Irradiation-enhanced Diffusion Research Articles

Irradiation response of microstructure and mechanical properties of NbMoVCr multi-principal element alloy coatings

In this study, NbMoVCr multi-principal element alloy (MPEA) coatings with near-equal atomic ratios were deposited onto the ferritic/martensitic (F/M) steel substrate by magnetron sputtering and then irradiated using 6 MeV Au2+ at room temperature (RT) and 550 °C, respectively. The results demonstrate that the NbMoVCr coatings have excellent irradiation resistance in terms of phase stability. RT-irradiation introduces both interstitial- and vacancy-type dislocation loops, and the dislocation loop density increases with the increase of irradiation damage dose. Meanwhile, RT-irradiation could also induce grain growth of the coating. However, under high-temperature irradiation, more recombination and annihilation of irradiation-induced defects, and recrystallization occur in the coating. In addition, irradiation also promotes the desegregation of V and Cr elements at the grain boundaries and enhances the uniformity of the distribution of coating elements. Irradiation softening and hardening in NbMoVCr coatings are also found to depend strongly on both irradiation damage dose and irradiation temperature. The creep deformation mechanism of the coating is dominated by diffusion; thus, the irradiation-enhanced diffusion reduces the creep resistance of the coating.

Effect of proton irradiation dose rate and implanted hydrogen ions on spinodal decomposition in thermally aged EQ308L stainless steel weld metal

Abstract Stainless steel welds with a ferritic phase are extensively utilized in nuclear power plants. Spinodal decomposition stands as the primary factor contributing to the degradation of their service performance. Ion irradiation has been widely employed to investigate the damage behavior of materials. However, varying irradiation parameters, such as dose rate and implanted ions, are believed to result in significant differences in the extent of spinodal decomposition. In this paper, proton irradiation experiments were conducted at different dose rates on thermally aged EQ308L stainless steel welds. Spinodal decomposition within the ferrite at various radiation depths was compared and analyzed by using atom probe tomography to unveil the effects of dose rate and implanted hydrogen ions. The results reveal that, under conditions of high dose rate irradiation, spinodal decomposition can be partially alleviated when compared to the initial thermally aged state. Conversely, under low dose rate irradiation conditions, spinodal decomposition exhibits an enhancement trend. This difference may be attributed to the interplay between irradiation-enhanced diffusion and atomic mixing. Therefore, the dose rate significantly influences the progression of spinodal decomposition. Furthermore, implanted hydrogen ions may also inhibit spinodal decomposition within the ferrite, potentially by promoting the recombination of irradiation defects.

Surface growth, inner defects and interface mixing of Cr coating on Zr alloy irradiated by 5 MeV protons at 400°C

Surface, inner and interface damages of Cr coatings irradiated by 5 MeV protons were characterized by XRD, SEM, TEM and nanoindentation in this work. After irradiation, a Cr7C3 growth layer with refined grains was formed on the surface due to the separation and redeposition of diffused Cr atoms. Irradiation-induced defects in the Cr coatings were voids, dislocations, and bubble-vacancies complexes. Importantly, irradiation softening was found for the arc-ion plated Cr coating, which should be due to the deformation recovery and recombination of inherent sinks and irradiation defects. At the coating-substrate interface, the ZrCr2 C14 phase, a multilayered interface and the preferential diffusion were found, meaning chained or layered Cr atoms may dominate the interface diffusion. In mechanism, formation mechanisms of irradiation defects are analyzed and the irradiation-enhanced diffusion, ballistic collision and the sharpening effect are employed to explain the interface mixing. A correlation coefficient of the anisotropic diffusion is proposed in the Cr flux equation for explaining the (110)-dominated diffusion. The results show that irradiation effects of coatings depend on the inherent defects and preferred orientation produced in the preparation process of coatings.

Enhanced irradiation tolerance of a medium entropy alloy via precipitation and dissolution of nanoprecipitates

High/medium entropy alloys (HEAs/MEAs) are good candidates for nuclear applications due to the excellent mechanical properties, good corrosion resistance and radiation resistance. In this work, a novel cobalt-free MEA was developed by introducing L12 nanoprecipitates. The microstructure evolution and radiation tolerance were evaluated after bombarded using 3 MeV Fe11+ ions at 500 °C. The evolution of nanoprecipitates was closely related to the irradiation dose, and dominated by irradiation-enhanced diffusion and ballistic dissolution mechanism. For the solid-solution MEA (without L12 nanoprecipitates), irradiation hardening occurred due to the irradiation-induced formation of precipitates, voids and dislocation loops. However, in MEA with L12 nanoprecipitates after aging, irradiation induced dissolution and reprecipitation of nanoprecipitates were observed. Different from the solid-solution MEA, the hardness kept almost unchanged in the aging sample after irradiation. The swelling rate of the solid-solution sample after irradiation is 9.4 × 10−6 %/dpa, while no swelling occurs in the aging sample under the same irradiation condition. Besides, the average size and number density of dislocation loops in the aging sample decreased by ∼ 40% and ∼ 28%, compared with the solid-solution sample. The precipitation and dissolution of nanoprecipitates substantially improved the radiation tolerance of the cobalt-free MEA.

Enhanced helium ion irradiation tolerance in a Fe-Co-Ni-Cr-Al-Ti high-entropy alloy with L12 nanoparticles

• Helium bubbles and large stacking faulted loops are observed as the dominant structural damage in the He ion irradiated HEA reinforced by L1 2 nanoparticles. • The L1 2 nanoparticles provide numerous interfaces for He entrapment and damage elimination, which suppresses He bubble growth. • A correlative TEM/APT characterization reveals that the RIS around He bubbles is dominated by the inverse Kirkendall mechanism. • Irradiation-induced dissolution and re-precipitation of the L1 2 nanoparticles can retain the main microstructure of the L1 2 -strengthened HEA and provide a sustainable irradiation resistance. L1 2 -strengthened high entropy alloys (HEAs) with excellent room and high-temperature mechanical properties have been proposed as promising candidates as structural materials for advanced nuclear systems. However, knowledge about their radiation response is fairly limited. In the present work, a novel HEA with a high density of L1 2 nanoparticles was irradiated with He ion at 500°C. Transmission electron microscope (TEM) and atom probe tomography (APT) were employed to study the evolution of microstructural stability and radiation-induced segregation. Similar to the single-phase FeCoNiCr HEA, the main microstructural features were numerous large faulted dislocation loops and helium bubbles. While the irradiation resistance of the present L1 2 -strengthened HEA is much improved in terms of reduced bubble size, which could be attributed to the considerable He trapping efficiency of the coherent precipitate/matrix interface and the enhanced capability of the interface for damage elimination when the matrix channel width is narrow. APT analysis revealed that an inverse Kirkendall mechanism-dominated radiation-induced segregation (RIS) occurs around bubbles, where a significant Co enrichment and Ni depletion can be clearly observed. In addition, the competing dynamics of ballistic mixing and elemental clustering that raised from the irradiation-enhanced diffusion in a highly supersaturated matrix, along with the low precipitation nucleation barrier due to the small lattice misfit, lead to a dynamical precipitation dissolution and re-precipitation appears under irradiation. Such a promising phenomenon is expected to promote a potential self-healing effect and could in turn provide a sustainable irradiation tolerance over the operational lifetime of a reactor.

Simulation of neutron irradiation-induced recrystallization of tungsten

<sec>Tungsten is the candidate for divertor target material in future fusion reactors. The tungsten divertor target is expected to long serve in a harsh environment of high temperature and high-energy neutron irradiation. This can lead to neutron irradiation-induced recrystallization of tungsten, thereby increasing the possibility of intergranular brittle failure and compromising the safe operation of the divertor. Thus, clarifying the mechanism of neutron irradiation-induced tungsten recrystallization is important. However, the current model, which only considers the irradiation-enhanced effect on recrystallization driving force, underestimates the irradiation effect on recrystallization compared with the results observed in recent high-temperature neutron irradiation experiments in the HFIR reactor. It indicates that other irradiation effects can also influence the recrystallization process.</sec><sec>In this study, we introduce the irradiation-enhanced grain boundary migration factor (<i>R</i>) into the established irradiation-induced recrystallization kinetic model, on the assumption that the grain boundary migration velocity is proportional to the self-diffusion coefficient. The simulation results show that after considering both irradiation-enhanced recrystallization driving force and grain boundary migration effect, the calculated half-recrystallization time (<inline-formula><tex-math id="M4">\begin{document}${t}_{{X}\text{}=\text{}0.5}$\end{document}</tex-math><alternatives><graphic xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="16-20230531_M4.jpg"/><graphic xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="16-20230531_M4.png"/></alternatives></inline-formula>) at 850 ℃ from the model matches the one obtained in the neutron irradiation experiment in the HFIR reactor. This result indicates that the irradiation-enhanced grain boundary migration effect is one of the important factors affecting irradiation-induced recrystallization. In addition, the difference between irradiated and unirradiated <i>t</i><sub><i>X</i>=0.5</sub> decreases with temperature increasing. This phenomenon is due to the fact that as the temperature increases, the contribution of irradiation defects to the driving force for recrystallization decreases owing to the irradiation defect recombination. Moreover, the increase of thermal activation diffusion coefficient is more significant than the increase of the irradiation-enhanced diffusion coefficient. These findings suggest that the thermal activation effect eventually dominates the recrystallization process over the irradiation effect as temperature increases.</sec>

A phase-field model with irradiation-enhanced diffusion for constituent redistribution in U-10wt%Zr metallic fuels

Constituent redistribution is a unique phenomenon to metal fuels that threatens the safety of such fuel forms. Therefore, it is imperative to establish models to understand the intrinsic mechanisms and predict the redistribution kinetics. In this work, we derived the conservative field equations of the phase-field model from near-equilibrium thermodynamic theory. A macroscopic constituent redistribution phase-field model was developed by introducing the effect of irradiation on the atom mobility and the effect of temperature on the interface mobility. An expression of phase boundary width, applicable to both microscopic and macroscopic scenarios was proposed. The interfacial parameters of the model and the Zr concentration distribution near the fuel surface were discussed at last. These works may help understand constituent redistribution characteristics and promote the application of the phase-field method in studying constituent redistribution in macroscopic scenarios.

Radiation-induced amorphization of M23C6 in F82H steel: An atomic-scale observation

The purpose of the present study is to clarify the instability behavior of M23C6 under irradiation, specifically the occurrence of radiation-induced amorphization (RIA). Ion irradiation of 10.5 MeV-Fe3+ at elevated temperatures from 573 to 623 K was conducted into the reduced activation ferritic/martensitic steels (F82H) and its model alloy (Fe-8Cr-0.1C). A bilayer contrast of the particle consisting of an amorphous-rim phase and inner crystalline core of M23C6 was observed in the irradiated F82H specimen, but not in the model alloy. From the high-resolution electron microscope observation, the preferential occupation site of W into M23C6 lattice was identified as 8c-site prior to irradiation in F82H specimen, which shifted to other sites due to chemical disordering upon irradiation. Evaluation of the intensity ratio between 8c and another site of M23C6, 8c/4a, then revealed that the extent of chemical disordering of W was mitigated at the amorphous-crystal interface region in comparison with the central of the particle. The hypothesis for the formation mechanism of an amorphous-rim in M23C6 was presumed as the deviation from the stoichiometric composition at the local interface due to the irradiation-enhanced diffusion and/or ballistic mixing under the current circumstances, although the efforts from experimental and/or simulation studies are still necessary to achieve a further understanding of the RIA behavior in M23C6.

Calculation of steady-state dynamical phase diagram in U-Mo binary system under irradiation

The steady-state dynamical phase diagram of U-Mo system under irradiation is studied by the thermodynamic model and effective free energy model. In the effective free energy model, we consider the irradiation-enhanced diffusion and ballistic mixing caused by recoil implantation and cascade collision. The effective free energy and diffusivity of each phase in U-Mo binary system at different temperatures were calculated. The dynamical phase diagrams of U-Mo alloys under different irradiation intensities were calculated and the differences between the dynamical phase diagram and thermodynamic equilibrium phase diagram was discussed. The calculated results show that, due to the irradiation-induced ballistic mixing, the high-temperature stable γ(U,Mo) phase can be stabilized near room temperature, which agree well with the experimental reports.

Cluster dynamics simulation of xenon diffusion during irradiation in UO2

Diffusion of fission gas in UO2 nuclear fuel impacts several important performance metrics, such as fission gas release, swelling, and thermal conductivity. Current empirical models of fission gas release have significant uncertainty, some of which derives from the bulk diffusion rate and its dependence on, for example, fuel chemistry and irradiation. We have applied the previously-developed Free Energy Cluster Dynamics (FECD) methodology in the code Centipede to calculate xenon cluster concentrations in UO2 under intrinsic (high temperature) and irradiation-enhanced (intermediate temperature) conditions in order to develop a model of the xenon diffusion coefficient based on the atomic scale mechanisms responsible for transport. While the diffusion mechanism for xenon in UO2 is adequately described by the Xe + U2O vacancy cluster for intrinsic conditions, a similar process is not capable of capturing measured in-pile fission gas diffusivity at intermediate temperatures. Therefore, a different diffusion mechanism must dominate under this regime. Using calculated atomistic data, we have shown that irradiation-enhanced diffusion at intermediate temperatures occurs via the larger Xe + U4Oy vacancy clusters, which have lower migration barriers and increase in concentration by several orders of magnitude compared to intrinsic conditions. This mechanism is enabled by the increased uranium vacancy concentration under irradiation due to Frenkel pair production. In addition, the fast migration of uranium interstitials with two attached oxygen interstitials lowers the total uranium interstitial concentration through reactions with sinks. This allows the extended defects, such as Xe + U4Oy vacancy clusters, to maintain high concentrations by limiting annihilation with attached vacancies. Predictions using the Xe + U4Oy diffusion mechanism are in good agreement with experiment, albeit with some differences in the Arrhenius slope, which we believe may be related to either experimental or model parameter uncertainty. Finally, an analytical expression suitable for application in fuel performance simulations was derived to capture the predictions of the Centipede simulations.

Steady-State Dynamic Phase Diagram Calculation of U-Ti and U-V Binary System under Irradiation

In this paper, by considering the irradiation-enhanced diffusion, a combination of effective free energy model and the thermodynamic model was used for studying the phase relationships in the U-Ti and U-V binary system under irradiation. The steady-state dynamical phase diagrams of U-Ti and U-V binary alloys under different irradiation intensities were calculated and compared with the conventional thermodynamic equilibrium phase diagram. The calculated results show that under irradiation the high-temperature stable (βTi, γU) and (γU,V) phases were stabilized at relatively low temperature resulting in invariant reactions at relatively low temperature. In addition, with the increase of the irradiation intensity, the temperature of the invariant reactions increased, and the phase regions of the (βTi, γU) and (γU,V) also increased.

Effect of friction stir welding and self-ion irradiation on dispersoid evolution in oxide dispersion strengthened steel MA956 up to 25 dpa

The effect of ion irradiation on the microstructure of oxide dispersion strengthened (ODS) MA956 steel, before and after friction stir welding (FSW), was studied. Both the base material (BM) and welded stir zone (SZ) were irradiated with 5 MeV Fe++ ions at 450 °C up to 25 displacements per atom (dpa). Characterization was performed using scanning transmission electron microscopy (STEM) and atom probe tomography (APT), with particular emphasis on the YAlO dispersoid characteristics and dislocation microstructures. After irradiation, the dispersoids in the BM increased in diameter and decreased in number density, which was explained by an Ostwald ripening mechanism. FSW caused significant coarsening and agglomeration of the dispersoids. After irradiation, both the diameter and number density of the SZ dispersoids increased, which was explained by an irradiation-enhanced diffusion mechanism. Dislocation loop and network behavior was also characterized and large dislocation loops of ≈20 nm diameter formed by 1 dpa in both the BM and SZ samples, whereas the network density remained nearly constant with irradiation.

Investigation of instability of M23C6 particles in F82H steel under electron and ion irradiation conditions

In order to clarify the instability of M23C6 in F82H steel under irradiation, both electron irradiation using a high voltage electron microscope (HVEM) and ion irradiation using an ion accelerator were performed. For the electron irradiation, in-situ observation under 2 MV electron irradiation and ex-situ high resolution electron microscopic (HREM) analysis were utilized to evaluate the response of M23C6 against irradiation. The temperature dependence of the irradiation induced instability of the carbide was first confirmed: 293 K < T < 573 K, by observation of lowering in contrast at the periphery of carbides, 698 K < T < 773 K, fragmentation at the interface between carbides and matrix, and at 773 K, formation and coarsening of new particles near the periphery of M23C6. HREM analysis showed the loss of the lattice fringe contrast at the pre-existing M23C6 precipitates at temperatures ranging from 473 to 773 K, indicating severe loss of crystallinity due to dissolution of the constituent atoms though irradiation-enhanced diffusion under the vacancy diffusion by the focused electron beam irradiation. For the ion irradiation, 10.5 MeV-Fe3+ ion was applied to bombard the F82H steel at 673 K to achieve the displacement damage of ≈20 dpa at the depth of 1.0 μm from surface. Cross-section TEM specimens were prepared by a focused ion beam technique. The shrinkage of carbide particles was observed especially near the irradiation surface. Besides, the lattice fringes at the periphery of carbide were observed in the irradiated M23C6 by the HREM analysis, which is different from that observed in the electron irradiation. It was clarified that the instability of M23C6 is dependent on the irradiation conditions, indicating that the flow rate of vacancy type defects might be the key factor to cause the dissolution of constituent atoms of carbide particles into matrix under irradiation.

Morphology variation, composition alteration and microstructure changes in ion-irradiated 1060 aluminum alloy

Morphology variation, composition alteration and microstructure changes in 1060 aluminum irradiated with 50 keV helium ions were characterized by field emission scanning electron microscopy (FESEM) equipped with x-ray elemental scanning, 3D measuring laser microscope and transmission electron microscope (TEM). The results show that, helium ions irradiation induced surface damage and Si-rich aggregates in the surfaces of irradiated samples. Increasing the dose of irradiation, more damages and Si-rich aggregates would be produced. Besides, defects such as dislocations, dislocation loops and dislocation walls were the primary defects in the ion implanted layer. The forming of surface damages were related with preferentially sputtering of Al component. While irradiation-enhanced diffusion and irradiation-induced segregation resulted in the aggregation of impurity atoms. And the aggregation ability of impurity atoms were discussed based on the atomic radius, displacement energy, lattice binding energy and surface binding energy.

A preliminary investigation of high dose ion irradiation response of a lanthana-bearing nanostructured ferritic steel processed via spark plasma sintering

A nanostructured ferritic steel with nominal composition of Fe-14Cr-1Ti-0.3Mo-0.5La2O3 (wt.%) was irradiated with Fe+2 ions at 475 °C for 100, 200, 300 and 400 dpa. Grain coarsening was observed for the samples irradiated for 200–400 dpa resulting in an increase of the average grain size from 152 nm to 620 nm. Growth of submicron grains at higher radiation doses is due to decreased pinning effect imparted by Cr-O rich nanoparticles (NPs) that underwent coarsening via Ostwald ripening. Dislocation density consistently increased with increasing irradiation dose at 300 and 400 dpa. The mean radius of lanthanum-containing nanoclusters (NCs) decreased and their number density increased above 200 dpa, which is likely due to solutes ejection caused by ballistic dissolution and irradiation-enhanced diffusion. Chromium, titanium, oxygen and lanthanum content of nanoclusters irradiated at 200 dpa and higher got reduced by almost half the initial value. The reduction in size of the nanoclusters accompanied with their higher number density and higher dislocation density led to significant radiation hardening with increasing irradiation dose.

Steady-state dynamical phase diagram calculation of U–Nb binary system under irradiation: Ballistic effect

A kinetic model involved in ballistic effects has been used for studying the steady-state configuration under irradiation, with considering the irradiation-enhanced diffusion. By applying this model to all the solid phases, the dynamical phase diagram of U–Nb system under irradiation is calculated and critically compared with the assessed thermodynamic equilibrium phase diagram. The calculated results indicate that the irradiation has little effect on the phase relationships of U–Nb system at high temperatures. However, it leads to the close of the miscibility gap of U–Nb system at relatively low temperatures. Meanwhile, the high-temperature stable γ(U,Nb) phase is stabilized by the irradiation, which is in agreement with the available experimental results.

Estimation of the solubility limit of Cr in Fe at 300 °C from small-angle neutron scattering in neutron-irradiated Fe–Cr alloys

The solubility limit of Cr in Fe (α-Fe–Cr) at low temperatures is a matter of debate. We report a direct estimation of the solubility limit at 300 °C from small-angle neutron scattering (SANS) data obtained for neutron-irradiated Fe–Cr alloys. The SANS results indicate that the equilibrium concentration of α′ was reached via irradiation-enhanced diffusion. The solubility limit was estimated using an iterative approach based on the SANS invariant and the lever rule of phase equilibrium.

XPS study of ion‐beam‐assisted formation of Si nanostructures in thin SiO2 layers

AbstractThin thermally grown SiO2 layers have been enriched with Si up to ∼15 at.% by Si ion implantation at room temperature. X‐ray photoelectron spectroscopy (XPS) was used to monitor the composition changes caused in the layers by bombardment with 3 keV Ar ions at elevated temperatures. Argon ion irradiation of pure SiO2 does not lead to the formation of Si nanoprecipitates. That was the case also for room‐temperature Ar bombardment of Si‐rich layers. Their ‘hot’ irradiation with Ar ions was found to enhance considerably the formation of segregated Si nanophase inclusions. The process starts at ∼500°C and becomes strongly pronounced at 650°C. At this temperature ion‐beam‐induced crystallization of Si nanoprecipitates may be achieved. The results obtained are explained in the framework of the physical model, based on the irradiation‐enhanced diffusion of excess Si atoms. Copyright © 2002 John Wiley &amp; Sons, Ltd.

Effects of AP-CITROX Decontamination Procedure on the Surface Oxide Layer Composition of Stainless Steel Originating from the Primary Circuit of a VVER-Type Nuclear Reactor

A real specimen originating from the primary circuit of a VVER-440 type pressurized water cooled nuclear reactor has been studied by Conversion Electron Mossbauer Spectroscopy (CEMS) in order to find out how the AP-CITROX decontamination procedure modifies the structure and composition of the surface oxide layer of stainless steel which is used in the steam generator. Other methods like voltammetry, gravimetry, and SEM-EDAX were also applied to characterize the samples and to help the interpretation of CEMS results. It was found that, in contradiction with expectations, the presence of the surface magnetite layer could not be convincingly identified even on the non-decontaminated sample. This finding together with the relatively weak Mossbauer signals indicated that the surface oxide layer is strongly Fe-depleted. It was also concluded that the upper layer of the bulk steel (under the oxide layer) has an altered composition probably due to irradiation-enhanced diffusion of the metallic constituents. It was established that the AP-CITROX decontamination procedure does not exert detrimental effects on the thickness and composition of the surface oxide layer.

Nanoscale analysis of electron irradiation-enhanced diffusion process on the multilayer interfaces of W-Al2O3-Ti/Cu

Nanoscale analysis of electron irradiation-enhanced diffusion process on the multilayer interfaces of W-Al2O3-Ti/Cu