental observations. These results advocate that the impurity with low concentration has significant effect on the evolution of irradiation-induced defects in materials, and contributes to our understanding of W performance under irradiation.

Enhanced radiation tolerance of the Ni-Co-Cr-Fe high-entropy alloy as revealed from primary damage

High-entropy alloys (HEAs) have received much attention for the development of nuclear materials because of their excellent irradiation tolerance. In the present study, the generation and evolution of irradiation-induced defects in the NiCoCrFe HEA were investigated by molecular dynamics (MD) simulations to understand the mechanisms of its irradiation tolerance compared with bulk Ni. The displacement cascades were simulated for the energies of primary knock-on atoms (PKA) ranging from 10 to 50 keV to understand the irradiation resistance in HEAs. In general, there are more displaced atoms produced in the thermal spike phase, but fewer defects survived at the end of the cascades in the NiCoCrFe alloy than in Ni. Both interstitial and vacancy clusters increase in size or number with increasing PKA energy in both materials, but they do so more slowly in the NiCoCrFe HEA. The delayed damage accumulations in the NiCoCrFe HEA are attributed to the high defect recombination caused by the following two mechanisms. First, the enhanced thermal spike and the low thermal conductivity of HEAs for heat dissipation result in the higher efficiency of defect recombination. Furthermore, the substantially small binding energies of interstitial loops in the NiCoCrFe HEA, as compared with those in Ni, are responsible for the delayed interstitial clustering in the NiCoCrFe HEA.

Fullerene-like defects in high-temperature neutron-irradiated nuclear graphite

Irradiation-induced defect evolution in graphite is particularly important for its application in graphite-moderated nuclear reactors. The evolution of defects directly influences macroscopically observed property changes in irradiated nuclear graphite which, in turn, can govern the lifetime of graphite components. This article reports novel defect structures and the irradiation response of microstructural features occurring in high-temperature irradiated nuclear graphite IG-110. High resolution transmission electron microscopy (HRTEM) was used to characterize specimens neutron-irradiated at a high temperature (≥800 °C) at doses of 1.73 and 3.56 atomic displacements per atom (dpa). Concentric shelled and fullerene-like defects were found to result in swelling along the c-axis and contraction along the a/b-axis of crystallites. Furthermore, such defects are shown to occur within, and partially fill, Mrozowski cracks prior to turnaround dose. In addition, in situ TEM under similar irradiation conditions was used to capture the real-time dynamic evolution of defects, providing unambiguous analysis of the evolution of the graphite structures during irradiation. Results suggest the mainstream theory for radiation damage in nuclear graphite (which assumes additional basal plane formation as the sole reason) to be an incorrect interpretation of defect evolution contributing to irradiation-induced property changes at higher temperatures.

The surface topography and microstructure of self-sintered nanopore graphite by Xe ions irradiation

Nanopore graphite prepared by self-sintered pitch coke for molten salt reactor was irradiated by 7 MeV Xe26+ to study the irradiation behaviors. The hardness and Young's modulus of self-sintered nanopore graphite (SSNG) were increased due to an increase in defects and close of micro-gap. The SSNG surface remained intact after irradiation. The pore and the gap between the filler particles of SSNG shrink due to the expansion of the filler particles. The microstructure evolution characterization of SSNG samples by the grazing incidence X-ray diffraction showed the increase in interlayer spacing and the expansion of crystallite size along c-axis, which caused the expansion of the filler particles. The defect density increase is the main reason for mechanical properties changes. The high-resolution transmission electron microscopy provides clear and visual evidence for this and showed the evolution of dislocation and interstitial loop. The microstructure evolution characterization of the SSNG is at the atomic scale after irradiation at different depths. The evolution of irradiation-induced defects in SSNG causes the change of microscopic morphology, and the evolution of morphology and defects plays an important role in the change of mechanical properties.

Enhanced Radiation Tolerance of the Ni-Co-Cr-Fe High-Entropy Alloy as Revealed from Primary Damage

High-entropy alloys (HEAs) have received much attention for the development of nuclear materials because of their excellent irradiation tolerance. In the present study, the creation and evolution of irradiation-induced defects in the NiCoCrFe HEA were investigated by molecular dynamics (MD) simulations to understand the mechanisms of their irradiation tolerance compared with bulk Ni. The displacement cascades were simulated for the energies of primary knock-on atoms (PKA) ranging from 10 to 50 keV to understand the irradiation resistance in HEAs. In general, there are more displacements produced in the thermal spike phase, but fewer defects survived at the end of the cascades in the NiCoCrFe alloy than in Ni. Both interstitial and vacancy clusters increase in size or number with increasing PKA energy in both materials, but they do so more slowly in the NiCoCrFe HEA. The delayed damage accumulations in the NiCoCrFe HEA are attributed to the high defect recombination caused by the following two mechanisms: the enhanced thermal spike and the low thermal conductivity of HEAs for heat dissipation result in the higher efficiency of defect recombination, and the substantially small binding energies of interstitial loops in the NiCoCrFe HEA, when compared with those in Ni, are responsible for the delayed interstitial clustering in the NiCoCrFe HEA.

Atomistic simulation study of clustering and evolution of irradiation-induced defects in zirconium

Zirconium (Zr) alloys have been widely used as structural materials for in-core components in water-cooled nuclear reactors. During normal operation, these materials are exposed to a neutron flux in the core of the reactor, resulting in material degradation such as irradiation-induced anisotropic growth, thus affecting their performance in the long-term. Experimental and theoretical studies have shown that irradiation-induced defects in zirconium lead to the formation of defect clusters and loops. The anisotropy in the migration of defects has been suggested to play an important role in irradiation growth in pure Zr and its alloys. However, the mechanisms that govern the microstructural evolution that lead to the observed anisotropic growth of Zr is still unclear. In the present work, we perform a molecular dynamics simulation study of irradiation-induced lattice defects in Zr to investigate the formation of clusters and loops. Irradiation-induced damage is modeled by constrained stochastic formation of vacancies and self-interstitial atoms in bulk Zr. Using this approach, the formation and evolution of defect clusters and loops were determined. The dynamic properties of lattice defect structures were investigated through the evaluation of their migration and diffusivity. We found that the diffusivity of vacancy and interstitial clusters is anisotropic and slow, while the diffusivity of large loops is relatively high and confined to the ⟨a⟩ plane.

Evolution of cellular dislocation structures and defects in additively manufactured austenitic stainless steel under ion irradiation

The evolution of irradiation-induced defects in additively manufactured (AM) austenitic stainless steel was investigated in situ by 1 MeV Kr ion irradiation at 450 and 600 °C in a transmission electron microscope. Cellular dislocation structure in AM steel act as sink/trap sites for the irradiation-induced defects, resulting in the lower density and smaller dislocation loops in AM steel than conventional forged (CF) steel at 450 °C. The higher stacking fault energy and local stress concentration induced by cellular dislocation structure in AM steel promotes the unfaulting process and the formation of network dislocation at 600 °C.

Defect evolution in Ni and NiCoCr by in situ 2.8 MeV Au irradiation

The evolution of radiation-induced defects in Ni and the single-phase concentrated solid solution alloy, NiCoCr, were investigated during in situ 2.8 MeV Au ion irradiation and post-irradiation analysis using transmission electron microscopy. Compared to Ni, both the size and area density of defect clusters decreased in NiCoCr under the same irradiation conditions, suggesting that the chemical complexity, i.e., randomness of lattice site occupations, of NiCoCr suppressed radiation-induced damage. One-dimensional glide of defect clusters was observed in Ni but not in the NiCoCr alloy. The structural nature of small defect clusters in NiCoCr were further investigated using high-angle annular dark field scanning transmission electron microscopy.

Neutron irradiation-induced defects in Cd0.9Zn0.1Te:In crystals

The survivability of γ-ray detector in radiation environment of energetic neutrons is a key issue for its application in spacecraft, accelerator, nuclear reactor, etc., especially for the case of CdZnTe detectors. Nevertheless, very little attention has been paid to the radiation damage mechanism of detectors and the evolution of radiation-induced defects in the detector medium, i.e. Cd0.9Zn0.1Te:In crystals. We proposed and simulated the radiation defects type and amount of Cd0.9Zn0.1Te:In crystals after neutron irradiation with the Monte-Carlo software SRIM. The calculated data are also compared with experimental results of detector performance and the crystal defects obtained using thermally stimulated current (TSC) measurement. We also discuss the influence of defect concentrations on crystal resistivity and detector energy resolution. It was shown that the concentration of vacancies, interstitials, and their related defects increased after neutron irradiation, which could be the reason for the worsening of Cd0.9Zn0.1Te:In γ-ray detector performance after the neutron irradiation.

Proton irradiation for radiation-induced changes in microstructures and mechanical properties of austenitic stainless steel

We report on microstructural and mechanical property changes as a function of radiation damage value in proton-irradiated austenitic stainless steel by means of advanced characterization techniques. The microstructural changes in proton-irradiated austenitic stainless steel were analyzed by transmission electron microscopy for observation of radiation-induced defects as well as the measurement of the chemical composition at grain boundaries. The radiation hardening after the proton irradiation was characterized by nano indentation for changes in hardness profiles with radiation damage.Various transition points for microstructural and mechanical property changes under proton irradiation are analyzed via material characterization of proton-irradiated austenitic stainless steels. The saturation is expected to occur at approximately 10 displacements per atom (dpa) for the radiation-induced segregation of Cr, Ni, and P and approximately 2.5 dpa for radiation hardening. The cavity formation is observed to occur at hydrogen concentration levels greater than 5E5 atomic parts per million (appm) H. It is also found that the transition from black dot to Frank loop happened above approximately 1 dpa.Profiles of radiation-induced segregation and radiation hardening as a function of dpa can be extended to the high irradiation condition, and can be compared with experimental data for neutron irradiation-induced segregation and radiation hardening. The radiation-induced segregation after the proton irradiation at 360 °C are in good agreement with that after neutron irradiation. On the other hand, it is observed that the evolution of radiation-induced defects and the corresponding radiation hardening exhibit sooner, that appears to be because of the dose rate effect.

Evolution of Dislocation Loops Induced by Different Hydrogen Irradiation Conditions in Reduced-Activation Martensitic Steel.

Hydrogen can be induced in various ways into reduced-activation ferritic/martensitic (RAFM) steels when they are used as structural materials for advanced nuclear systems. However, because of the fast diffusion of hydrogen in metals, the effect of hydrogen on the evolution of irradiation-induced defects was almost neglected. In the present work, the effect of hydrogen on the evolution of dislocation loops was investigated using a transmission electron microscope. Specimens of reduced-activation ferritic/martensitic (RAFM) steels were irradiated with hydrogen ions to 5 × 1020 H+ • m−2 at 523–823 K, and to 1 × 1020 H+ • m−2 − 5 × 1020 H+ • m−2 at 723 K. The experimental results reveal that there is an optimum temperature for dislocation loop growth, which is ~723 K, and it is greater than the reported values for neutron irradiations. Surprisingly, the sizes of the loops produced by hydrogen ions, namely, 93 nm and 286 nm for the mean and maximum value, respectively, at the peak dose of 0.16 dpa under 723 K, are much larger than that produced by neutrons and heavy ions at the same damage level and temperature. The results indicate that hydrogen could enhance the growth of loops. Moreover, 47.3% a0 <111> and 52.7% a0 <100> loops were observed at 523 K, but a0 <111> loops disappeared and only a0 <100> loops existed above 623 K. Compared with the neutron and ion irradiations, the presence of hydrogen promoted the formation of a0 <100> loops.

Molecular dynamics study of hydrogen-vacancy interactions in α-zirconium

Hydrogen is known to have detrimental effects on the mechanical properties of zirconium and its alloys, mainly due the formation of brittle hydride phases. The behaviour of hydrogen dissolved in zirconium is not fully understood at the atomic scale. In particular, the influence of hydrogen on the behaviour of defects in zirconium still needs to be clarified. In this paper, we report a molecular dynamics study of hydrogen behaviour in zirconium using an improved Zr-H interatomic potential. We determined the diffusivity of hydrogen in zirconium, and the interaction of hydrogen atoms with a single or a pair of atomic vacancies. The computed hydrogen diffusion is found to be in good agreement with experimentally determined values. Our results also indicate that hydrogen interactions with vacancies in zirconium promote vacancy clustering. Moreover, hydrogen is found to substantially decrease the mobility of divacancies. From the present simulation findings, it is concluded that the presence and concentration of hydrogen impurities may have a significant influence on the evolution of irradiation-induced defects in zirconium alloys.

Effects of Fe concentration on the ion-irradiation induced defect evolution and hardening in Ni-Fe solid solution alloys

Understanding alloying effects on the irradiation response of structural materials is pivotal in nuclear engineering. To systematically explore the effects of Fe concentration on the irradiation-induced defect evolution and hardening in face-centered cubic Ni-Fe binary solid solution alloys, single crystalline Ni-xFe (x = 0–60 at%) alloys have been grown and irradiated with 1.5 MeV Ni ions. The irradiations have been performed over a wide range of fluences from 3 × 1013 to 3 × 1016 cm−2 at room temperature. Ion channeling technique has shown reduced damage accumulation with increasing Fe concentration in the low fluence regime, which is consistent to the results from molecular dynamic simulations. No irradiation-induced compositional segregation was observed in atom probe tomography within the detection limit, even in the samples irradiated with high fluence Ni ions. Transmission electron microscopy analyses have further demonstrated that the defect size significantly decreases with increasing Fe concentration, indicating a delay in defect evolution. Furthermore, irradiation induced hardening has been measured by nanoindentation tests. Ni and the Ni-Fe alloys have largely different initial hardness, but they all follow a similar trend for the increase of hardness as a function of irradiation fluence.