Effect of simultaneous helium implantation on the microstructure evolution of Inconel X-750 superalloy during dual-beam irradiation

This study focuses on investigation into the effect of helium implantation on microstructure evolution in Inconel X-750 superalloy during dual-beam (Ni⁺/He⁺) irradiation. The 1 MeV Ni⁺ ions with the damage rate of 10^?3 dpa/s as well as 15 keV He⁺ ions using rate of 200 appm/dpa were simultaneously employed to irradiate specimens at 400 °C to different doses. Microstructure characterization has been conducted using high-resolution analytical transmission electron microscopy (TEM). The TEM results show that simultaneous helium injection has significant influence on irradiation-induced microstructural changes. The disordering of γ′ (Ni₃ (Al, Ti)) precipitates shows noticeable delay in dose level compared to mono heavy ion irradiation, which is attributed to the effect of helium on promoting the dynamic reordering process. In contrast to previous studies on single-beam ion irradiation, in which no cavities were reported even at high doses, very small (2–5 nm) cavities were detected after irradiation to 5 dpa, which proved that helium plays crucial role in cavity formation. TEM characterization also indicates that the helium implantation affects the development of dislocation loops during irradiation. Large 1/3 〈1?1?1〉 Frank loops in the size of 10–20 nm developed during irradiation at 400 °C, whereas similar big loops detected at higher irradiation temperature (500 °C) during sole ion irradiation. This implies that the effect of helium on trapping the vacancies can help to develop the interstitial Frank loops at lower irradiation temperatures.     

Evolution of dislocation loops in austenitic stainless steels implanted with high concentration of hydrogen

Abstract

It has been found that under certain conditions, hydrogen retention would be strongly enhanced in irradiated austenitic stainless steels. To investigate the effect of the retained hydrogen on the defect microstructure, AL-6XN stainless steel specimens were irradiated with low energy (100 keV) H₂⁺ so that high concentration of hydrogen was injected into the specimens while considerable displacement damage dose (up to 7 dpa) was also achieved. Irradiation induced dislocation loops and voids were characterised by transmission electron microscopy. For specimens irradiated to 7 dpa at 290 °C, dislocation loops with high number density were found and the void swelling was observed. At 380 °C, most of dislocation loops were unfaulted and tangled at 7 dpa, and the void swellings were observed at 5 dpa and above. Combining the data from low dose in previous work to high dose, four stages of dislocation loops evolution with hydrogen retention were suggested. Finally, molecular dynamics simulation was made to elucidate the division of large dislocation loops under irradiation.     

Study of defect evolution by TEM with in situ ion irradiation and coordinated modeling

The paper describes a novel transmission electron microscopy (TEM) experiment with in situ ion irradiation designed to improve and validate a computer model. TEM thin foils of molybdenum were irradiated in situ by 1?MeV Kr ions up to ～0.045 displacements per atom (dpa) at 80°C at three dose rates ?5?×?10^?6, 5?×?10^?5, and 5?×?10^?4?dpa/s – at the Argonne IVEM-Tandem Facility. The low-dose experiments produced visible defect structure in dislocation loops, allowing accurate, quantitative measurements of defect number density and size distribution. Weak beam dark-field plane-view images were used to obtain defect density and size distribution as functions of foil thickness, dose, and dose rate. Diffraction contrast electron tomography was performed to image defect clusters through the foil thickness and measure their depth distribution. A spatially dependent cluster dynamic model was developed explicitly to model the damage by 1?MeV Kr ion irradiation in an Mo thin foil with temporal and spatial dependence of defect distribution. The set of quantitative data of visible defects was used to improve and validate the computer model. It was shown that the thin foil thickness is an important variable in determining the defect distribution. This additional spatial dimension allowed direct comparison between the model and experiments of defect structures. The defect loss to the surfaces in an irradiated thin foil was modeled successfully. TEM with in situ ion irradiation of Mo thin foils was also explicitly designed to compare with neutron irradiation data of the identical material that will be used to validate the model developed for thin foils.     

In situ study of self-ion irradiation damage in W and W–5Re at 500 °C

In situ self-ion irradiations (150?keV?W⁺) have been carried out on W and W–5Re at 500?°C, with doses ranging from 10¹⁶ to 10¹⁸ W⁺m^?2 (～1.0?dpa). Early damage formation (10¹⁶W⁺m^?2) was observed in both materials. Black–white contrast experiments and image simulations using the TEMACI software suggested that vacancy loops were formed within individual cascades, and thus, the loop nucleation mechanism is likely to be ‘cascade collapse’. Dynamic observations showed the nucleation and growth of interstitial loops at higher doses, and that elastic loop interactions may involve changes in loop Burgers vector. Elastic interactions may also promote loop reactions such as absorption or coalescence or loop string formation. Loops in both W and W–5Re remained stable after annealing at 500?°C. One-dimensional hopping of loops (b?=?1/2 ?111>) was only seen in W. At the final dose (10¹⁸W⁺m^?2), a slightly denser damage microstructure was seen in W–5Re. Both materials had about 3–4?×?10¹⁵ loops m^?2. Detailed post-irradiation analyses were carried out for loops of size???4?nm. Both b?=?1/2 ?111? (～75%) and b?= ?100> (～25%) loops were present. Inside–outside contrast experiments were performed under safe orientations to determine the nature of loops. The interstitial-to-vacancy loop ratio turned out close to unity for 1/2 ?111? loops in W, and for both 1/2 ?111? and ?100? loops in W–5Re. However, interstitial loops were dominant for ?100? loops in W. Re seemed to restrict loop mobility, leading to a smaller average loop size and a higher number density in the W-Re alloy.     

Effect of helium concentration on irradiation damage of Fe-ion irradiated SIMP steel at 300℃ and 450℃

SIMP steel is newly developed fully martensitic steel for lead-cooled fast reactors and accelerator-driven systems.It is important to evaluate its radiation resistance via high flux neutron irradiation, where dense He atoms can be formed via(n, α) transmutation reaction. Co-irradiation with Fe and He ions, instead of neutron, was performed. Specimens were irradiated with 6.4-Me V Fe ions to the damage dose of 5 dpa at a depth of 600 nm. Three different helium injection ratios of 60-appm He/dpa(dpa: displacements per atom), 200-appm He/dpa and 600-appm He/dpa at a depth of 600 nm,were performed. Two different irradiation temperatures of 300℃ and 450℃ were carried out. The effect of helium concentration on the microstructure of Fe-irradiated SIMP steel was investigated. Microstructural damage was observed using transmission electron microscopy. The formed dislocation loops and bubbles depended on the helium injection ratio and irradiation temperature. Lots of dislocation loops and helium bubbles were homogeneously distributed at 300℃, but not at 450℃. The causes of observed effects are discussed.     

Migration behaviour of vacancies and damage structure recovery in a Fe-based Fe-Cr-Mn-Cu-Mo multi-component alloy

ABSTRACT

A Fe-based multi-component alloy, 60Fe-12Cr-10Mn-15Cu-3Mo, which presents higher yield stress than typical stainless steels (such as 304, 316, and 340), was used to investigate the thermal stability of irradiation-induced defects. Neutron irradiation was carried out at approximately 323 and 643?K using up to 1.3 × 10^?3 and 4.5 × 10^?4 dpa (displacements per atom), respectively. While no defects were accumulated at the high temperature of 643?K, single vacancies were formed after irradiation at the low temperature of 323?K to 1.3 × 10^?3 dpa, and the vacancies became mobile at 423?K. As a result, vacancy clusters were formed. However, as the annealing temperature increased the size of vacancy clusters decreased. Coincidence Doppler broadening measurements indicated that Cu precipitates were the sites of vacancy cluster formation, and the recovery of vacancy clusters became prominent while annealing the irradiated sample at temperatures higher than 423?K. Recovery of vacancy clusters at 573?K, which was not a high temperature, was also observed even in the sample that was irradiated using 2.5?MeV Fe ions at room temperature to 0.6 dpa at damage peak.     

Study of vacancy defects and their thermal stability in MeV Fe ion irradiated RAFM steel using positron beam Doppler broadening spectroscopy

Indian reduced activation ferritic-martensitic steel was irradiated with 1.1?MeV Fe ions to various doses from 1 to100?dpa at room temperature. The depth profiling of irradiation-induced vacancy-type defects and the defect-recovery under post-irradiation annealing was studied using variable low-energy positron beam Doppler broadening spectroscopy. The influence of irradiation-induced defects on the microstructural properties was studied by glancing incidence x-ray diffraction (GIXRD) and nanoindentation technique. Positron annihilation study showed the signatures of reduced vacancy concentration at the peak damage region due to injected interstitial effect from 30 to 100?dpa and the widening of vacancy-interstitial recombination-rich region towards the end of ion range with the increase in dose. The GIXRD results were analysed by modified Williamson–Hall plot method, and the variation of coherent domain size and micro-strain with irradiation dose was studied. Irradiation-induced hardening was observed in the nanoindentation study. The features observed in the GIXRD and nanoindentation study are correlated with the depth-resolved defect distribution observed in the positron annihilation study.     

超临界水冷堆燃料包壳材料的辐照损伤研究进展

超临界水冷堆（SCWR）是第四代核电站的主力堆型之一,高温、高压、超临界水环境下的辐照损伤问题是其燃料包壳材料面临的最大挑战。SCWR燃料包壳候选材料主要包括锆合金、奥氏体不锈钢、铁素体/马氏体不锈钢、镍基合金、ODS合金五大类,奥氏体不锈钢是最有希望的候选材料。介绍了近年来在这个领域国际上的主要研究进展。作者所在团队也对多种SCWR的候选材料进行了辐照损伤研究,包括:镍基合金C-276和718、铁素体/马氏体钢P92、奥氏体不锈钢AL-6XN和HR3C。对AL-6XN的氢离子辐照实验发现,辐照产生的缺陷主要是间隙型位错环,伯格斯矢量为1/3<111>,在较高剂量（5～7 dpa）辐照下,出现空洞肿胀。在氢滞留的影响下,位错环有着独特的演化规律,总结提出了位错环的四阶段演化过程。The Supercritical Water-cooled Reactor (SCWR) is one of the prior Generation IV advanced reactors. Irradiation damage is one of the key issues of fuel cladding materials which will suffer serious environment, such as high temperature, high pressure, high irradiation and supercritical water. The candidate materials contain zirconium alloys, austenitic stainless steels, ferritic/martensitic stainless steels, Ni-base alloys and ODS alloys. Austenitic stainless steels are the most promising materials. This paper summarized the international researches on irradiation effects in fuel cladding materials for SCWR. The group of authors also has done many researches in this field, including nickel-base alloy C-276 and 718, ferritic/martensitic steel P92 and austenitic stainless steel AL-6XN and HR3C. In AL-6XN austenitic stainless steels irradiated by hydrogen ions, dislocation loops were the dominant irradiation defects. At higher irradiation dose (5～7 dpa), the voids were found. All the dislocation loops were confirmed to be 1/3<111> interstitial type dislocation loops, and four evolution stages of dislocation loops with hydrogen retention were suggested.     

Bubble-loop complexes induced by helium irradiation and effect of pre-irradiation and post-irradiation of hydrogen in reduced-activation ferritic/martensitic steel

In the present work, the synergistic effect of high concentration hydrogen and helium on the dislocation loops and bubbles as well as their correlations in reduced-activation ferritic/martensitic (RAFM) steels is investigated. Such an effect was transmuted from 14?MeV neutron irradiation and has been one of the most challenging issues for RAFM steels for future fusion reactors. After low dose (0.18?dpa) high concentration (5000 appm) single-ion helium irradiation at 723?K, very large dislocation loops were observed, and the majority of bubbles were inside dislocation loops, forming bubble-loop complexes. These bubble-loop complexes defects were also present in hydrogen/helium and helium/hydrogen sequential-ion irradiated steels. Pre-irradiated hydrogen ion effectively inhibited the later growth of loops induced by helium post-irradiation, and the higher the ratio of hydrogen to helium fluence, the greater the effect of inhibition. At high fluence of hydrogen pre-irradiation, the structure of bubble-loop complexes disappeared. On the other hand, hydrogen post-irradiation promoted the growth of loops induced by helium pre-irradiation, and the higher the ratio of hydrogen to helium fluence, the greater the effect of promotion. The mechanisms for hydrogen/helium synergistic effects are discussed.     

6061-Al合金的自离子辐照损伤效应

铝合金是国内外研究堆的主要结构材料,在前期300#研究堆主要结构材料铝合金辐照性能研究的基础上,通过离子辐照研究6061-Al合金的微观结构损伤和引起的硬度变化,以开展较高辐照剂量下6061-Al合金损伤效应的前期探索。结果表明,经过自离子辐照后,6061-Al合金中产生了夹角为72°的位错环等缺陷,随着辐照剂量从0.218×1016 cm?2增加到4.367×1016 cm?2,缺陷密度明显增加,但选区电子衍射表明合金保持了很好的晶体结构,并没有发生非晶化。纳米压痕测试表明,不同辐照剂量下,样品中产生了不同程度的硬化,且微观硬度随着辐照剂量的增加而增加,当剂量增加到2.183×1016和4.367×1016 cm?2时,辐照硬化达到饱和,约为11%。研究结果可为初步预测较高中子辐照剂量下6061-Al合金结构和性能的变化提供数据支撑。     

Defect aggregates in neutron-irradiated β-Si3N4

Abstract

Microstructural analysis of the defect aggregates formed in bulk samples of polycrystalline β-Si₃N₄ neutron-irradiated to a dose of ～2.0 × 10²⁶n/m² at temperatures of 1100 K and 925 K has been carried out. This study has shown that the defect aggregates formed are faulted dislocation loops lying on the {1010} planes with a Burgers vector of b ? 1 /10<1125>. The vector is non-rational but corresponds to the insertion of an extra layer of [SiN4] tetrahedra on the {10l0} planes plus an additional shear in the loop plane. The formation of these loops is dependent upon the temperature of irradiation. In the sample irradiated at 1100 K their formation is additionally dependent upon whether or not a particular grain contains pre-existing c-axis dislocations. If no c-axis dislocations are present then independent nucleation of the loops is apparent; if there are pre-existing c-axis dislocations then the loops form from an apparent dissociation between the arcs of the irradiation-induced helical c-axis dislocation. In the sample irradiated at 925 K only independent nucleation of the loops occurs, regardless of whether or not there are any pre-existing c-axis dislocations in the grains.     

Effect of irradiation at low doses and incubation of voids within the rate theory approach

The evolution of defects in a material under irradiation is studied at low doses (∼5 dpa or less) using rate equations. It is shown that as a function of temperature at a critical valueT _c a transition occurs in the behaviour of the solutions of the rate equations. BelowT _c the voids show incubation effects. An expression is derived for the critical dislocation density at which the void growth starts. This is related to the trapped vacancy fraction ε in vacancy dislocation loops. AboveT _c the incubation effects are shown to be related to the gas production rate which becomes the rate controlling parameter in determining the evolution of the defects. A gas-bubble to void transition occurs at a critical void radius and expressions are derived for the critical void size and dose at which the transition appears. It is shown that closely related to this is the incubation dose for interstitial loops. Finally, these features are corroborated by actual numerical integration of the rate equations.     

Evaluation of Argon ion irradiation hardening of ferritic/martensitic steel-T91 using nanoindentation,X-ray diffraction and TEM techniques

In the present study, microstructures of Ferritic-martensitic T-91 steel irradiated at room temperature for 5, 10 and 20?dpa using 315?KeV Ar⁺⁹ ions have been characterized by grazing incident X-ray diffraction (GIXRD) and by transmission electron microscopy (TEM). Line profiles of GIXRD patterns have shown that the size of domain continuously reduced with increasing dose of radiation. TEM investigations of irradiated samples have shown the presence of black dots, the number density of which decreases with increasing dose. Microstructures of irradiated samples have also revealed the presence of point defect clusters, such as dislocation loops and bubbles. In addition, dissolution of precipitates due to irradiation was also observed. Nano-indentation studies on the irradiated samples have shown saturation behavior in hardness as a function of dose which could be correlated with the changes in the yield strength of the alloy.     

反应堆压力容器钢A508-3辐照硬化的剂量率效应研究

针对不同剂量率对国产反应堆压力容器钢(Reactor Pressure Vessel, RPV)A508-3辐照硬化的影响，利用3.5 MeV的Fe离子在3种不同剂量率(0.1, 0.5和 1.0 dpa/h)下将样品辐照至4个不同剂量点(0.1, 0.3, 1.0和3.0 dpa)，采用纳米压痕技术表征样品在不同辐照条件下的硬化效应。结果表明，在高剂量率(1.0 dpa/h)下，材料的硬度随剂量的增大快速增加，在0.3 dpa以后逐渐达到饱和；中剂量率(0.5 dpa/h)和低剂量率(0.1 dpa/h)下，样品硬度随剂量的变化趋势与高剂量率相似，但在0.3 dpa以后材料的硬度随剂量仍在缓慢增加。在辐照剂量低于0.3 dpa时，不同剂量率引起的辐照硬化差异较小，但在辐照剂量大于0.3 dpa时，不同剂量率辐照下的硬化效应差异显著。数据拟合结果表明，在实验剂量范围内辐照硬化的饱和值随剂量率的增加呈减函数关系。     

Superior tolerance of Ag/Ni multilayers against Kr ion irradiation: an in situ study

Monolithic Ag and Ni films and Ag/Ni multilayers with individual layer thickness of 5 and 50?nm were subjected to in situ Kr ion irradiation at room temperature to 1 displacement-per-atom (a fluence of 2?×?10¹⁴?ions/cm²). Monolithic Ag has high density of small loops (4?nm in diameter), whereas Ni has fewer but much greater loops (exceeding 20?nm). In comparison, dislocation loops, ～4?nm in diameter, were the major defects in the irradiated Ag/Ni 50?nm film, while the loops were barely observed in the Ag/Ni 5?nm film. At 0.2?dpa (0.4?×?10¹⁴?ions/cm), defect density in both monolithic Ag and Ni saturated at 1.6 and 0.2?×?10²³/m³, compared with 0.8?×?10²³/m³ in Ag/Ni 50?nm multilayer at a saturation fluence of ～1?dpa (2?×?10¹⁴?ions/cm²). Direct observations of frequent loop absorption by layer interfaces suggest that these interfaces are efficient defect sinks. Ag/Ni 5?nm multilayer showed a superior morphological stability against radiation compared to Ag/Ni 50?nm film.     

Effect of irradiation on microstructure and hardening of Cr–Fe–Ni–Mn high-entropy alloy and its strengthened version

ABSTRACT

A single-phase fcc high-entropy alloy (HEA) of 20%Cr–40%Fe–20%Mn–20%Ni composition and its strength with yttrium and zirconium oxides version was irradiated with 1.4?MeV Ar ions at room temperature and mid-range doses from 0.1 to 10 displacements per atom (dpa). Transmission electron microscopy (TEM), scanning transmission electron microscopy with energy dispersive X-ray spectrometry (STEM/EDS) and X-ray diffraction (XRD) were used to characterise the radiation defects and microstructural changes. Nanoindentation was used to measure the ion irradiation effect on hardening. In order to understand the irradiation effects in HEAs and to demonstrate their potential advantages, a comparison was performed with hardening behaviour of 316 austenitic stainless steel irradiated under an identical condition. It was shown that hardness increases with irradiation dose for all the materials studied, but this increase is lower in high-entropy alloys than in stainless steel.     

Heavy-ion irradiations of Fe and Fe–Cr model alloys Part 1: Damage evolution in thin-foils at lower doses

The evolution of radiation damage in Fe and Fe–Cr alloys under heavy-ion irradiation was investigated using transmission electron microscopy. Thin foils were irradiated with 100 or 150 keV Fe⁺ and Xe⁺ ions at room temperature (RT) and 300°C. Dynamic observations followed the evolution of damage and the early stages in damage development are reported. Small (2–5 nm) dislocation loops first appeared at doses between 10¹⁶ and 10¹⁷ ions m^?2 in all materials. Loop number densities depended strongly on the foil orientation in pure Fe but not in Fe–Cr alloys. Number densities did not depend strongly on Cr content. For a given material, defect yields were higher for Xe⁺ ions than for Fe⁺ ions, and were higher at RT than at 300°C. Loops with both ?100? and ½?111? Burgers vectors were identified. The proportion of ?100? loops was larger, especially in pure Fe. Dynamic observations showed that: the contrast of some new loops developed over intervals as long as 0.2 s; hopping of ½?111? loops was induced by the ion and electron beams and was pronounced in ultra-pure iron; and many loops were lost during and after ion irradiation by glide to the foil surface. The number of loops retained was strongly dependent on the foil orientation in Fe, but less so in Fe–Cr alloys. This is due to lower loop mobility in Fe–Cr alloys, probably due to pinning by Cr atoms. Reduced loop loss probably explains the higher loop number densities in Fe–Cr alloys compared with pure Fe.     

Crystal plasticity modeling of irradiation growth in Zircaloy-2

Abstract

A physically based reaction-diffusion model is implemented in the visco-plastic self-consistent (VPSC) crystal plasticity framework to simulate irradiation growth in hcp Zr and its alloys. The reaction-diffusion model accounts for the defects produced by the cascade of displaced atoms, their diffusion to lattice sinks and the contribution to crystallographic strain at the level of single crystals. The VPSC framework accounts for intergranular interactions and irradiation creep, and calculates the strain in the polycrystalline ensemble. A novel scheme is proposed to model the simultaneous evolution of both, number density and radius, of irradiation-induced dislocation loops directly from experimental data of dislocation density evolution during irradiation. This framework is used to predict the irradiation growth behaviour of cold-worked Zircaloy-2 and trends compared to available experimental data. The role of internal stresses in inducing irradiation creep is discussed. Effects of grain size, texture and external stress on the coupled irradiation growth and creep behaviour are also studied and compared with available experimental data.     

Recrystallization of hexagonal silicon carbide after gold ion irradiation and thermal annealing

Silicon carbide (SiC) single crystals with the 6H polytype structure were irradiated with 4.0-MeV Au ions at room temperature (RT) for increasing fluences ranging from 1?×?10¹² to 2?×?10¹⁵ cm^?2, corresponding to irradiation doses from ~0.03 to 5.3 displacements per atom (dpa). The damage build-up was studied by micro-Raman spectroscopy that shows a progressive amorphization by the decrease and broadening of 6H-SiC lattice phonon peaks and the related growth of bands assigned to Si–Si and C–C homonuclear bonds. A saturation of the lattice damage fraction deduced from Raman spectra is found for ~0.8?dpa (i.e. ion fluence of 3?×?10¹⁴ cm^?2). This process is accompanied by an increase and saturation of the out-of-plane expansion (also for ~0.8?dpa), deduced from the step height at the sample surface, as measured by phase-shift interferometry. Isochronal thermal annealing experiments were then performed on partially amorphous (from 30 to 90%) and fully amorphous samples for temperatures from 200 °C up to 1500 °C under vacuum. Damage recovery and densification take place at the same annealing stage with an onset temperature of ~200 °C. Almost complete 6H polytype regrowth is found for partially amorphous samples (for doses lower than 0.8 dpa) at 1000 °C, whereas a residual damage and swelling remain for larger doses. In the latter case, these unrelaxed internal stresses give rise to an exfoliation process for higher annealing temperatures.