Radiation-induced Segregation Research Articles

Radiation tolerance of commercial and advanced alloys for core internals: a comprehensive microstructural characterization

Thirteen austenitic stainless steels, nickel-base alloys, and ferritic alloys were irradiated using 2 MeV protons at 360 °C to a damage level of 2.5 displacements per atom (dpa). Comprehensive microstructural characterization was performed for irradiation-induced features, including dislocation loops, voids, precipitates, and radiation induced segregation (RIS). Dislocation loops formed in all alloys except 14YWT, while voids were observed in alloys 316 L, 310, C22, and 14YWT. Irradiation-induced formation of γ′ precipitates was observed in alloys 316 L, 310, 800, and 690; the irradiation-enhanced, long-range ordered Ni2Cr phase (Pt2Mo-type) was observed in alloys 690, C22, 625, 625Plus, 625DA, and 725; and G-phase was observed in alloy T92. No irradiation-induced precipitates were observed in alloys X750, 718 or 14YWT. Precipitation of the γ′ phase can be understood through segregation and clustering of Si, Al, and Ti. Overall, austenitic stainless steels are generally susceptible to irradiation damage in the form of loops, voids, precipitates, and RIS. Ni-base alloys have this same type of dislocation loops and RIS behaviors but are more resistant to void swelling. Ferritic alloys showed better resistance to loop formation, void swelling and irradiation-induced precipitation. From the degree of irradiation-induced microstructural change, alloy T92 was identified as the most radiation resistant among these alloys.

Effect of Si-Addition in Stainless Steel on Surface Oxidation in High Temperature Water Environments

Being used as the internal components of pressurized water reactor (PWR), austenitic stainless steels (SSs) are susceptible to irradiation-assisted stress corrosion cracking (IASCC) that is a critical issue for the long term safe operation. Radiation induced segregation (RIS) at grain boundaries in austenitic SSs, such as the depletion of Cr and enrichment of Ni and Si, is thought to contribute to IASCC. Studying the oxidation behavior of the model alloys simulating grain boundary RIS in high temperature water could provide evidence for the effect of RIS on IASCC. In this study, the model alloys 9Cr-25Ni-1.5Mo-5Si and 9Cr-25Ni-1.5Mo-3Si were used to simulate the depletion of Cr and enrichment of Ni and Si at grain boundaries of irradiated 316L SS. The oxide film properties of 316L SS and model alloys in of high temperature water environments were investigated. After the immersion in high-temperature water environments, the surface morphologies of the oxide films formed on the immersed specimens were examined using JEOL JSM-7500F field emission scanning electron microscope (FESEM). The oxide films were analyzed using INVIA laser Raman spectrometer with laser wavelength of 514.5 nm. The compositions of oxide particles were analyzed by the energy-dispersive spectrometry (EDS) attached to the FESEM. Fig.1 shows the SEM morphologies of the 316L SS, 9Cr-25Ni-1.5Mo-3Si and 9Cr-25Ni-1.5Mo-5Si alloy surfaces after immersion in pure water at 290 oC (pH at 290 oC of about 5.6) for 120h. The surfaces of all the test specimens were covered with polyhedral particles of different sizes. Larger size outer layer oxides are formed on 9Cr-25Ni-1.5Mo-3Si and 9Cr-25Ni-1.5Mo-5Si specimen surface. On the surface of 316L SS, there were only sparsely distributed large-size particles. Si addition in the alloy matrix favors the formation of the oxide layer in this environment. Fig. 2 shows the SEM morphologies of 316L SS, 9Cr-25Ni-1.5Mo-5Si and 9Cr-25Ni-1.5Mo-3Si alloys surfaces after the immersion in high temperature water of pH 9.5 (pH adjusted with monoethanolamine) at 290 oC for 24h, 48h and 120h. Compact and homogeneous oxide particles were formed on the surfaces of 316L SS and model alloys. The size of the oxide particles on model alloys was smaller than that on 316L SS. The oxide particle size of 9Cr-25Ni-1.5Mo-5Si and 9Cr-25Ni-1.5Mo-3Si was similar. The morphologies of particles became more stereoscopic with increasing immersion time. The outer layer oxide particles formed on the 316L SS and model alloys decreased with increasing solution pH value. The pH value has a significant influence on the oxidation of 316L SS and model alloys. These results show that the effect of Si in the alloy on oxide film properties is strongly dependent on the water chemistry. Si in the alloy facilitated the formation of large-sized outer layer in 290 oC water of pH 5.6, which refined the outer layer oxide particles in 290 oC water of pH 9.5. High solubility of Si-bearing oxide in high temperature water can modify the oxide film on the surface. These results show that Si in the alloy would play a role in IASCC through the modification of oxide film. Figure 1

Multiscale modeling of Radiation Induced Segregation in iron based alloys

In metallic alloys of fission or fusion reactors, microstructural evolution results from a dynamic equilibrium between thermodynamic forces and the production of defects by irradiation. The migration of defects can lead to the formation of clusters of defects (nano-cavities, dislocation loops, nano-precipitates) or variations in chemical composition close to the defect sinks, a phenomenon known as Radiation Induced Segregation (RIS). To predict the effect of irradiation conditions (type of irradiation particle, dose rate, temperature), phenomenological diffusion models exist that have to be assembled to give an overall description. Our objective is to describe these models and to propose numerical implementations to solve them starting at the atomic scale (DFT energy calculations, Kinetic Monte Carlo, Self-Consistent Mean Field approaches) to reach the mesoscopic one using the phase field modeling. This multi-scale approach is illustrated by a short review of recent studies focusing on dilute and concentrated iron based alloys.

Microstructure evolution of T91 irradiated in the BOR60 fast reactor

Microstructures of T91 neutron irradiated in the BOR60 reactor at five temperatures between 376 °C and 524 °C to doses between 15.4 and 35.1 dpa were characterized using transmission electron microscopy (TEM), scanning transmission electron microscopy (STEM), and atom probe tomography (APT). Type a<100> dislocation loops were observed at 376–415 °C and network dislocations dominated at 460 °C and 524 °C. Cavities appeared in a bimodal distribution with a high density of small bubbles less than 2 nm at irradiation temperatures between 376 °C and 415 °C. Small bubbles were also observed at 460 °C and 524 °C but cavities greater than 2 nm were absent. Enrichment of Cr, Ni, and Si at the grain boundary was observed at all irradiation temperatures. Radiation-induced segregation (RIS) of Cr, Ni and Si appeared to saturate at 17.1 dpa and 376 °C. The temperature dependence of RIS of Cr, Ni and Si at the grain boundary, which showed a peak Cr enrichment temperature of 460 °C and a lower peak Ni and Si enrichment temperature of ∼400 °C, was consistent with observations of RIS of Cr in proton irradiated T91, suggesting that the same RIS mechanism may also apply to BOR60 irradiated T91. G-phase and Cu-rich precipitates were observed at 376–415 °C but were absent at 460 °C and 524 °C. The absence of G-phase at 524 °C could be related to the minimal segregation of Ni and Si in that condition.

Energetics analysis of interstitial loops in single-phase concentrated solid-solution alloys

Systematic energetics analysis on the shape preference, relative stability and radiation-induced segregation of interstitial loops in nickel-containing single-phase concentrated solid-solution alloys have been conducted using atomistic simulations. It is shown that the perfect loops prefer rhombus shape for its low potential energy, while the Frank faulted loops favor ellipse for its low potential energy and the possible large configurational entropy. The decrease of stacking fault energy with increasing compositional complexity provides the energetic driving force for the formation of faulted loops, which, in conjunction with the kinetic factors, explains the experimental observation that the fraction of faulted loops rises with increasing compositional complexity. Notably, the kinetics is primarily responsible for the absence of faulted loops in nickel-cobalt with a very low stacking fault energy. We further demonstrate that the simultaneous nickel enrichment and iron/chromium depletion on interstitial loops can be fully accounted for by their energetics.

Self-ion emulation of high dose neutron irradiated microstructure in stainless steels

Solution-annealed 304L stainless steel (SS) was irradiated to 130 dpa at 380 °C, and to 15 dpa at 500 °C and 600 °C, and cold-worked 316 SS (CW 316 SS) was irradiated to 130 dpa at 380 °C using 5 MeV Fe++/Ni++ to produce microstructures and radiation-induced segregation (RIS) for comparison with that from neutron irradiation at 320 °C to 46 dpa in the BOR60 reactor. For the 304L SS alloy, self-ion irradiation at 380 °C produced a dislocation loop microstructure that was comparable to that by neutron irradiation. No voids were observed in either the 380 °C self-ion irradiation or the neutron irradiation conditions. Irradiation at 600 °C produced the best match to radiation-induced segregation of Cr and Ni with the neutron irradiation, consistent with the prediction of a large temperature shift by Mansur's invariant relations for RIS. For the CW 316 SS alloy irradiated to 130 dpa at 380 °C, both the irradiated microstructure (dislocation loops, precipitates and voids) and RIS reasonably matched the neutron-irradiated sample. The smaller temperature shift for RIS in CW 316 SS was likely due to the high sink (dislocation) density induced by the cold work. A single self-ion irradiation condition at a dose rate ∼1000× that in reactor does not match both dislocation loops and RIS in solution-annealed 304L SS. However, a single irradiation temperature produced a reasonable match with both the dislocation/precipitate microstructure and RIS in CW 316 SS, indicating that sink density is a critical factor in determining the temperature shift for self-ion irradiations.

MODELING RADIATION-INDUCED SEGREGATION IN BINARY ALLOYS

A computer code to calculate the time and space dependencies of the defect and component concentrations of a binary alloy is developed. This code is based on the theoretical model for radiation-induced segregation governed by first and second Fick’s laws with taking into account inverse Kirkendall effect. The system of three coupled partial differential equations for concentrations of one of components and point defects is converted to the system of ordinary differential equations in time by a discretization procedure. Numerical solutions of this system are obtained under appropriate initial and boundary conditions by means of the MATLAB. The modeling of radiation-induced segregation in binary Fe-Cr alloys is carried out for various initial concentrations of components, temperatures, dose rates and doses. The process of achievement of steady state at dose rising is demonstrated. The calculated concentration profiles are compared with experimental profiles published in literature. The sensitivity of model to input parameters is done and the capabilities of proposed model are estimated. This computer code for multicomponent alloys is developed.

Effect of post-irradiation annealing on the irradiated microstructure of neutron-irradiated 304L stainless steel

Post-irradiation annealing was performed on a 304L SS that was irradiated to 5.9 dpa in the Barsebäck 1 BWR reactor. Evolution of dislocation loops, radiation-induced solute clusters and radiation-induced segregation at the grain boundary was investigated following thermal annealing at 500 °C and 550 °C up to 20 h. Dislocation loops, Ni-Si and Al-Cu clusters, and enrichment of Ni, Si and depletion of Cr at the grain boundary were observed in the as-irradiated condition. Dislocation loop size did not change significantly after annealing at 550 °C for 5 h but the loop number density decreased considerably and loops mostly disappeared after annealing at 550 °C for 20 h. The average size of Ni-Si and Al-Cu clusters increased while the number density decreased with annealing. The increase in cluster size was due to diffusion of solutes rather than cluster coarsening. Significant volume fractions of Ni-Si and Al-Cu clusters still remained after annealing at 550 °C for 20 h. Substantial recovery of Cr and Ni at the grain boundary was observed after annealing at 550 °C for 5 h but neither Cr nor Ni was fully recovered after 20 h. Annihilation of dislocation loops, driven by the thermal vacancy concentration gradient caused by the strain field and stacking fault associated with the loops appeared to be faster than annihilation of solute clusters and recovery of Ni and Si at the grain boundary, both of which are driven by the solute concentration gradients.

Atomistic study of the hardening of ferritic iron by Ni-Cr decorated dislocation loops

The exact nature of the radiation defects causing hardening in reactor structural steels consists of several components that are not yet clearly determined. While generally, the hardening is attributed to dislocation loops, voids and secondary phases (radiation-induced precipitates), recent advanced experimental and computational studies point to the importance of solute-rich clusters (SRCs). Depending on the exact composition of the steel, SRCs may contain Mn, Ni and Cu (e.g. in reactor pressure vessel steels) or Ni, Cr, Si, Mn (e.g. in high-chromium steels for generation IV and fusion applications). One of the hypotheses currently implied to explain their formation is the process of radiation-induced diffusion and segregation of these elements to small dislocation loops (heterogeneous nucleation), so that the distinction between SRCs and loops becomes somewhat blurred. In this work, we perform an atomistic study to investigate the enrichment of loops by Ni and Cr solutes and their interaction with an edge dislocation. The dislocation loops decorated with Ni and Cr solutes are obtained by Monte Carlo simulations, while the effect of solute segregation on the loop's strength and interaction mechanism is then addressed by large scale molecular dynamics simulations. The synergy of the Cr-Ni interaction and their competition to occupy positions in the dislocation loop core are specifically clarified.

Radiation-induced swelling and radiation-induced segregation & precipitation in dual beam irradiated Ferritic/Martensitic HT9 steel

Ferritic/Martensitic HT9 steel was irradiated at 432°C to 16.6 displacements per atom (dpa) (at the depth of 600nm) using a defocused beam of 5MeV Fe++ ions, while co-implanted with 3.22appm He at the same depth. The helium concentration profile was designed so to follow the damage curve with a 0.22appmHe/dpa ratio at the depth from 300 to 1000nm in the material. The depth-dependence of the cavity size and number density were characterized by both transmission electron microscopy (TEM) and scanning transmission electron microscopy (STEM) imaging methods. A comparison between the two techniques was done showing good agreement. Cavity number density and the resulting swelling were found to be suppressed by the injected interstitial effect in the vicinity of the ion induced damage peak. The region between 300 and 750nm depth which excludes the injected interstitial effect was thus proposed for improved cavity swelling analysis. The swelling ratio in this region was found to be ~(0.86–1.02)×10−2%/dpa. In addition, ChemiSTEM characterization revealed radiation-induced segregation occurring throughout the irradiated region and precipitation of G-phase particles. Segregation of Ni to cavity surfaces was also observed and its possible synergistic influence on swelling was discussed.

Radiation-induced segregation at grain boundaries in AL-6XN stainless steels irradiated by hydrogen ions

The effect of high concentration of hydrogen on the segregation of radiation-induced segregation (RIS) in AL-6XN stainless steels has been investigated by transmission electron microscopy (TEM) with energy-dispersive X-ray spectroscopy. Specimens were irradiated with 100keV H2+ ions from 1dpa to 5dpa at 380°C to investigated the dose dependence of grain boundary RIS. A specimen was irradiated to 5dpa at 290°C to study the effect of irradiation temperature. The trends of Cr depletion and Ni enrichment with irradiation dose is similar to that of other austenitic steels reported in the literatures, but the higher concentration of hydrogen made the RIS profile wider. An abnormal phenomenon that the degree of RIS increased with decreasing irradiation temperature was found, indicating that with the retention of hydrogen in the steels, temperature dependence of RIS is dominated by the quantity of retained hydrogen, rather than by thermal segregation processes.

Helium effects on IASCC susceptibility in as-implanted solution annealed, cold-worked and post-implantation annealed 316L steel

ABSTRACTThe susceptibility to irradiation-assisted stress corrosion cracking (IASCC) at relatively low dose is dominated by conventional mechanisms such as radiation-induced segregation, hardening, creep/relaxation and radiolysis. Large amount of helium (He) can be accumulated in reactor internal components made of nickel containing steels of pressurised water reactors after long-term operation. The increase of He content has been correlated to IASCC susceptibility in reactor internals subjected to neutron irradiation. However, the question whether He increases or even dominates the IASCC susceptibility at high doses is still open. In this work, He was homogeneously implanted at 300°C to 300 and 1000 appm. Slow strain rate tests were performed in as-implanted solution annealed and cold-worked samples and in post-implantation annealed samples. The results show that homogenised implanted He up to 1000 appm alone (without significant displacement damage) does not produce intergranular cracking and IASCC in high-temperature air and in hydrogenated high-temperature water, respectively.

Cluster dynamics modeling of Mn-Ni-Si precipitates in ferritic-martensitic steel under irradiation

Mn-Ni-Si precipitates (MNSPs) are known to be responsible for irradiation-induced hardening and embrittlement in structural alloys used in nuclear reactors. Studies have shown that precipitation of the MNSPs in 9-Cr ferritic-martensitic (F-M) alloys, such as T91, is strongly associated with heterogeneous nucleation on dislocations, coupled with radiation-induced solute segregation to these sinks. Therefore it is important to develop advanced predictive models for Mn-Ni-Si precipitation in F-M alloys under irradiation based on an understanding of the underlying mechanisms. Here we use a cluster dynamics model, which includes multiple effects of dislocations, to study the evolution of MNSPs in a commercial F-M alloy T91. The model predictions are calibrated by data from proton irradiation experiments at 400 °C. Radiation induced solute segregation at dislocations is evaluated by a continuum model that is integrated into the cluster dynamics simulations, including the effects of dislocations as heterogeneous nucleation sites. The result shows that MNSPs in T91 are primarily irradiation-induced and, in particular, both heterogeneous nucleation and radiation-induced segregation at dislocations are necessary to rationalize the experimental observations.

The role of grain boundary microchemistry in irradiation-assisted stress corrosion cracking of a Fe-13Cr-15Ni alloy

A novel explanation based on the variation of grain boundary composition is proposed to elucidate the localized nature of irradiation-assisted stress corrosion cracking, i.e., intergranular cracking is observed only at specific sites of random high-angle grain boundaries, in a Fe-13Cr-15Ni austenitic model alloy subjected to proton irradiation and straining in a high-temperature water environment. Specifically, this work presents electron microscopy characterization of multiple cracked and un-cracked grain boundary sites and the neighboring oxides. The depletion of Cr and enrichment of Ni, as expected due to radiation-induced segregation, is observed at grain boundaries far from the crack tip (for cracked sites) or the metal surface (for un-cracked sites), and such modification of grain boundary composition is enhanced in the vicinity of the corrosion reaction front. Unexpectedly, grain boundary sites beyond the crack tip always present a lower Ni content and higher Cr content than the un-cracked sites on the same grain boundary. Overall, it is proposed that the site-specific susceptibility to stress corrosion cracking is governed by the grain boundary microchemistry, which determines not only the quantity of Cr, but also the efficiency of Cr transport to the reaction front and thus the protectiveness of the inner oxide. These effects of local composition may be further coupled with the local structure of and the local stresses interacting with the random high-angle grain boundaries.

Effect of irradiation temperature on microstructural changes in self-ion irradiated austenitic stainless steel

We investigated the microstructural and hardness changes in austenitic stainless steel after Fe ion irradiation at 400, 300, and 200 °C using transmission electron microscopy (TEM) and nanoindentation. The size of the Frank loops increased and the density decreased with increasing irradiation temperature. Radiation-induced segregation (RIS) was detected across high-angle grain boundaries, and the degree of RIS increases with increasing irradiation temperature. Ni–Si clusters were observed using high-resolution TEM in the sample irradiated at 400 °C. The results of this work are compared with the literature data of self-ion and proton irradiation at comparable temperatures and damage levels on stainless steels with a similar material composition with this study. Despite the differences in dose rate, alloy composition and incident ion energy, the irradiation temperature dependence of RIS and the size and density of radiation defects followed the same trends, and were very comparable in magnitude.

Microstructural evolution of neutron-irradiated T91 and NF616 to ∼4.3 dpa at 469 °C

Ferritic-martensitic steels such as T91 and NF616 are candidate materials for several nuclear applications. This study evaluates radiation resistance of T91 and NF616 by examining their microstructural evolutions and hardening after the samples were irradiated in the Advanced Test Reactor to ∼4.3 displacements per atom (dpa) at an as-run temperature of 469 °C. In general, this irradiation did not result in significant difference in the radiation-induced microstructures between the two steels. Compared to NF616, T91 had a higher number density of dislocation loops and a lower level of radiation-induced segregation, together with a slightly higher radiation-hardening. Unlike dislocation loops developed in both steels, radiation-induced cavities were only observed in T91 but remained small with sub-10 nm sizes. Other than the relatively stable M23C6, a new phase (likely Sigma phase) was observed in T91 and radiation-enhanced MX → Z phase transformation was identified in NF616. Laves phase was not observed in the samples.

Interpreting radiation-induced segregation and enhanced radiation tolerance of single-phase concentrated solid-solution alloys from first principles

Most recently, a host of irradiation experiments conducted on single-phase concentrated solid-solution alloys has been lacking interpretations from first principles. Here, the formation energetics and migration kinetics of interstitial defects in pure Ni and two Ni-based alloys (NiCo and NiFe) have been investigated by first principles calculations, based on which we are able to provide significant insights into two key aspects of experimental observations concerning the radiation-induced segregation and the enhanced radiation tolerance.

Application of the FRI crack growth model for neutron-irradiated stainless steels in high-temperature water of a boiling water reactor environment

This study considered a methodology to reflect the effect on material properties by neutron irradiation onto the FRI model of the crack growth rate (CGR) of stress corrosion cracking developed at Tohoku University. Yield strength and strain hardening exponent were evaluated by a tensile test for irradiated stainless steels, and the relationship between mechanical property and strain to fracture of the oxide film was derived. Effect of the radiation-induced segregation on CGR was also discussed. The experimental tendencies for the CGR to increase with dose and almost saturate above 3dpa were well replicated by the model calculations.

A review: applications of the phase field method in predicting microstructure and property evolution of irradiated nuclear materials

Complex microstructure changes occur in nuclear fuel and structural materials due to the extreme environments of intense irradiation and high temperature. This paper evaluates the role of the phase field method in predicting the microstructure evolution of irradiated nuclear materials and the impact on their mechanical, thermal, and magnetic properties. The paper starts with an overview of the important physical mechanisms of defect evolution and the significant gaps in simulating microstructure evolution in irradiated nuclear materials. Then, the phase field method is introduced as a powerful and predictive tool and its applications to microstructure and property evolution in irradiated nuclear materials are reviewed. The review shows that (1) Phase field models can correctly describe important phenomena such as spatial-dependent generation, migration, and recombination of defects, radiation-induced dissolution, the Soret effect, strong interfacial energy anisotropy, and elastic interaction; (2) The phase field method can qualitatively and quantitatively simulate two-dimensional and three-dimensional microstructure evolution, including radiation-induced segregation, second phase nucleation, void migration, void and gas bubble superlattice formation, interstitial loop evolution, hydrate formation, and grain growth, and (3) The Phase field method correctly predicts the relationships between microstructures and properties. The final section is dedicated to a discussion of the strengths and limitations of the phase field method, as applied to irradiation effects in nuclear materials.

Characterization of the martensite phase formed during hydrogen ion irradiation in austenitic stainless steel

Microstructural changes in austenitic stainless steel caused by hydrogen ion irradiation were investigated using transmission electron microscopy (TEM). It has been confirmed that the irradiation induced the formation of martensite along the grain boundary; the martensite phase exhibited a crystal orientation relationship with the adjacent austenite phase. The results of this study also indicate that the concentration of Cr in the martensite phase is lower compared to that in the austenite matrix. The TEM results showed the development of asymmetric radiation-induced segregation (RIS) near the grain boundary, which leads to local changes in the chemical composition such as reduction of Cr near the grain boundary. The asymmetric RIS serves as a prerequisite for the formation of the martensite under hydrogen irradiation.