Radiation-induced Segregation Research Articles

Radiation-induced Fe segregation in the dual-phase FeCrNiMnAl high-entropy alloy under high-dose helium ion irradiation

This study investigates the radiation-induced Fe segregation in the dual-phase FeCrNiMnAl high-entropy alloy (HEA) under the helium (He) ion irradiation with a dose of 5 × 1017 ions/cm2. The superior bubble swelling resistance of the chemical-ordered NiAlMn phase with B2 structure is observed compared to the FCC FeCrMn matrix, which can be attributed to their differences in atomic packing factors. Fe segregation at phase boundaries (PBs) is detected after irradiation, as evidenced by a higher concentration of Fe at PBs than in the matrix by about 10 at.%. This phenomenon can be ascribed to the migration of Fe interstitials induced by irradiation cascades and the punching mechanism during bubble growth. In essence, the Fe segregation in this study is driven by the radiation-enhanced diffusion and metastable state of FCC-structured FeCrMn with supersaturated Fe atoms. These insights provide theoretical foundations and innovative perspectives for the radiation-induced segregation (RIS) in HEAs.

Demonstration of an integrated-heating load frame for quantitatively assessing microscale tensile properties of copper and Zircaloy-2

Small-scale mechanical testing (SSMT) is of great interest in the nuclear materials community as it permits the use of surrogate irradiation techniques, such as ion beam irradiation when investigating the impact of irradiation damage on mechanical properties. The design of a micro-electro-mechanical-system (MEMS) micro-tensile stage is demonstrated to enable micron-sized specimen testing up to failure under ambient and reactor-relevant temperatures. Copper and Zircaloy-2 microscale specimens, having gauge lengths of 200 μm, gauge widths of 48 μm, and specimen-dependent constant gauge thicknesses of 10 μm to 26 μm, are fabricated using micro-wire electrical discharge machining and focused ion beam milling. From each gauge section, microstructure data is captured prior to testing using electron backscatter diffraction analysis, and an optimized digital image correlation speckle pattern is deposited on the specimen surface. The speckle pattern is tracked during deformation and post-processed to obtain strain-field evolution maps throughout deformation. Strain maps provide a means to account for and explain microstructure-dependent features observed in the captured stress-strain curves. The combination of using micro-wire electrical discharge machining, focused ion beam cleaning, and ion beam induced platinum deposition for micron-sized specimen preparation as well as utilizing microfabrication techniques to fabricate an integrated heating microtensile load frame reported herein serve as a demonstration of SSMT approaches that can be adopted to tackle challenging problems in nuclear materials research, including but not limited to the effects of ion beam irradiation on mechanical performance and providing experimental data to support small-scale modeling efforts of embrittling precipitates or radiation-induced segregation.

TEM Characterization of Radiation-Induced Segregation at Irradiation-Induced Dislocation loops in Al0.3CoCrFeNi and CoCrFeMnNi High Entropy Alloys

TEM Characterization of Radiation-Induced Segregation at Irradiation-Induced Dislocation loops in Al0.3CoCrFeNi and CoCrFeMnNi High Entropy Alloys

APPLICATION OF RUTHERFORD BACKSCATTERING METHOD FOR STUDYING RADIATION-INDUCED SEGREGATION OF HIGH-ENTROPY NICKEL-BASED ALLOYS

In this study, radiation-induced segregation was studied in high-entropy alloys (HEA) CoCrFeNi, CoCrFeMnNi, irradiated with helium ions He2+ with an energy of 40 keV at room temperature. Changes in the concentrations of HEAs and their depth distributions were studied by Rutherford Backscattering (RBS) and energy dispersive x-ray spectroscopy (EDS) methods. Measurements using the RBS and EDS methods showed that non-irradiated HEAs have a composition close to equiatomic, where the average concentration for CoCrFeNi is 24.8 atomic percents (at.%), and for CoCrFeMnNi – 20 at.%. The EDS results were significantly different from the RBS in Ni/Co concentrations, and indicated no significant changes in element distribution in both HEAs after irradiation. According to the RBS data, the largest changes in concentrations during irradiation in both HEAs relate to the enrichment of Ni atoms. In CoCrFeNi, upon irradiation, Ni/Co atoms undergo the greatest segregation, and in CoCrFeMnNi, the Ni/Co/Fe concentrations change significantly. In CoCrFeMnNi, the change in element concentrations with increasing irradiation fluence was more pronounced than in CoCrFeNi. In CoCrFeMnNi, changes in concentrations of all elements at both fluences reached 0.5–17% (0.1–3.1 at.%) and exceeded changes in CoCrFeNi, which reached 2–11% (0.5–1.9 at.%). It was found that the resistance to segregation when irradiated with helium ions under these conditions was lower for CoCrFeMnNi than for CoCrFeNi. In CoCrFeNi and CoCrFeMnNi, changes in the concentrations of Co, Fe, Cr, and Mn were significantly less than changes near sinks and defect clusters when irradiated with nickel ions with similar doses in other studies at temperatures close to the halfmelting temperature of nickel HEAs. The RBS study showed a uniform distribution of atoms in depth and resistance to segregation in CoCrFeNi, CoCrFeMnNi when irradiated with helium ions.

Enhanced radiation hardness of lead halide perovskite absorber materials via incorporation of Dy2+ cations

Continuously growing evidence of the excellent radiation hardness of perovskite solar cells stimulates their intense exploration in the context of potential aerospace applications. However, there is still a very limited understanding of the fundamental aspects such as the mechanisms of the interaction of complex lead halides with different types of ionizing radiation and associated aging pathways, even though this knowledge is essentially important for the development of new materials with improved properties. Herein, we made one of the first steps to fill this gap through a systematic study of the aging behavior of two perovskite absorber materials under exposure to three different stressors: 60Co gamma-rays, 8.5 MeV electron fluences, and light. The application of a set of complementary spectroscopy and microscopy techniques allowed us to identify and compare perovskite degradation pathways caused by each type of the ionizing radiation. In particular, radiation-induced phase segregation with the formation of δ-phases of CsPbI3 and FAPbI3 appears to be a common route in the case of double cation perovskites. Most importantly, we have revealed that the incorporation of dysprosium cations as a replacement of 1 % of Pb2+ ions in the Cs0.12FA0.88Pb0.99Dy0.01I3 material formulation results in a remarkably improved radiation hardness: this modification blocks the formation of metallic lead and strongly suppresses segregation of Cs-rich and FA-rich phases in perovskite films under exposure to gamma rays or high-energy electrons. Thus, these results open up a promising research direction for designing radiation-resistant perovskite absorbers through rational compositional engineering.

High-temperature dimensional stability and radiation behavior of high-entropy rare earth tungstate ceramics

Materials with high-temperature dimensional stability and resistance to radiation damage are receiving increasing attention in advanced nuclear energy systems. In this research, (Sm0.2Eu0.2Gd0.2Dy0.2A0.2)2W3O12-type (A = Y, Tm,Ho) high-entropy ceramics is presented. The coefficients of thermal expansion of (Sm0.2Eu0.2Gd0.2Dy0.2Y0.2)2W3O12 (HE-RE2W3O12) is 7.21 × 10−6 K−1 at 300–1100K, which is the lowest among the three synthesized high-entropy ceramics and indicates good dimensional stability at high temperatures. After irradiation with Kr+ fluence of 4.58 × 1016 ions/cm2 at 400 °C, both HE-RE2W3O12 and Gd2W3O12 maintained crystalline, but the lattice volume expansion rate of HE-RE2W3O12 (0.71 %) is much lower than that of Gd2W3O12 (1.49 %). Additionally, in the post-irradiated HE-RE2W3O12, the Raman spectra vibration peak remained unchanged and no radiation-induced segregation was observed. Nanoindentation tests displayed that the degradation of hardness and elastic modulus in HE-RE2W3O12 is rarely degrade. Overall, these results provide new insights for research on advanced nuclear materials in the future.

Influence of magnetic properties on elemental vacancy migration energy in Fe49.5Mn29.4Co10.1Cr10.1C0.9 high-entropy alloy

Randomly mixing ferromagnetic (FM) and antiferromagnetic (AFM) elements in high-entropy alloys (HEAs) can create fluctuating local magnetic moments that influence the energetics of point defects. In this study, we employed first-principles calculations to investigate the influence of magnetic properties on vacancy migration energy in Fe49.5Mn29.4Co10.1Cr10.1C0.9, alongside equiatomic NiCoFeCrMn alloy. By examining structures with paramagnetism, ferromagnetism, and no spin polarization, our study reveals significant impacts of magnetic interactions on vacancy migration barriers, potentially altering the sequence of elemental migration energies if overlooked. In Fe49.5Mn29.4Co10.1Cr10.1C0.9, the order of vacancy migration barriers is Co > Fe > Mn > Cr across all magnetic states, suggesting the dominant roles of atomic properties and inherent chemical bonding. Conversely, the NiCoFeCrMn HEA exhibits a pronounced magnetic state-dependent elemental migration energy order, indicating that magnetic interactions significantly influence vacancy migration behavior in this alloy. In addition, while FM elements generally exhibit higher migration barriers, AFM elements display lower barriers in the investigated Cantor alloys, with notable variations between the studied compositions. These findings underscore the critical role of magnetism in accurate migration energy calculations, which is important for studying chemically biased diffusion and radiation-induced segregation in HEAs.

Extending damage accumulation of commercial reactor irradiated 316 stainless steel with ion irradiation

Austenitic 316 stainless steel flux thimble tubes (FTTs) removed from a commercial pressurized water reactor (PWR) were re-irradiated using nickel ions to evolve the irradiated microstructure to higher damage levels than those achieved in reactor. The microstructures of the neutron-plus-ion irradiated samples were compared with those of neutron irradiated only samples at the same doses to evaluate the effectiveness of ion irradiation to extend the damage range and match that created in reactor. Ion irradiations effectively captured, both qualitatively and semi-quantitatively, the evolution of dislocation loops, nanocavities, nanocluster size and density, and radiation-induced segregation (RIS) with dose. Only Ni-Si clusters were observed in ion irradiated FTT samples while Ni-Si-Mn clusters that were frequently observed in reactor irradiated samples above ∼41 dpa were absent in ion irradiated samples probably due to ballistic mixing at the high ion irradiation damage rate. Two samples were ion irradiated to ∼160 dpa to predict the irradiated microstructure that may develop due to an increase of the damage accumulation by extending a PWR life from 40 years to 60 years. The extended ion irradiation results indicated that significant microstructure changes were unlikely to occur even to very high doses near common PWR relevant temperature conditions. Overall, the microstructure resulting from ion irradiation following neutron irradiation matched that of neutron to the same dose reasonably well in most cases.

Coupling of radiation and grain boundary corrosion in SiC

Radiation and corrosion can be coupled to each other in non-trivial ways and such coupling is of critical importance for the performance of materials in extreme environments. However, it has been rarely studied in ceramics and therefore it is not well understood to what extent these two phenomena are coupled and by what mechanisms. Here, we discover that radiation-induced chemical changes at grain boundaries of ceramics can have a significant (and positive) impact on the corrosion resistance of these materials. Specifically, we demonstrate using a combination of experimental and simulation studies that segregation of C to grain boundaries of silicon carbide leads to improved corrosion resistance. Our results imply that tunning of stoichiometry at grain boundaries either through the sample preparation process or via radiation-induced segregation can provide an effective method for suppressing surface corrosion.

Post-irradiation examination of commercial tantalum alloys following neutron irradiation

Post irradiation examination of two commercial tantalum alloys (based on Ta-8 %W) revealed significant irradiation hardening following neutron irradiation in the High Flux Isotope Reactor at Oak Ridge National Laboratory. Tensile samples of commercial Ta alloys T-111 and ASTAR-811C were irradiated to a total neutron fluence (E > 0.1 MeV) in the range (2–4)x1021 n/cm2 (0.39–0.75 dpa) at an irradiation temperature near 900 °C. Tensile testing conducted at room temperature and at 800 °C revealed significant irradiation hardening under all test conditions. Samples that were tensile tested at room temperature exhibited brittle failure, whereas ductility was maintained for elevated temperature testing. The irradiation hardening and embrittlement noted for both alloys was attributed to irradiation-induced dislocation loop formation as confirmed through scanning transmission electron microscopy analysis. Corresponding energy dispersive x-ray spectroscopy revealed radiation-induced segregation of Hf in these loop populations. This Hf enrichment is shown to result in eventual co-precipitation of elongated (Hf,O)-rich precipitates along the dislocation loops. These results show that irradiation hardening is enhanced via irradiation-enhanced precipitation at elevated irradiation temperatures. Consequently, the minimum recommended operating temperature window for Ta alloys should be increased to account for observed hardening/embrittlement at irradiation temperatures as high as 900 °C (1,173 K).

Dislocation channel broadening–A new mechanism to improve irradiation-assisted stress corrosion cracking resistance of additively manufactured 316 L stainless steel

Additively manufactured (AM) 316 L stainless steel (SS) after hot isostatic pressing (HIP) was found to exhibit superior resistance to irradiation-assisted stress corrosion cracking (IASCC) in high-temperature water, as compared to wrought 316 L SS. The well-accepted IASCC factors of radiation-induced segregation (RIS) and radiation hardening are not accurate descriptions of IASCC susceptibility in this case. A decreased strain localization along grain boundaries, caused by dislocation channel broadening, was confirmed to suppress crack initiation. A unique distribution of irradiation-induced defects in HIP AM 316 L SS eased dislocation cross-slip compared to those in the wrought counterpart, thus increasing the channel width near the grain boundaries. For the first time, this study highlights the importance of dislocation channel broadening as a potential mechanism to further improve the IASCC resistance of 316 L SS and provides direct experimental evidence based on commercial-grade materials.

Effect of Production Bias on Radiation-Induced Segregation in Ni-Cr Alloys

We present an in-depth investigation into the Radiation-Induced Segregation (RIS) phenomenon in Ni-Cr alloys. All the pivotal factors affecting RIS such as surface’s absorption efficiency, grain size, production bias, dose rate, temperature, and sink density were systematically studied. Through comprehensive simulations, the individual and collective impacts of these factors were analyzed, enabling a refined understanding of RIS. A notable finding was the significant influence of production bias on point defects’ interactions with grain boundaries/surfaces, thereby playing a crucial role in RIS processes. Production bias alters the neutrality of these interactions, leading to a preferential absorption of one type of point defect by the boundary and consequent establishment of distinct surface-mediated patterns of point defects. These spatial patterns further result in non-monotonic spatial profiles of solute atoms near surfaces/grain boundaries, corroborated by experimental observations. In particular, a positive production bias, signifying a higher production rate of vacancies over interstitials, drives more Cr depletion at the grain boundary. Moreover, a temperature-dependent production bias must be considered to recover the experimentally reported dependence of RIS on temperature. The severity of radiation damage and RIS becomes more pronounced with increased production bias, dose rate, and grain size, while high temperatures or sink density suppress the RIS severity. Model predictions were validated against experimental data, showcasing robust qualitative and quantitative agreements. The findings pave the way for further exploration of these spatial dependencies in subsequent studies, aiming to augment the comprehension and predictability of RIS processes in alloys.

Investigations of microstructure and tensile properties of neutron irradiated 6061 aluminum alloys

Effects of neutron irradiation on the microstructure and mechanical properties of CMRR (China Mianyang Research Reactor) structural material pure aluminum and 6061 alloy were studied by scanning/transmission electron microscopy (S/TEM) and tensile tests. The fast neutron (E > 0.1 MeV) fluences of 1.09–2.89 × 1022 n•cm−2 produced damage doses of 1.6–4.2 displacements per atom (dpa). Voids and faulted 1/3<111> dislocation loops were observed in both pure aluminum and 6061 alloy. In particular, voids tended to preferentially aggregate on coarse Cu-rich and fine Mg2Si precipitates which contributes to the higher irradiation tolerance in terms of void formation for 6061 alloy, suggesting that the uniformly distributed precipitates can act as trapping sites for defect clusters during irradiation and thus promote recombination. The size and density of Mg2Si precipitates showed a slight increase and Cu-rich precipitates emerged in 6061 alloy after neutron irradiation. Moreover, radiation-induced segregation was also detected along the grain boundaries. After tensile testing, the pure aluminum showed a lower yield strength and larger uniform elongation than 6061 alloy, where a moderate uniform elongation was still reserved after irradiation to 4.2dpa. The severely reduced ductility in 6061 alloy is supposed to be caused by the transmutation product silicon which contributes to the increase of dispersed multiple precipitates after irradiation. Our results indicate that precipitates especially Mg2Si play a decisive role in void suppression, but are not suitable in diminishing the radiation hardening.

The effect of secondary phases on microstructure and irradiation damage in an as-built additively manufactured 316 L stainless steel with a hafnium compositional gradient

Additive manufacturing (AM) or rapid prototyping has become a crucial tool for reducing both cost and time while increasing efficiency in qualifying structural materials for reactor use. In this study, directed energy deposition (DED) was used to develop an as-built 316 L stainless steel sample with three regions of increasing Hf-dopant to study the effects of Hf on the irradiation response of the material. Morphological and microstructural changes were analyzed before and after 2 MeV proton irradiation at 360 °C to a damage of 2.5 dpa at ∼ 10 µm below the surface. The addition of Hf effectively suppressed radiation-induced damage (dislocation loops, radiation-induced segregation) due to enhanced point defect recombination. The radiation damage seen in the as-built sample was further compared to a thermo-mechanically treated counterpart of the same fabrication and irradiation parameters which was found to behave superiorly. The increased radiation resistance of this material may be attributed to the as-built microstructure, which includes undissolved Hf particles, delta ferrite grains and cellular sub-grain boundaries that can hinder defect motion.

The time dependence of proton irradiation effect on the intergranular oxidation of 316 L stainless steel in high-temperature hydrogenated water

The time dependence of proton irradiation effect on the intergranular oxidation of 316 L stainless steel in simulated PWR primary water was clarified for the first time by ruling out the interference from grain boundary structure dependence. Interestingly, proton irradiation has an acceleration effect on the intergranular oxidation of 316 L stainless steel after shorter-term oxidation (less than 500 h) but a mitigation effect at later stage (more than 500 h). It was found that Cr depletion at the pristine grain boundary due to radiation-induced segregation (RIS) plays a dominant role at the early stage of oxidation, resulting in a deeper intergranular oxidation than in the non-irradiated region. Nevertheless, the Cr content at the intergranular oxide tip in the irradiated region builds up faster than in the non-irradiated region with time due to the enhanced solute (especially Cr) diffusivity in the irradiated region. The positive effect of enhanced Cr diffusivity on resistance to intergranular oxidation gradually dominates after longer-term immersion and eventually leads to a shallower intergranular oxide penetration in the irradiated region. The faster intergranular oxidation in irradiated region due to RIS at the early stage partly explains the accelerated crack propagation of irradiated material as the oxidation condition at the crack tip is similar to that at the early stage.

Convection-Induced Compositional Patterning at Grain Boundaries in Irradiated Alloys.

We consider the stability of precipitates formed at grain boundaries (GBs) by radiation-induced segregation in dilute alloys subjected to irradiation. The effects of grain size and misorientation of symmetric-tilt GBs are quantified using phase field modeling. A novel regime is identified where, at long times, GBs are decorated by precipitate patterns that resist coarsening. Maps of the chemical Péclet number indicate that arrested coarsening takes place when solute advection dominates over thermal diffusion right up to the precipitate-matrix interface, preventing interfacial local equilibrium and overriding capillary effects. This contrasts with liquid-solid mixtures where convection always accelerates coarsening.

Impact of defect annihilation mechanism on the grain size dependence and the dimensionality effect of radiation induced segregation

Radiation-induced segregation (RIS) of alloying elements at defect sinks critically affects material performance. This work computes and compares RIS at grain boundaries in a Fe-Ni-Cr alloy for a wide range of grain sizes with 1D, 2D, and 3D models. A transition in grain size dependence from linear to gradual convergence is identified, along with a dimensionality effect in modeling RIS using 1D and 2D models compared with more realistic 3D models. Further analysis shows that both the grain size dependence and the dimensionality effect are governed by the defect annihilation mechanism. In the absorption dominated regime, 1D and 2D models overestimate RIS, and RIS depends linearly on grain size. In the recombination dominated regime, the dimensionality effect vanishes gradually, and RIS converges gradually over grain size. The dimensionality effect can be corrected using the ratio of recombination over sink absorption, allowing for realistic RIS simulation using the 1D model.

Contributions of Ni-content and irradiation temperature to the kinetic of solute cluster formation and consequences on the hardening of VVER materials

The study of solute clusters induced by neutron irradiation is fundamental to the understanding of the embrittlement phenomena of Reactor Pressure Vessel (RPV) steels. In this paper, the influence of Ni-content and irradiation temperature on these cluster formation are investigated using atom probe tomography.VVER 1000-BM (base metal) and 1000-W (weld) steels were chosen for their high Ni content of 1.2 and 1.7 at%, respectively. They were irradiated up to doses of 0.28 dpa at two temperatures (265 and 300 °C). The irradiations were performed at SCK CEN in the BR2 reactor.After irradiation, Mn, Ni and Si rich clusters were observed. A careful study of their composition indicates that these clusters actually contain iron. Moreover, the evolution of Fe cluster concentration with the dose indicates a constant supply of solutes throughout the irradiation thanks to the flux coupling. With increasing dose, both the cluster size and number density increase.The effect of temperature has to be dissociated from that of Ni content. Both of them result in an increase of the cluster number density at a given dose. The irradiation temperature has a more pronounced effect than chemical composition (mainly Ni effect). This can be explained by the fact that more matrix damage is produced at lower irradiation temperature. The influence of Ni on cluster number density is significantly higher at the lower temperature. All these observations support the hypothesis of radiation-induced segregation, even if a thermodynamic contribution to solute clustering cannot be excluded, in particular in the higher Ni steel irradiated at low temperature.In all cases, the yield strength increases proportionally to the volume fraction of radiation-induced defects (√(Vf)).

Atomistic investigation on radiation-induced segregation in W-Re alloy under cumulative flux irradiation: A hybrid simulation with MD and VC-SGC-MC

Due to its great performance, tungsten (W) and W-based alloys have been considered as the candidates for plasma-faced materials in future fusion reactors. However, the radiation-induced segregation under the reactor environment is one of the key issues to limit the long-term service properties of W-based alloys, and the understanding of nucleation and growth mechanisms of Re-rich precipitates is still incomplete. In the present study, both molecular dynamics (MD) and variance-constrained semi-grand canonical ensemble Monte Carlo (VC-SGC-MC) methods are employed to investigate the nucleation and growth mechanisms of Re clusters precipitated in W-Re alloys under cumulative flux irradiation. It is found that all segregated Re-rich clusters in the model alloys without irradiation are a plate-like structure (χ-phase) which orientates along the <110>, while they change to a needle-like structure (σ-phase) orienting along the <111> direction after irradiation. The combination of defects and Re atoms in W-Re alloys are considered with the most probable nucleation precursors. The Re-rich clusters will pin the interstitials and form stable nucleation sites. During the process of cumulative flux irradiation, newborn defects will be trapped by the nucleation site promoting the growth of Re clusters.

Transmission electron microscopy investigation of phase transformation and fuel constituent redistribution in neutron irradiated U-10wt.%Zr fuel

Uranium-10 wt.% zirconium (U-10 wt.%Zr) is a primary candidate for fast reactor nuclear fuels. However, there is a lack of data characterizing the crystallographic phases and chemistry of the neutron irradiated fuel. In the current study, the microstructural evolution of a U-10 wt.%Zr fuel neutron irradiated to a burnup of 5.7 at.% was investigated using scanning transmission electron microscopy, energy dispersive X-ray spectroscopy, and selected area electron diffraction to determine the major phases, alterations in microstructure, and variations in local chemical composition at different localities of a fuel cross-section. The current study revealed that the irradiated U-10 wt.%Zr fuel was comprised of various major phases, including α-U, β-U, and δ-UZr2, as well as amorphous and crystalline solid fission product (FP) precipitates within different regions of the fuel cross-section. Regions A and A/B in the center of the fuel were comprised of U-rich, U-intermediate, and U-lean localities with α-U and δ-UZr2 composing the major phases. Regions B and C in the intermediate and peripheral fuel localities, respectively, were comprised of α-U grains, U-rich (β-U) grains with Zr-rich precipitates, U-intermediate grains (δ-UZr2), and solid FP precipitates. The major phases identified were associated with the nanoscopic chemical concentrations, the phase diagram, and the as-characterized specimen temperature, with the exception of the non-equilibrium β-U phase identified in Region B. The U-lean localities in Regions A, A/B, B, and C were Zr-enriched pathways along subgrain/grain boundaries, suggesting that Zr is susceptible to radiation-induced segregation in U-Zr fuels and indicating a new mechanism for constituent redistribution.