Radiation-induced Segregation Research Articles

Investigating radiation-induced segregation on intragranular defects in a 316L(N)

Radiation induced segregation (RIS) on lattice defects and phase stability under irradiation have been investigated in an optimized solution annealed 316L(N) with an increased content of N and a significant amount of Nb. Primary Z-phase nitrides were observed within the as-received microstructure. 5 MeV Fe3+ ion irradiation was conducted at 450 °C up to 1dpa. Post-ion irradiation characterization was performed via nanoscale techniques coupling and correlation to associate radiation induced defects such as cavities and dislocation loops to the local chemistry. Cr, Fe and Ni observed RIS trends were expected, but P was solely measured at the vicinity of cavities. Void-precipitate association has been detected thanks to transmission electron microscopy (TEM) and it is believed to explain the low reduced atomic density measured at the location of a feature within an atom probe tomography (APT) volume. Nb-rich novel features were observed after irradiation and first results are brought to establish a link between these features and the primary Z-phase.

Phase stability and microstructural evolution in neutron-irradiated ferritic-martensitic steel HT9

Ferritic Martensitic (F/M) steel HT9 specimens were irradiated in the Advanced Test Reactor up to 4.16 dpa in three temperature ranges (roughly from 300 to 600 °C). The post-irradiation microstructure, including dislocation structure , precipitation and radiation-induced segregation (RIS) was characterized using analytical scanning / transmission electron microscopy (S/TEM), and atom probe tomography (APT). Irradiation hardening was measured using nanoindentation . The results reveal a distinctive pattern of dislocation and precipitate evolution at high temperature, around 600 °C, where various defects and precipitates formed in the low dose regime followed by a recovering process with increasing dose. Dislocation loops formed in all temperature ranges, and the growth of dislocation loops is unconstrained above certain critical temperature, contributing to the increasing dislocation density even prior to doses of 0.5 dpa at 600 °C. Ni/Mn/Si clusters were identified in all temperature ranges and the compositions of these clusters converged to G phase stoichiometrically. Significant coarsening of G phase particles was observed at 600 °C, accompanied by the formation of G phase on grain boundaries. α ’ precipitates were only found in the medium and low temperature ranges (below 500 °C). The number density and volume fraction were higher in the low temperature specimens, while larger particles were observed in the medium temperature range. RIS of Cr, Ni, Mn, Si, P was identified at dislocation lines , grain boundaries and phase boundaries, and the temperature dependence is consistent with previous studies. The RIS of Cr to the existing VN particles was confirmed by APT and may accelerate the transition of VN to Cr-rich nitrides . The irradiation hardening contribution from dislocation loops, dislocation lines, G phase and α ’ phase was parsed based on a linear dispersed barrier hardening model. The results suggest that most irradiation hardening at high temperature is due to increasing dislocation density with dose.

Enhanced irradiation tolerance of Fe30Cr25Ni20Co15Mn10 high-entropy alloy via nanotwin boundaries

• There is no nanoscale precipitation and irradiation-induced hardening in Fe 30 Cr 25 Ni 20 Co 15 Mn 10 HEA at a high dose irradiation (∼ 70 dpa). • The HEA presents a very low swelling rate of ∼ 0.002%/dpa. • The RIS is observed for the first time at nanotwin boundaries in the HEA after irradiation. • The irradiation-induced segregation level is obviously suppressed in the HEA due to the sluggish atom diffusion. High-entropy alloys (HEAs) have been considered as a promising candidate for nuclear engineering due to their intriguing irradiation tolerance. The microstructure stabilities, including the radiation-induced precipitation (RIP), swelling, and segregation (RIS) in HEAs, are of great significance for applications. In this paper, the Fe 30 Cr 25 Ni 20 Co 15 Mn 10 HEA was irradiated by high-energy Au 2+ ions with a peak dose of ∼ 70 displacements per atom (dpa). The formation of voids, nanoscale precipitation, segregation, and mechanical properties before and after irradiation was characterized carefully, using scanning/transmission electron microscopy (S/TEM), atom probe tomography (APT), and nanoindentation. The results indicate that the HEA presents a very low swelling rate of ∼ 0.002%/dpa. After irradiation, no nanoscale precipitation is observed, confirming the excellent stability of the HEA microstructure. The RIS is observed for the first time at nanotwin boundaries in the HEA after irradiation. The elements, Co and Ni, are enriched slightly, while Fe, Cr, and Mn depleted on the nanotwin boundaries after irradiation. The irradiation-induced segregation level is obviously suppressed in the HEA due to the sluggish atom diffusion. The hardness remains almost unchanged after irradiation. The nanotwin boundaries act as defect sinks to absorb irradiation-induced vacancies and interstitials, reducing the swelling rate and relieving hardening. The low swelling, weak RIS level, and almost no hardening after a high dose irradiation indicate that the Fe 30 Cr 25 Ni 20 Co 15 Mn 10 HEA possesses an outstanding irradiation tolerance.

Impact of the local microstructure fluctuations on radiation-induced segregation in dilute Fe-Ni and Ni-Ti model alloys: A combined modeling and experimental analysis

From a systematic atom probe tomography (APT) characterization of the radiation-induced segregation (RIS) in dilute Fe-Ni and Ni-Ti model alloys, we highlight fluctuations of the solute local concentration up to the scale of the APT specimens. We deduce the RIS at dislocation loops from a solute diffusion equation, that is solved at steady state, within the Voronoi’s volume occupied by a single loop. From a statistical sampling of the Voronoi’s volume and the dislocation loop radius modeled after the characterization of the microstructure by transmission electron microscopy, we provide the full RIS distribution. The present statistical approach of RIS demonstrates that the fluctuation of local solute concentrations in Fe-Ni and Ni-Ti mainly results from the dispersion in size and density of the dislocation loop population. Besides, we highlight the impact of the post-treatment parameters used in the APT protocol on the extracted RIS profiles. In Ni-Ti alloys, the simulated Ti-depletion profiles are in very good agreement with the measured ones. Furthermore, the dispersion of the loop radius and density is shown to play a critical role on the fluctuations of the Ti local concentration. In Fe-Ni, the identification of discrepancies between the simulated Ni-enrichment profiles and the measured ones provides a signature of additional operating mechanisms of the solute redistribution, such as radiation-induced precipitation.

Effects of low-temperature neutron irradiation on the microstructure and tensile properties of duplex 2304 stainless steel and its electron-beam welds

A lean duplex stainless steel material (2304-grade) in its base metal and electron beam (e-beam) welded conditions were studied microstructurally and mechanically as a function of irradiation conditions to evaluate its use as a structural material at low temperatures (60–100 °C). Neutron irradiation up to a fluence of 1.40 × 1019 n/cm2 (E > 0.1 MeV) or ~0.011 dpa decreased the total elongation of both base metal and e-beam welded samples. Overall, radiation hardening was observed in all the samples. The transversely cut irradiated samples showed some nonuniform quasi-cleavage and shearing in their fracture surfaces, indicating the variance of ductile nature of the two-phased (deformable austenite and harder ferrite) duplex structure. The e-beam welded samples also showed quasi-cleavage fracture, which is a characteristic of radiation-induced embrittlement. These observations of the e-beam welded samples were attributed to the formation of coarse ferrites, grain boundary and intragranular phases such as γ2 and γ3, and minor impurity phases such as CrN and Cr2N in the weld pool and/or heat-affected zone of the samples. Radiation-induced elemental segregation was also identified in the post-irradiated base metal.

양성자 조사가 316 스테인리스강의 미세조직과 표면산화 특성에 미치는 영향

Austenitic 316 stainless steel was irradiated with protons accelerated by an energy of 2 MeV at 360 ℃, the various defects induced by this proton irradiation were characterized with microscopic equipment. In our observations irradiation defects such as dislocations and micro-voids were clearly revealed. The typical irradiation defects observed differed according to depth, indicating the evolution of irradiation defects follows the characteristics of radiation damage profiles that depend on depth. Surface oxidation tests were conducted under the simulated primary water conditions of a pressurized water reactor (PWR) to understand the role irradiation defects play in surface oxidation behavior and also to investigate the resultant irradiation assisted stress corrosion cracking (IASCC) susceptibility that occurs after exposure to PWR primary water. We found that Cr and Fe became depleted while Ni was enriched at the grain boundary beneath the surface oxidation layer both in the non-irradiated and proton-irradiated specimens. However, the degree of Cr/Fe depletion and Ni enrichment was much higher in the proton-irradiated sample than in the non-irradiated one owing to radiation-induced segregation and the irradiation defects. The microstructural and microchemical changes induced by proton irradiation all appear to significantly increase the susceptibility of austenitic 316 stainless steel to IASCC.

Microstructure-Dependent Rate Theory Model of Radiation-Induced Segregation in Binary Alloys

Conventional rate theory often uses the mean field concept to describe the effect of inhomogeneous microstructures on the evolution of radiation induced defect and solute/fission product segregation. However, the spatial and temporal evolution of defects and solutes determines the formation and spatial distribution of radiation-induced second phase such as precipitates and gas bubbles/voids, especially in materials with complicated microstructures and subject to high dose radiation. In this work, a microstructure-dependent model of radiation-induced segregation (RIS) has been developed to investigate the effect of inhomogeneous thermodynamic and kinetics properties of defects on diffusion and accumulations of solute A in AB binary alloys. Four independent concentrations: atom A, interstitial A, interstitial B, and vacancy on [A, B] sublattice are used as field variables to describe temporal and spatial distribution and evolution of defects and solute A. The independent concentrations of interstitial A and interstitial B allow to describe their different generation rates, thermodynamic and kinetic properties, and release the assumptions of interstitial generation and sink strength used in the conventional rate theory. Microstructure and concentration dependent chemical potentials of defects are used to calculate the driving forces of defect diffusions. With the model, the effects of defect chemical potentials and mobilities on the RIS in polycrystalline AB model alloys have been simulated. The results demonstrated the model capability in predicting defect evolution in materials with inhomogeneous thermodynamic and kinetic properties of defects. The model can be extended to materials with complicated microstructures such as a wide range of grain size distribution, coating structure and multiphases as well as radiation-induced precipitation subject to severe radiation damage.

Microstructure and microchemistry of laser welds of irradiated austenitic steels

This article investigates the integrity of laser welds on neutron irradiated, He-containing steels. Life extension of the current fleet of light water reactors could necessitate repair of cracks on irreplaceable internal components, but heat input of weld repairs exacerbates the problem by initiating He-induced cracking. Laser welding is a promising low-heat-input technology thought to limit the extent of He-induced cracking. In this study, we produce laser welds in a hot cell on AISI 304L stainless steel plates previously irradiated in the Experimental Breeder Reactor (EBR)-II. We select a systematic set of three specimens spanning fluences ~1–28 displacements per atom (dpa) at ~415–430 °C and He concentrations ~0.2–8 atomic parts per million (appm) amounting to ~0.2–2.8% swelling. He-induced cracking is observed only in specimens containing ≥3 appm He. Laser welding nearly eliminates all irradiation-induced cavities and reduces the dislocation loop number density, similar to conventional post-irradiation annealing. Microchemically, laser welding induces Cr-rich precipitation and suppresses grain boundary radiation-induced segregation. The mechanism of He-induced cracking is discussed in the context of these microchemical changes. The weld heat input is calculated and suggests that further refinement of laser welding parameters may improve the cracking resistance for higher dose and He conditions.

Precipitation in reactor pressure vessel steels under ion and neutron irradiation: On the role of segregated network dislocations

The ultrahigh density of Mn-Ni-Si-Cu precipitates that form in irradiated reactor pressure vessels (RPV) result in hardening and embrittlement, which may limit the service life of aging nuclear reactors. Here, we quantitatively analyze solute segregation to, and precipitation on, network dislocations in irradiated low alloy RPV steels. The analysis is based on a broad thermodynamic framework, which allows proper identification of possible non-equilibrium perturbative effects of irradiation, like radiation induced segregation (RIS). For the first time, we quantify the segregation of Cu, Mn, Ni and Si to ≈ 5–10 nm network dislocation segments, typically separated by a string-of-pearl type array of MnNiSi precipitates (MNSPs); the MNSPs are appended to Cu rich co-precipitates (CRPs) in steels bearing this element. Detailed atom probe tomography (APT) data are reported for 16 RPV steels, with a wide matrix of tailored compositions, irradiated by both neutrons and 6.4 MeV Fe3+ ions at 320 and 290 °C, respectively. A key aspect of this study is that the very high displacement per atom (dpa) doses and radiation enhanced diffusion (RED) result in nearly full phase separation, allowing isolation of thermodynamic (equilibrium or non-equilibrium) effects. We characterize segregation and precipitates with a comprehensive set of descriptors under both ion and neutron irradiation, including: a) solute enrichment factors (EFs) at dislocations, as a function of alloy composition; and, b) matrix versus dislocation precipitates and their respective characteristics. Formation of these dislocation features can be due to RIS, or thermodynamic in origin, or both. Coupled with simple models, we show that the nucleation MNSP features in low Cu steels is enhanced by segregation. However, in typical supersaturated RPV steels, rapid growth of the precipitates is thermodynamically driven to a quantitatively predictable, nearly full phase separation mole fraction.

The effects of irradiation on CrMnFeCoNi high-entropy alloy and its derivatives

High-entropy alloys (HEAs), based on equiatomic or near-equiatomic mixture of usually four or more elements, have attracted considerable attention as there are indications that along with surprising microstructural simplicity of some alloys, they may also offer intriguing combinations of mechanical and other properties. Amongst these properties, there is growing interest in the irradiation response of HEAs and their potential to withstand the neutron bombardment environment of future civil nuclear power plants. One of the first proposed HEAs is face-centred cubic CrMnFeCoNi, also known as the Cantor alloy, and the irradiation response of the Cantor alloy and its sub-systems are the focus of this review. Using irradiation analogues (electrons, heavy ions and He) to neutron bombardment and considering simulations, advanced microstructural analysis and property measurement, the Cantor alloy and its derivatives are shown to exhibit encouraging irradiation resistance that, in many instances, is superior to more traditional dilute alloys of the same elements. The beneficial aspects are high phase stability and resistance to radiation-induced segregation, smaller size but higher number density of dislocation loops, significantly lower extent of swelling and improved resistance to He bubble growth. The future research directions for irradiation resistant HEAs are also suggested.

Multiscale modeling of the effects of temperature, radiation flux, and sink strength on point-defect and solute redistribution in dilute Fe-based alloys

In this work, we investigate the radiation-induced segregation (RIS) resulting from the coupling between the atomic and point defect (PD) fluxes towards the structural defects of the microstructure. This flux coupling depends on the migration mechanisms of PDs and atoms, including thermal diffusion mechanisms and forced atomic relocations (FAR) occurring in displacement cascades. We derive an analytic model of the PD and solute RIS profiles accounting for PD production and mutual recombination, the FAR mechanism, and the overall sink strength of the microstructure controlling the elimination of PDs at structural defects. From this model, we present a parametric investigation of diffusion and RIS properties in dilute Fe-$B$ ($B$ = P, Mn, Cr, Si, Ni, and Cu) binary alloys, in the form of quantitative temperature/radiation flux/sink strength maps. As in previous works, we distinguish three kinetic domains for the diffusion and RIS properties: the recombination domain, the sink domain, and the thermal domain. Both our analytical approach and numerical applications demonstrate that the diffusion and RIS behaviors of PDs and solute atoms largely differ from one kinetic domain to another. Moreover, at high radiation flux, low temperature, and large sink strength, FARs tend to destroy the solute RIS profiles and therefore reduce the overall amount of RIS by forcing the mixing of solute and host atoms, especially close to PD sinks. Finally, we provide quantitative criteria to emulate in-reactor RIS behaviors by ion irradiation.

Correlative STEM-APT characterization of radiation-induced segregation and precipitation of in-service BWR 304 stainless steel

Radiation induced segregation and precipitation phenomena in an in-service boiling water reactor 304 stainless steel component were investigated using directly correlated 3D-atom probe tomography and scanning transmission electron microscopy. Significant quantitative differences in measured segregation at grain boundaries were found between the atom probe and energy dispersive spectroscopy measurements of the exact same locations. In particular, a much stronger Si segregation (~10 atomic% via atom probe versus ~4 atomic% via electron microscopy) and different Cr profile shapes were detected that are critical to models of radiation induced segregation and stress corrosion cracking behavior. These quantitative differences highlight the need for comparative microscopy and critical evaluation of limitations in each analytical method. Elemental segregation to dislocations and conjoined-clusters were also highlighted by atom probe; confirming and expanding upon what has been observed in test reactor neutron and accelerator-based ion irradiations.

Striking effect of solute elements (Mn, Ni) on radiation-induced segregation/precipitation in iron-based model alloys

The evolution of radiation damage in steels is a major issue for the safe operation of nuclear power plants. Mn and Ni contribution to the formation and evolution of microstructural features remains a controversial issue. The present study aims at investigating their effects on microstructure by irradiating undersaturated BCC Fe-3.3at.%Ni and Fe-2.8at.%Mn model alloys. Two different ion irradiation conditions have been applied to study the effect of irradiation dose rate in both of these Fe-based alloys at a temperature of 673 K. In all cases, irradiation promotes the formation of dislocation loops. In the FeNi alloy, radiation-induced segregation is observed on loops, leading to precipitation of a FCC phase in the low dose rate irradiation condition only. In the case of FeMn, even if no precipitation was revealed, highly enriched decorated dislocation loops are observed, with a higher Mn segregation at radiation defects than Ni.

Segregation behavior and phase instability of Eurofer97 after neutron irradiation to 72 dpa

A reduced activation ferritic/martensitic steel, Eurofer97, was neutron irradiated in the vicinity of 300 °C in the High Flux Isotope Reactor (HFIR) up to 72 dpa. Advanced analytical scanning transmission electron microscopy and conventional transmission electron microscopy were applied to investigate the radiation-induced segregation and phase instability behavior after neutron irradiation. Amorphization was observed in M23C6 carbides. Cr-rich clusters were seen within the matrix, near the lath boundaries and close to the M23C6 carbides. Cr enrichment and Fe depletion were detected at both prior austenite grain boundaries and lath boundaries, despite different segregation magnitude. In addition, the enrichment of Ni, the depletion of V, and tiny cavities (presumably helium bubbles) are also found at lath boundaries. This work interrogates the evolution of microstructures after neutron irradiation, which provides detailed understanding on the microstructural aspects controlling the mechanical integrity of Eurofer97 under high-dose neutron damage.

Solute-point defect interactions, coupled diffusion, and radiation-induced segregation in fcc nickel

Radiation-induced segregation (RIS) of solutes in materials exposed to irradiation is a well-known problem. It affects the life-time of nuclear reactor core components by favouring radiation-induced degradation phenomena such as hardening and embrittlement. In this work, RIS tendencies in face-centered cubic (fcc) Ni-X (X = Cr, Fe, Ti, Mn, Si, P) dilute binary alloys are examined. The goal is to investigate the driving forces and kinetic mechanisms behind the experimentally observed segregation. By means of ab initio calculations, point-defect stabilities and interactions with solutes are determined, together with migration energies and attempt frequencies. Transport and diffusion coefficients are then calculated in a mean-field framework, to get a full picture of solute-defect kinetic coupling in the alloys. Results show that all solutes considered, with the exception of Cr, prefer vacancy-mediated over interstitial-mediated diffusion during both thermal and radiation-induced migration. Cr, on the other hand, preferentially migrates in a mixed-dumbbell configuration. P and Si are here shown to be enriched, and Fe and Mn to be depleted at sinks during irradiation of the material. Ti and Cr, on the other hand, display a crossover between enrichment at lower temperatures, and depletion in the higher temperature range. Results in this work are compared with previous studies in body-centered cubic (bcc) Fe, and discussed in the context of RIS in austenitic alloys.

The Studies of Irradiation Hardening and RIS on IASCC of Reactor Internals Bolts

IASCC of stainless steel has been the most important issue for internals BFBs. The inspection data analysis indicates that there is a closed relation between irradiation fluence and cracked BFBs distribution. Then the nanoindentation and 3DAP tests were carried out to study the hardening and radiation induced segregation (RIS) behaviors of the reactor internals stainless steel specimens irradiated with 6 MeV Xe ions at room temperature. It is indicated that higher irradiation damage will cause more significant hardening and RIS and consequently increase the IASCC susceptibility.

Atom probe tomography characterization of high-dose ion irradiated MA957

Because the microstructure of oxide dispersion strengthened (ODS) alloys largely determines its mechanical properties in nuclear reactors, it is critical to understand the irradiation response of microstructure after high-dose radiation damage for safety and performance concerns. This paper presents a comprehensive characterization of microstructural responses to ion irradiation of the ODS ferritic alloy MA957 using atom probe tomography and compares the observations to current theories on mechanisms for microstructural evolution during irradiation. The MA957 specimens were irradiated by 1.8 MeV Cr3+ ions to doses and at temperatures relevant to advanced fast reactors (100 dpa and 500 dpa at 400–500°C). The stability of YTiO particles, formation of alpha prime (α′), and radiation-induced segregation were quantified and analyzed. The results show that the evolution of YTiO particles with respect to dose and temperature is consistent with the Nelson-Hudson-Mazey rate theory model that considers the competing effects of recombination and ballistic dissolution. The formation of Cr-rich precipitates (αʹ) was observed in specimens irradiated at temperatures up to 450°C, and it shows a strong dependence on temperature. Irradiation-induced segregation on grain boundaries was investigated in the 100 dpa specimens. In addition to the commonly studied element Cr, this study also analyzed segregation of Y and Ti. While the magnitude of segregation varies due to intrinsic variability of grain boundary character, clear trends with respect to irradiation temperature were observed. Overall, these results revealed relatively small but distinct effects of irradiation on microstructure evolution to 100 or 500 dpa, demonstrating the excellent radiation tolerance of ODS alloys.

Effects of grain boundary structures on primary radiation damage and radiation-induced segregation in austenitic stainless steel

Grain boundary (GB) engineering is crucial in the austenitic stainless steel (ASS) design for nuclear energy applications. In this work, the influence of different GB structures on radiation defect recombination and radiation-induced segregation (RIS) at different temperatures were investigated using molecular dynamics simulation. Four typical GBs in ASSs were selected as model structures. Results showed that GBs remained stable at various temperatures and they all exhibited better self-healing performance than single crystals in terms of radiation defects. However, except Σ3(112) GB, other three GBs cannot inhibit the radiation induced segregation, while promoting the radiation defect recombination. Calculation results showed that the higher Σ value of GBs can lead to a greater lattice mismatch near GBs, which not only results in stronger sink strength for radiation induced defects, but also provides more sites for solute atoms and causes greater segregations eventually. Owing to the intrinsic low Σ and large inclination angle characteristic, Σ3(112) GB achieves an excellent balance between the defect-absorption and RIS. This phenomenon provides a feasible route for the future GB design in ultra-high radiation tolerant materials.

Experimental and theoretical analysis of equilibrium segregation and radiation-induced segregation of Cr at grain boundaries in a reduced activation ferritic/martensitic (RAFM) steel

Helium ion irradiation at 350 °C was performed to study equilibrium segregation and radiation-induced segregation (RIS) of Cr at grain boundaries in reduced activation ferritic/martensitic steels. Cr concentration at grain boundary was measured by scanning transmission electron microscopy with an energy-dispersive spectrometer. The measured Cr concentration at grain boundaries in heat treated zone was 11.7 and 12.8 wt.% in irradiated zone, respectively, which matched well to the calculated results from Mclean and modified Perk model. Equilibrium segregation and RIS of Cr mechanisms were theoretically analysed. The analysis indicates that as temperature rises, equilibrium Cr segregation decreases monotonically, while RIS of Cr has a bell-shape profile, which increases first and then decreases. It is also shown that at low and high temperatures, equilibrium segregation of Cr is higher than that of RIS; at intermediate temperatures, equilibrium Cr segregation is lower than RIS.

Microstructural characterization of cold-worked 316 stainless steel flux thimble tubes irradiated up to 100 dpa in a commercial Pressurized Water Reactor

Two flux thimble tubes (FTTs) made of 15% cold-worked 316 stainless steel (SS) were harvested from Ringhals Pressurized Water Reactor (PWR) Unit 2, with peak damages of 76 and 100 displacements per atom (dpa) after 29 and 34 years’ service, respectively. Specimens sectioned from parent tubes were comprehensively characterized with nominal damage levels of ∼0, ∼41, ∼74, 76, and 100 dpa at a nominal temperature range of 285–323 °C. Both FTTs contained helium and hydrogen gases as transmutation products. The helium follows a production rate of ∼9.8 appm/dpa, while environmental factors complicate hydrogen production obscuring an exact H/dpa ratio. Irradiation-induced dislocation loops, nano-cavities, solute clusters, and microsegregation were all observed. The dislocation loops and nano-cavities indicated saturation at 41 dpa. The solute clusters continued to evolve with Ni–Si clusters formed at 41 dpa, and Ni–Si–Mn–P clusters formed at 74 and 100 dpa, but neither clusters exhibited distinct diffraction patterns at any damage levels. Solute clusters were observed to frequently be co-located with dislocation loops, but fully decorated loops were rarely detected. Significant radiation-induced segregation (RIS) was observed around grain boundaries at all damage levels. The modified inverse Kirkendall (MIK) model captured the RIS behavior of major elements. Large cavities within or around an Mn–S rich region were observed for the first time. Through all the damage levels, void swelling is always below 0.05%, making significant dimensional change unlikely in core internals when used at similar conditions. Meanwhile, the role of overwhelming nanocavities, presumably helium bubbles, should be considered in other potential degradation mechanisms, including irradiation-assisted stress corrosion cracking, embrittlement, and loss of fracture toughness, which remain the concerns for extended operation of nuclear power plants.