This study investigates the influence of nanoscale oxide dispersoid characteristics on irradiation hardening in FeCrAl-based oxide dispersion strengthened (ODS) ferritic steels. Four ODS alloys with systematically varied oxide size and number density-achieved through Al and Zr additions-along with a conventional HT9 steel, were irradiated with Fe-56 ions to 0.8 dpa at 563 K. Vickers micro-hardness tests revealed that the Zr-added and Al-free ODS steels exhibited significantly lower irradiation hardening than their Al-containing counterparts, while HT9 showed the highest hardening. This enhanced radiation tolerance is attributed to the high number density of fine (similar to 2-10 nm) oxide particles, which act as efficient sinks for point defects, suppressing dislocation loop formation. A power-law correlation was established between the oxide-related sink strength and the irradiation-induced hardness increment. The improved irradiation resistance is primarily attributed to the increased sink strength. Assuming interstitial-type dislocation loops govern hardening, the dispersed barrier hardening (DBH) model yielded predictions consistent with experimental data, confirming the critical role of microstructure in defect mitigation.

For the purpose of clarifying the basic mechanism of material failure caused by irradiation damage, the microstructures of materials under extreme radiation conditions were evaluated by transmission electron microscopy (TEM). In the present study, swift heavy ion irradiation (345MeV Fe ions) was employed to introduce dislocation loops in the MA956 ODS steel, and the subsequent S/TEM characterization was conducted. While TEM has traditionally used as the primary tool for dislocation loops imaging, recent advancements in scanning transmission electron microscopy (STEM) technology have increased interest in defects imaging. The principal STEM-based approaches for dislocation/dislocation loops characterizing include diffraction-selected on-zone STEM (DsoZ-STEM), diffraction-contrast imaging STEM (DCI-STEM), and weak-beam dark-field STEM (WBDF-STEM). This study compared these STEM methods for dislocation loops imaging, demonstrating the usage of objective lens aperture in the enhancement of dislocation loops contrast. Experimental findings indicated that DCI-STEM method performed under a systematic row diffraction condition with sg>0 surpassed the other approaches, yielding an enhanced contrast of dislocation loops with a clean background free of dynamical diffraction artifacts.

In this study, Molecular Dynamics (MD) and Object Kinetic Monte Carlo (OKMC) simulations are adopted to explore the long-term defect evolution behavior of Fe-9Cr alloys. The effects of injected Fe ion concentrations (0–1000 appm/dpa) on irradiated microstructures are systematically evaluated at two typical irradiation temperatures of 573 K and 673 K. The results reveal that the modulation effects of injected self-ions are more prominent under high-temperature and high-dose irradiation conditions. At irradiation doses below 1 dpa, injected ions exert a minor influence on matrix defects. As the irradiation dose increases, such an effect becomes considerable: at 10 dpa, the density of large defect clusters varies by more than one order of magnitude under different injected ion concentrations. High-concentration injected ions substantially promote the nucleation and growth of self-interstitial atom (SIA) clusters while inhibiting the formation and evolution of voids. At high irradiation doses, high ion injection concentrations significantly raise the production ratio of <100>-type SIA loops. Furthermore, this work also elucidates the fundamental mechanisms by which injected ions affect the diffusion and segregation of solute atoms.

FeCrAl ODS steel is the promising candidate cladding material for the advanced nuclear systems. The effects of alloy compositions on oxides and the strengthening are vital to materials design and development, which deserve in-depth investigation. Here, detailed crystal structures and interface coherency of totally more than 1000 oxides in Fe-15Cr-3.3Al ODS model alloys with 0.51 wt% Zr and 0.46 wt% Hf addition (denoted as P1 and P2, respectively) were investigated. The oxides in P1 are mainly trigonal Y4Zr3O12 and Y4Al2O9 (YAM, yttrium aluminum monoclinic) while in P2, they are Y2Hf2O7 (pyrochlore) and YAM. The Y2Hf2O7 oxides occupy 74.5 % of oxides in P2 while Y4Zr3O12 oxides account for 65.9 % in P1, manifesting that 0.46 wt% Hf addition can greatly suppress the formation of YAM oxides than 0.51 wt% Zr addition. The mean diameter of oxides in P2 is smaller than P1 due to the formation of smaller Y2Hf2O7 and YAM oxides. Almost all Y4Zr3O12 and Y2Hf2O7 oxides are coherent with the matrix whereas some YAM oxides show a semi-coherency relationship. The semi-coherent YAM oxides take a percentage of 5.0 % in P2, in contrast to that of 8.9 % in P1, indicating an improvement of oxide coherency by Hf addition. A size dependence of coherency of YAM oxides was observed in both specimens. The strengthening mechanism of ODS model alloys is analyzed by using dispersion barrier hardening (DBH) model. The calculated barrier strength factor, alpha for coherent oxides is 0.20, which is smaller than that for semi-coherent oxides, 0.36.

Oxide dispersion strengthened (ODS) steel is considered a promising candidate material for advanced nuclear systems. The stability of oxides plays a critical role in maintaining mechanical properties during service. In a fission environment, structural materials experience damage not only from neutrons but also from fission products. This study focuses on the impact of electronic energy loss on the structural damage of ODS steel MA956 under 345 MeV Fe ion irradiation at fluences of 1012-1014 ions/cm2 at different temperatures. Detectable hardening was observed at an extremely low fluence of 2.0 x 1012 ions/cm2 and lower temperature (-70--50 degrees C). An evident hardening was observed under the irradiation at a fluence of 4.5 x 1014 ions/cm2 and a temperature of 160-185 degrees C. The hardness values of the irradiated samples at low temperatures showed only a slight increase or no changes after annealing at 500 degrees C compared to the as-irradiated samples, indicating that the damaged structures slightly extended or survived the annealing treatment. TEM results indicated that the size and density of oxides slightly increased in the irradiated samples, accompanied by a narrower size distribution. The proportion of oxides smaller than 12 nm decreased, while a higher proportion was observed in the size range of 12 to 20 nm in the irradiated samples, with no apparent differences in matrix damage, except for the irradiated samples at 160-185 degrees C, where dislocation loops were observed. Estimations based on the dispersed barrier hardening (DBH) model suggested that the detectable hardening observed at such low irradiation fluence was attributed to changes in the oxides, due to the electronic energy loss. The evident hardening in the irradiated samples at 160-185 degrees C resulted from the changes in the oxides and the formation of dislocation loops.

Herein, we used a multiscale simulation method to investigate the variation rule of the long-term feature of irradiation defects versus primary knock-on atom (PKA) energies (5 keV–100 keV) in dilute FeMnNi alloys. Simulation results presented a significant effect of PKA energy on irradiation microstructure, the density of large defect clusters increases significantly with increasing PKA energy, and the difference in the number density of visible defect clusters can reach several times to about one order of magnitude at the dose rate of 1 × 10−4dpa/s. The migration events of point defects (PD) decrease significantly with increasing PKA energy, suggesting that a higher PD flux at lower PKA energy could probably enhance the radiation-induced segregation (RIS) in alloys. Additionally, since the generation mechanism involves specific SIA-cluster sizes, the production ratio of <100>-type SIA loops could also influenced by PKA energies. Our simulation results align with experimental differences in the microstructure of light and heavy-ion irradiated alloys. This work clarified the mechanism of PKA energy effects in dilute alloys and provides a certain scientific basis for the comparison and emulation of different particle irradiation.

Silicon carbide (SiC) and its composites are promising structural materials for advanced nuclear energy applications. Due to the lack of advanced nuclear energy devices, ion beam irradiation is widely used to emulate reactor neutron irradiation. At the same time, different types of ion beam irradiation could produce different primary knock-on atom (PKA) energy spectra. PKA energy determines cascade damage sizes and defect clustering distributions, which may influence the irradiated materials' long-term microstructural evolution and mechanical properties. This work used SRIM and Geant4 software to investigate the PKA characteristic produced by 1 MeV different noble gas ions and neutrons in the silicon carbide. The PKA energy spectra and weighted energy spectra of C and Si are calculated, respectively. The simulation results show that the PKA energy spectra calculated by the two kinds of software have obvious differences, but the weighted average PKA energies are close to each other. Simulation results verified that the weighted average PKA energy of Kr and Xe ion irradiation is close to that weighted average PKA energy spectrum for neutron irradiation of advanced reactors. The simulation results provide scientific references for understanding the difference in irradiation effects of different types of ions and also provide fundamental bases for the simulation of primary defect damage and long-term defect evolution in irradiated SiC.

The Nano-indentation deformation behavior of the heavy ion irradiated RAFMs is investigated in the present work. Specimens of the CLF-1 steel were irradiated with 123.4 MeV 20Ne ion at both -50 °C and 400 °C, respectively, to achieve a mean displacement damage level of 0.15 dpa. A quasi-uniform distribution of the displacement damage was produced in the specimens by utilizing an energy degrader at the irradiation chamber. The results of nano-indentation experiments indicate that during the loading stage, the irradiated materials exhibit a lower strain rate, but during the holding stage after reaching the maximum load, the irradiated materials show a higher strain rate than the un-irradiated material. Transmission electron microscopy observation showed that the low-temperature irradiation produced a high density of irradiation-induced dislocation loops, which impeded the glide of dislocation lines during the loading stage. However, as the load increases gradually to the holding stage, the dislocation lines can overcome the pinning by the dislocation loops. The strain that did not occur due to the pinned effect of the dislocation loops in the loading stage was released during the holding stage, resulting in a higher strain rate in the irradiated specimen during the holding stage. Based on the Peierls mechanism, a quantitative relationship between the number density of the irradiation-induced dislocation loops and the strain rate during the holding stage was established. The calculated results agree well with the experimentally obtained strain rate during the holding stage.

Vanadium (V) alloys are attractive candidate materials for tritium breeding modules (TBM) of blankets in fusion reactors. However, irradiation-induced hardening and embrittlement give rise to significant challenges for the application of V alloys in advanced nuclear reactors, particularly in the presence of inert gas atoms. In this study, V-5Cr-5Ti alloys were irradiated by 122 MeV Ne ions (at doses of 0.1 and 0.3 dpa) and 352 MeV Fe ions (at doses of 0.06, 0.20 and 0.35 dpa) at a temperature around 100 degrees C, and subsequently annealed at the temperature ranging from 212 to 500 degrees C. During the annealing process, the hardening produced by Fe ion irradiation started to recover at a temperature higher than 212 degrees C and was eliminated at a temperature higher than 375 degrees C. For the sample with Ne ions irradiation, the irradiation hardening of the sample with 0.3 dpa can be sustained at a temperature up to 450 C-o, but an evident reduction of hardening in the sample with 0.1 dpa was observed after annealing at 212 C-o. This suggested that the existence of inert gas atoms had a thermal stabilization effect on irradiation hardening, but it required a critical atom concentration. The different annealing behaviors of irradiation hardening at different doses under Ne ions irradiation were considered to be from the coupling effect of bubble growth and the interaction between interstitial clusters and Ne atoms.

The precipitation of solutes is the primary factor that leads to the occurrence of irradiation hardening and embrittlement, thus limiting the service life of reactor pressure vessel (RPV) steels. In this study, four model alloys (Fe-1.35Mn-0.75Ni, Fe-1.35Mn-0.20Si, Fe-0.75Ni-0.20Si and Fe-1.35Mn-0.75Ni-0.20Si) were used to investigate the influence of solute elements on grain size, hardness, irradiation hardening, and defect clustering behavior. The materials were irradiated with 3.5 MeV Fe ions to doses of 0.1, 0.3, 1.0, and 3.0 dpa at 290 degrees C. Results indicated that the addition of solute Mn resulted in grain refinement while the addition of solutes Ni and Si led to grain coarsening. The strength of the model alloys was affected by the change in grain size. A definite dose rate effect was found in all four model alloys under irradiation, where irradiation with a higher dose rate led to a lower irradiation hardening effect and irradiation with a lower dose rate induced a higher hardening effect. Atom probe tomography (APT) results revealed that the precipitated solutes were always coupled with interstitial defect clusters. Solute Mn had a strong influence on cluster nucleation and a weak influence on cluster size. Solutes Si and Ni significantly promoted the nucleation and growth of clusters, and Ni had a greater influence on the nucleation and growth of clusters than Si. Meanwhile, solute Si consistently exhibited a low fraction in the clusters when the solute Mn element was presented in the alloys.

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Gas-filled bubbles present significant challenges in materials as they make materials porous and weaken the strength of materials. Here, argon bubbles that are formed during the fabrication of oxide dispersion strengthened (ODS) steels are investigated. The Ar bubbles are attached to nano-oxides which are dispersed in the matrix. MA956 ODS specimen was irradiated with 362 MeV 129Xe26+ at room temperature to about 7 dpa at peak damage. The behavior of Ar bubbles in different irradiation depths was investigated by using transmission electron microscopy (TEM) via cross-section specimens. The results show that Ar-bubbles in the peak damage and ions deposition region experience an apparent dissolution. Further, we simulate the behavior of isolated Ar bubble and attached Ar bubble to oxide under primary knock-on atoms (PKA) with different energy and numbers by using molecular dynamics (MD). The results show that high energy PKA (exceeding 30 keV) and multiple cascades can contribute to the effective ejection of Ar atoms from bubbles. Meanwhile, the interface between oxide and bubble contributes less to the bubble dissolution. Besides that, the effects of thermal ejection and excess self-interstitial atoms generated from the implantation of foreign atoms on the dissolution of Ar bubbles are discussed.

The precipitation of solutes is a major cause of irradiation hardening and embrittlement limiting the service life of reactor pressure vessel (RPV) steels. Impurities play a significant role in the formation of precipitation in RPV materials. In this study, the effects of carbon on cluster formation and irradiation hardening were investigated in an RPV alloy Fe-1.35Mn-0.75Ni using C and Fe ions irradiation at 290 °C. Nanoindentation results showed that C ion irradiation led to less hardening below 1.0 dpa, with hardening continuing to increase gradually at higher doses, while it was saturated under Fe ion irradiation. Atom probe tomography revealed a broad size distribution of Ni–Mn clusters under Fe ion irradiation, contrasting a narrower size distribution of small Ni–Mn clusters under C ion irradiation. Further analysis indicated the influence of carbon on the cluster formation, with solute-precipitated defects dominating under C ion irradiation but interstitial clusters dominating under Fe ion irradiation. Simulations suggested that carbon significantly affected solute nucleation, with defect clusters displaying smaller size and higher density as carbon concentration increased. The higher hardening at doses above 1.0 dpa was attributed to a substantial increase in the number density of defect clusters when carbon was present in the matrix.

The present study aimed to investigate the mechanical property of heavy ions irradiated Reduced Activation Ferrite/Martensitic (RAFM) steels. The CLF-1 steel was irradiated by 33.5 MeV Fe ions at room temperature, resulting a peak displacement damage of 20 displacements per atom (dpa). Experimental analysis was conducted using Electron Backscatter Diffraction (EBSD) mapping-assisted Focused Ion Beam (FIB) technique to prepare micro-pillars oriented along crystal directions <0 0 1>. Subsequently, these fabricated micro-pillars underwent compression tests utilizing a flat indenter within the Nano-indentation setup. Engineering Stress-strain curves were generated for specimens both un-irradiated and irradiated. The yield strength of the un-irradiated and irradiated specimens in the <0 0 1> directions was determined by using the engineering stress-strain curves. In addition, the critical resolved shear stress (CRSS) of the un-irradiated and irradiated specimens was calculated. Microstructural analysis revealed that the formation of dense and fine dislocation loops were appeared in the material after irradiation. By combining the results of microstructure analysis with the dispersion barrier hardening model, the changes in CRSS caused by irradiation were calculated. Furthermore, the number of dislocation lines participating in deformation at a strain of 20 % was quantified, with the irradiated specimen exhibiting an increase of 33 % in comparison to the un-irradiated specimen.

Irradiation defects of V-5Cr-5Ti alloys with different pre-irradiation treatment after irradiation to 0.5 dpa at 100 degrees C with energetic Fe ions were investigated. For the cold-worked specimens, dislocation loops decrease in size while increasing in number density with the increase of the deformation percentage. For the annealed specimens, dislocation loops have minor difference in size whilst the number density decreases with the increasing defor-mation percentage. Meanwhile, dislocation loops decorating dislocations were observed. The obstacle strength factor (alpha) is calculated according to the dispersion barrier hardening model (DBH). For the cold-worked speci-mens, alpha increases with increasing size of dislocation loops. A linear relation, alpha = 0.075 * d -0.074 is established. At a similar size, dislocation loops in the annealed specimens have a greater alpha than the cold-worked specimens, possibly due to the lock of the tangled network dislocations and/or the hinder of the segmented movement of the tangled network dislocations by dislocation loops.

This work studies the defect features in a dilute FeMnNi alloy by an Object Kinetic Monte Carlo (OKMC) model based on the "grey-alloy" method. The dose rate effect is studied at 573 K in a wide range of dose rates from 10−8 to 10−4 displacement per atom (dpa)/s and demonstrates that the density of defect clusters rises while the average size of defect clusters decreases with increasing dose rate. However, the dose-rate effect decreases with increasing irradiation dose. The model considered two realistic mechanisms for producing <100>-type self-interstitial atom (SIA) loops and gave reasonable production ratios compared with experimental results. Our simulation shows that the proportion of <100>-type SIA loops could change obviously with the dose rate, influencing hardening prediction for various dose rates irradiation. We also investigated ways to compensate for the dose rate effect. The simulation results verified that about a 100 K temperature shift at a high dose rate of 1 × 10−4 dpa/s could produce similar irradiation microstructures to a lower dose rate of 1 × 10−7 dpa/s irradiation, including matrix defects and deduced solute migration events. The work brings new insight into the OKMC modeling and the dose rate effect of the Fe-based alloys.

The dose rate effect significantly impacts accelerated assessment experiments of the radiation resistance of critical structural materials in nuclear reactors. This study investigated the role of the dose rate on the irradiation hardening and defect cluster formation in a low-Cu reactor pressure vessel (RPV) steel A508-3 under 3.5 MeV Fe13+ ions irradiation at 290 °C, with chosen dose rates of 2.8 × 10–5, 1.4 × 10–4 and 2.8 × 10–4 dpa/s. Nano-indentation and atom probe tomography (APT) techniques were employed to characterize the mechanical property changes and defect clustering behavior, respectively. Nanoindentation results revealed that for a fixed dose, greater hardening was observed in the sample irradiated at a lower dose rate, and conversely, less hardening at a higher dose rate. APT results for samples with 3.0 dpa demonstrated that defect clusters exhibited characteristics of small size and low density under irradiation at a dose rate of 2.8 × 10–4 dpa/s, and large size and high density at a dose rate of 2.8 × 10–5 dpa/s. The ratio of solutes to Fe elements in a cluster also varied with changes in dose rate. The change in hardening, cluster size, number density and cluster composition suggested that the evolution of defect clusters was primarily driven by ballistic mixing, rather than radiation-enhanced diffusion, under heavy ions irradiation within the dose rate range of 2.8 × 10–5–2.8 × 10–4 dpa/s.

目的 研究在563 K的温度条件下,高能重离子辐照导致国产RPV钢A508-3的硬化行为.方法 使用回旋加速器提供的352.8 MeV Fe21+对A508-3钢试样进行辐照,使其依次达到0.15、0.30、1.50 dpa的损伤水平.借助辐照终端的梯度减能装置,在样品表面至25μm的深度范围内产生一个准均匀分布的原子离位损伤坪区,并使用纳米压痕仪和维氏显微硬度计测量了试样的硬度.结果 考虑到压痕尺寸效应的影响,采用Nix-Gao模型对硬度数据进行拟合,由于高能Fe离子产生了较厚的损伤区,辐照试样中的软基效应明显减弱.辐照后的试样出现明显的硬化现象,且随着辐照剂量的增大,硬化增量有饱和趋势.通过辐照前和辐照至0.15 dpa试样的微硬度数据,得出其与纳米硬度之间的线性关系(Hv0=0.85 H0).结论 借助A508-3辐照前后的硬度数据,计算发现试样的屈服强度增量与辐照剂量之间存在一种幂函数关系,这为A508-3钢中子-离子辐照硬化的映射关系研究提供了基础数据.

Herein, the defect features under irradiation in the Fe-C system are studied by an Object Kinetic Monte Carlo (OKMC) model. The model was based on recent parameters and was validated by comparing the numerical estimates with the experimentally obtained defect features in neutron-irradiated iron. Systematic simulations of defects evolution at 70 °C in a wide range of dose rates from 10−8 to 10−4 dpa/s were carried out. The simulation results demonstrate that at the lower dose range (< 0.02 dpa), a higher dose rate irradiation leads to a higher interstitial-loop density and irradiation hardening. In contrast, the dose rate does not obviously influence the defect features at the high dose range. Defect features under two carbon concentrations 50 and 100 appm show a similar dose-rate effect, and a higher carbon concentration leads to a higher interstitial loop density and irradiation hardening. We attribute the dose rate effect change with dose to the competition between sinks absorption and SIAs-vacancies annihilation. Based on the interaction model of <100> production, the ratio of <100>-type interstitial loops is higher at the higher dose rate irradiation, consistent with the experimental results.

目的 尝试采用计算模拟方法探究剂量率对辐照微结构特征的影响,探究常温辐照下剂量率效应的机理.方法 采用动力学蒙特卡罗(OKMC)方法,结合近些年第一性原理和分子动力学的计算参数,研究了常温下铁-碳体系中辐照缺陷随剂量率的变化特征.结果 在较低的剂量范围(<0.01 dpa)内,间隙型位错环的密度随剂量率的增加而增加;但在较高的剂量范围内,高剂量率辐照呈现较低的间隙型位错环密度、较大的缺陷尺寸.通过比较不同剂量率下的位错线对点缺陷的吸收数目,把剂量率在不同剂量范围内的特征归结于位错吸收与缺陷复合之间的竞争.结论 在较低的剂量范围内,位错吸收具有重要影响,随着剂量率的增加,位错吸收的缺陷数目显著减少;而在较高的剂量范围内,基体中间隙-空位缺陷的复合随剂量率的增加而显著增加,以至于高剂量率辐照可能产生较低的辐照硬化.文中的工作对理解剂量率效应的机理提供了一定的科学依据,为离子束模拟中子辐照提供了一定的科学参考.