Segregation of Cr at tilt grain boundaries in Fe-Cr alloys: A Metropolis Monte Carlo study

In this work, the Metropolis Monte Carlo (MMC) method employing the isothermal-isobaric statistical ensemble is applied to investigate segregation at grain boundaries in bcc Fe-Cr alloys with varying Cr content from 5 to 14 at.%. Several different 〈1 1 0〉 tilt grain boundaries, namely: Σ19{3 3 1}, Σ9{2 2 1}, Σ3{1 1 1}, Σ3{1 1 2}, Σ11{1 1 3}, Σ9{1 1 4} with misorientation angle varying in the range 26-141° were considered. Systematic MMC simulations were performed employing a two band empirical many-body potential in the temperature range 300-900 K. It was found that the binding energy of substitutional Cr to the GB core is essentially determined by the structure of the GB interface and varies in the range 0.05-0.35 eV. At this, the binding energy increases with the GB excess volume. MMC simulations revealed that either a local atomic rearrangement or segregation of Cr at the considered GBs occurs depending on the combination of temperature, alloy composition and GB structure. Influence of temperature and GB structure on the local atomic rearrangement and precipitation of α′ particles is demonstrated.     

Void swelling in a 9Cr ferritic/martensitic steel irradiated with energetic Ne-ions at elevated temperatures

In this work the void swelling behavior of a 9Cr ferritic/martensitic steel irradiated with energetic Ne-ions is studied. Specimens of Grade 92 steel (a 9%Cr ferritic/martensitic steel) were subjected to an irradiation of ²⁰Ne-ions (with 122 MeV) to successively increasing damage levels of 1, 5 and 10 dpa at a damage peak at 440 and 570 °C, respectively. And another specimen was irradiated at a temperature ramp condition (high flux condition) with the temperature increasing from 440 up to 630 °C during the irradiation. Cross-sectional microstructures were investigated with a transmission electron microscopy (TEM). A high concentration of cavities was observed in the peak damage region in the Grade 92 steel irradiated to 5 dpa, and higher doses. The concentration and mean size of the cavities showed a strong dependence on the dose and irradiation temperature. Enhanced growth of the cavities at the grain boundaries, especially at the grain boundary junctions, was observed. The void swelling behavior in similar 9Cr steels irradiated at different conditions are discussed by using a classic void formation theory.     

Effect of impurity and alloying elements on Zr grain boundary strength from first-principles computations

Effects of twenty impurity and alloy elements on the strength of a Zr(0 0 0 1)/Zr(0 0 0 1) ∑7 twist grain boundary were studied using a first-principles density functional approach. A ranking in the order of most weakening to most strengthening was: Cs, I, He, Te, Sb, Li, O, Sn, Cd, H, Si, C, N, B, U, Ni, Hf, Nb, Cr, and Fe. Segregation energies for these elements to the grain boundary and the Zr(0 0 0 1) surface were also calculated. Calculations showed that the weakening grain boundary elements He, I, and Cs have a strong driving force for segregation to the grain boundary from bulk Zr. Zircaloy cladding failures (pellet-clad interactions) in commercial fuel systems and separate effects test results provide context for these computational results.     

The effect of Cr concentration on radiation damage in Fe-Cr alloys

Displacement cascades in Fe-Cr alloys were studied using molecular dynamics computer simulations. We considered random Fe-5Cr and Fe-15Cr alloys, as well as Fe-10Cr alloys with and without Cr-rich precipitates. In the simulations two versions of a two-band embedded atom method potential were used, and the cascades were induced by recoils with energies up to 20 keV. We found that the average number of surviving Frenkel pairs and the fraction of vacancies and self-interstitials in clusters was approximately the same in pure Fe and random Fe-Cr alloys (regardless of Cr concentration). A noticeable effect of the presence of Cr in the Fe matrix was only observed in the enrichment of self-interstitials by Cr in Fe-5Cr. The calculated change in the short range order parameter showed that Fe-5Cr tends towards ordering (negative short range order parameter) and Fe-15Cr towards segregation (positive short range order parameter) of Cr atoms. In simulations with the Cr-rich precipitate, enhanced cascade splitting and segregation of self-interstitial defects created inside the precipitates towards the precipitate-matrix interface region was observed. The number of Frenkel pairs and their clustered fraction was not affected by the presence of the precipitate.     

Grain boundary diffusion and segregation of Cr in α-Zr

⁵¹Cr diffusion along grain boundaries in polycrystalline α-Zr was measured by means of the radiotracer technique in the temperature range 449-680 K. The use of Harrison´s C and B kinetics provided direct data about grain boundary diffusivity (D_gb) and the apparent grain boundary diffusivity (P_gb) in the temperature range of power reactors service. The grain boundaries segregation factor s of Cr in α-Zr was determined at the limit of very dilute solute concentration.     

Modeling solute-vacancy trapping at oversized solutes and its effect on radiation-induced segregation in Fe-Cr-Ni alloys

Rate theory modeling was used to simulate the effects of oversized solute additions on radiation-induced segregation in austenitic stainless steels. The purpose was to understand the effects of a solute-vacancy trapping mechanism on radiation-induced segregation and to define key parameters that most affect segregation behavior. Sensitivity analysis of the model showed the solute-vacancy binding energy to be the most important model parameter. Binding energies from ab initio first principles were calculated for oversized solutes of Pt, Ti, Hf and Zr, with energies of 0.31, 0.39, 0.71 and 1.08 eV, respectively. Differences in binding energies, despite similar sizes of the atoms, suggests that the short-range electronic interactions play an important role in determining binding energy. The model results show oversized solutes to be most effective at reducing grain boundary Cr depletion at temperatures of 450-500 °C for a dose rate applicable to proton irradiations. The reduction increases with increasing oversized solute concentration, where it saturates at approximately 0.1 at.%.     

Effects of oversized solutes on radiation-induced segregation in austenitic stainless steels

Zirconium or hafnium additions to austenitic stainless steels caused a reduction in grain boundary Cr depletion after proton irradiations for up to 3 dpa at 400 °C and 1 dpa at 500 °C. The predictions of a radiation-induced segregation (RIS) model were also consistent with experiments in showing greater effectiveness of Zr relative to Hf due to a larger binding energy. However, the experiments showed that the effectiveness of the solute additions disappeared above 3 dpa at 400 °C and above 1 dpa at 500 °C. The loss of solute effectiveness with increasing dose is attributed to a reduction in the amount of oversized solute from the matrix due to growth of carbide precipitates. Atom probe tomography measurements indicated a reduction in amount of oversized solute in solution as a function of irradiation dose. The observations were supported by diffusion analysis suggesting that significant solute diffusion by the vacancy flux to precipitate surfaces occurs on the time scales of proton irradiations. With a decrease in available solute in solution, improved agreement between the predictions of the RIS model and measurements were consistent with the solute-vacancy trapping process, as the mechanism for enhanced recombination and suppression of RIS.     

Atomic scale investigation of radiation-induced segregation in austenitic stainless steels

Radiation-induced precipitation and segregation in a cold-worked 316 austenitic stainless steel irradiated with 10 MeV Fe⁵⁺ ions were characterized by atom probe tomography. Ni and Si enrichment and Cr depletion were observed in roughly spherical and torus-shaped clusters, believed to be due to solute enrichment and depletion at dislocation loops. Solute segregation was also observed at network dislocations. These observations are consistent with the phenomenon of radiation-induced segregation. Radiation-induced segregation at grain boundaries was also studied at the near atomic scale. Comparison of these observations with results from the literature shows a difference in the magnitude of the peak concentration of segregated solutes.     

First-principles investigation of energetics and site preference of He in a W grain boundary

We perform first-principles calculations based on density functional theory to investigate energetics and site preference of He in a bcc-W Σ = 5 grain boundary (GB). The segregation energy is calculated to be −1.37 eV, indicating that He prefers to segregate in the W GB. The formation energy of He in the W GB is positive and thus He is quite hard to dissolve in the W GB, similar to its behavior in the bulk. Because of its closed-shell electronic structure, He is shown to preferably occupy either interstitial or substitutional site with larger space provided by the GB, changing the GB electronic structure. Moreover, segregation of He gives rise to the W GB expansion. These structure variations can have a large effect on the mechanical properties of the W GB.     

Grain boundary influence on displacement cascades in UO2: A molecular dynamics study

The influence of grain boundaries on the primary damage state created by a recoil nucleus in UO₂ matrix is studied here by molecular dynamics simulations. This study is divided in two steps: (1) the study of the structural properties of several symmetrical tilt boundaries for different misorientation angles ranging from 12.7° to 61.9°; and (2) the study of displacement cascades near these grain boundaries. For all the grain boundaries studied, the structure around the interface up to about 2 nm presents a perturbed but stable fluorite lattice. The type of defect at the interface depends directly on the value of the misorientation angles. For the small angles (12.7° and 16.3°) the interface defects correspond to edge dislocations. For higher misorientation angles, a gap of about 0.3 nm exists between the two halves of the bicrystal. This gap is composed of Schottky defects involving numerous vacancies along the interface. About 10 keV displacement cascades were initiated with an uranium projectile close to the interface. In all the cases, numerous point defects are created in the grain boundary core, and the mobility of these defects increases. However, cascade morphologies depend strongly on the grain boundary structure. For grain boundaries with edge dislocations, the evolution of the displacement cascades is similar to those carried out in monocrystals. On the other hand, cascades initiated in grain boundaries with vacancy layer defects present an asymmetry on the number of displaced atoms and the number of point defects created.     

The microstructure of neutron-irradiated Fe-Cr alloys: A small-angle neutron scattering study

The effect of Cr on the irradiation-induced microstructure of neutron-irradiated Fe-Cr alloys is not yet known in detail. Small-angle neutron scattering was applied in order to provide the characteristics of nm-sized defects averaged over macroscopic volumes. Results are reported for a set of Fe-Cr alloys of Cr levels of 2.5, 5, 9 and 12.5 at.%, irradiated at 300 °C up to neutron exposures of 0.6 and 1.5 dpa. We have found that the incoherent magnetic scattering of the unirradiated alloys exhibits a systematic variation with the Cr content and that there is an irradiation-induced increase of the coherent magnetic scattering for each of the irradiated conditions. The effect of Cr on size and type of irradiation-induced scatterers is discussed. For 12.5 at.%Cr, the scatterers are unambiguously identified as α′ particles. For 2.5 and 5 at.%Cr, the scatterers are tentatively interpreted as clusters enriched with alloying Cr and impurity C. For 9 at.%Cr, a mixture of both kinds of scatterers explains the experimental findings.     

Angular Distributions of Resonance Neutron Scattering from Fe-56

Chromium depletion near grain boundaries of austenitic stainless steel during irradiation was investigated. Specimens were kept at 1,473 K for 30 min, and were quenched into the water. Irradiations were done using 400 keV He⁺ ions at 573, 673 and 773 K up to 10dpa with a dose rate of 2.4×10^?4 dpa/s. After irradiation, the Cr concentration profile near the grain boundary was measured using an analytical electron microscope with a 1 nm beam diameter. At 573 K, Cr depletion is small, and its concentration at the grain boundary decreases to 15.5 mass% at 3 dpa from the initial concentration of 18.5 mass%. At 673 and 773 K, Cr concentration at the grain boundary rapidly decreases between 0 and 0.2dpa, and then gradually approaches a constant value, 7.0 mass% at 673 K and 5.0 mass% at 773 K. Two stages are found in radiation induced segregation (RIS) behavior, one stage in which Cr depletion and Ni enrichment balance and another in which Fe depletion and Ni enrichment balance.

These experimental results were compared with the calculations based on the vacancy-induced inverse Kirkendall effect. Predicted Cr segregation at 673 and 773 K above 3dpa agrees with the experimental results. But Cr depletions at low doses which were obtained in the experiments are much faster than calculated. At 573 K in the experiments, depletion is smaller than calculated up to 10dpa.     

Segregation of atoms in NixPdy alloys during ion irradiation

Sputtering of Ni₅Pd and NiPd₅ alloys by 10 keV Ar ions has been studied using the binary-collision simulation. Special attention was given to the angular distributions of sputtered atoms at the steady-state conditions. The results of simulations were compared with the experimental data published recently. For both targets, the concentrations of Ni and Pd atoms in the top monolayer were extracted from the experimental data. The results of simulations favor segregation of Pd in Ni₅Pd and segregation of Ni in NiPd₅. The total concentration of surface vacancies was found to be about 10-30%.     

Effect of Zirconium Addition to Austenitic Stainless Steels on Suppression of Radiation Induced Chromium Segregation at Grain Boundaries under Ion Irradiation

The effect of Zr addition to austenitic stainless steels on the suppression of radiation induced Cr segregation at grain boundaries under 400 keV He⁺ irradiation was studied. Type 316L stainless steel and steels with addition of 0.07, 0.21 or 0.41 mass% Zr were kept at 1,423K for 30 min, and then they were quenched into the water. Irradiation was done at 773K with the dose rate of 2.4×10^?4dpa/s. The total dose was 0.85 or 3.4dpa. After irradiation, profiles of Cr concentration across the grain boundaries were measured using an analytical electron microscope with 1 nm beam diameter. Concentration of Cr at the grain boundary is decreased by radiation induced segregation. However, it increased with the addition of Zr, and the Cr segregation is almost completely suppressed when Zr is added more than 0.21 mass%.

The effect of Zr addition on suppression of Cr segregation was analyzed focussing on the interaction between dissolved Zr atoms and point defects. The effect is based on vacancy trapping by the Zr atom, and the extent to which it suppresses Cr segregation can be empirically evaluated using a radiation induced segregation model by changing the effective vacancy migration energy.     

SANS investigation of a neutron-irradiated Fe-9 at%Cr alloy

Available experimental results indicate that the addition of Cr to Fe and steels significantly influences the response of Fe-Cr alloys and ferritic/martensitic high-Cr steels to neutron irradiation. A level of 9 at%Cr is of particular interest because this composition is close to the boundary of the Fe-Cr miscibility gap. Furthermore, it corresponds to the composition of several candidate steels for application in nuclear technology. However, experimental evidence has been incomplete so far. The reported study by means of small-angle neutron scattering is devoted to the effect of neutron irradiation at 300 °C up to fluences of 0.6 and 1.5 dpa on the microstructure of an Fe-9 at%Cr alloy. We have observed a pronounced irradiation-induced increase of scattering cross-sections for both magnetic and nuclear scattering. Bimodal size distributions of irradiation-induced defect-solute clusters have been reconstructed. The restrictions on the composition of these clusters have been discussed in terms of the scattering contrast. We have found that vacancy clusters and α′-particles alone cannot explain the full set of experimental findings. The remaining inconsistency can be solved by taking into account a contribution of impurity carbon.     

Generation of c-component edge dislocations in α-zirconium during neutron irradiation - An atomistic study

The nucleation and multiplication of c-component edge dislocation segments during neutron irradiation in zirconium and its alloys is known to have important consequences to their in-reactor deformation behavior. Although there are ample experimental observations showing the close correlation between the edge-type and the screw-type of c-dislocations, the relation between them is unclear. In this paper, we performed atomistic study of the interaction between a [0 0 0 1] screw dislocation and a vacancy cluster in the form of a platelet on the basal plane. The local minimum-energy configuration was obtained using the conjugate-gradient method, with boundary relaxation achieved via a modified Green’s function method. Under stress-free conditions, the vacancy clusters maintained their cavity nature. With a [0 0 0 1] screw dislocation in the close neighborhood, vacancy clusters containing more than 23 vacancies collapse into faulted vacancy loops. Interaction at even closer range leads to the disappearance of the vacancy cluster and the development of an edge component on the originally straight screw dislocation in the form of a helical line. The implications of these findings are discussed in relation to the experimentally observed behavior of growth acceleration in zirconium and its alloys.     

Self-diffusion near symmetrical tilt grain boundaries in UO2 matrix: A molecular dynamics simulation study

Molecular dynamics simulations have been carried out to study the influence of grain boundaries in stoichiometric UO₂ on uranium and oxygen self-diffusions over a large range of temperature varying from 300 K to 2100 K. The study was carried out on two symmetrical tilt grain boundaries, Σ5 and Σ41, which have respectively two different atomic structures. Firstly, the study of the temperature effect on the grain boundary core structure is presented. With the raise of temperature, the grain boundary core grows with an increase of disorder. Secondly, self-diffusion near both grain boundaries is studied. It has been found that grain boundaries accelerate the uranium and oxygen self-diffusion rates over several nanometres from the grain boundary interface. Uranium and oxygen self-diffusion are anisotropic, with a high acceleration along the grain boundary interface. Using the self-Van Hove correlation functions, hopping mechanisms were identified for Σ41 in all directions while for Σ5 hopping mechanism takes place along the grain boundary interface and random diffusion appears in the perpendicular direction of the grain boundary plane.     

The effect of vacancy created by ion irradiation on the ordering of FePt: A first-principle study

Ion irradiation has been used to promote ordering processes and to modify the magnetic properties of magnetic thin films. The major reason for ion irradiation reducing the ordering temperature is the introduction of a number of vacancies. The vacancy and its influence on the ordering temperature and magnetic properties in L1₀ ordered FePt are investigated by first-principle simulation. The vacancy formation energy for Fe and Pt in FePt alloy are 1.45 and 2.25 eV respectively. The calculated order-disorder transition temperature of Fe₅₀Pt₅₀ is 1680 K. The order-disorder transition temperatures for Fe vacancy and Pt vacancy models are about 50 K and 200 K lower than that of the stoichiometric Fe₅₀Pt₅₀ alloy respectively. The results suggested that the vacancy in FePt alloy favors the ordering process. The saturation magnetization of stoichiometric L1₀ FePt is 1070 emu/cc and these of Fe and Pt vacancy are 1027 and 1075 emu/cc, respectively.     

Fe-Cr-V ferritic steels for possible nuclear applications

Basic mechanical and metallurgical properties of specific ferritic Fe-Cr-V alloys and steels with 5, 10, and 15 wt.% vanadium were investigated. Vanadium is an effective carbide former and can also form a brittle sigma phase with chromium. Therefore, the microstructural investigations focused on the determination and analysis of possible precipitations. The present study showed that sigma phase precipitates increase significantly in alloys with 10 wt.% Cr and 10 wt.% V. The addition of carbon led to grain refinement due to the stabilizing effect of VC. In this way, precipitation hardening as well as fine grain strengthening could be quantified for this class of material. However, compared to typical martensitic steels, the strength of the considered ternary Fe-Cr-V alloys and steels is still lower.     

A systematic approach for the study of radiation-induced segregation/depletion at grain boundaries in steels

High Cr ferritic steels are candidate materials for structural applications in Gen-IV and fusion nuclear reactors. However, the relative contributions of irradiation conditions and materials microstructures on radiation-induced segregation or depletion of Cr at grain boundaries in ferritic steels are unclear. Here, the possibility of systematically analyzing the chemistry of the same grain boundary of known character during irradiation is demonstrated using a combination of electron back-scattered diffraction, atom-probe tomography and focused ion beam specimen preparation. This method provides a dynamic evolution of grain boundary chemistry as function of dose, spatial variations within the grain boundary plane, and quantification of minor solute elements such as carbon otherwise difficult to obtain experimentally.