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.     

Ab initio formation energies of Fe-Cr alloys

We have calculated ab initio lattice parameters, formation energies, bulk moduli and magnetic moments of Fe-Cr alloys. The results agree well with available experimental data. In addition to body centered cubic (bcc) alloys, which are representative of ferritic steels used in fast neutron reactors, face centered cubic (fcc) and hexagonal close packed (hcp) phases were considered in order to complete a theoretical database of thermodynamic properties. Calculations were done for the ferromagnetic phase, as well as for a phase with local moment disorder, simulating the magnetic structure at high temperatures. For the latter case, the formation energy of the alloy is strictly positive smooth function of chromium concentration, in agreement with experiments performed at high temperature. In the ferromagnetic case, a negative mixing enthalpy is found for chromium concentrations below 6%. Our observation is consistent with the experimentally observed inversion of the ordering trend, as well as with formation of the chromium rich α^′ phase at Cr-concentrations above 9%, occurring at T<900 K.     

Elevated-temperature heavy-ion damage in (Co0.78Fe0.22)3V ordered alloy

Irradiation-induced changes in microstructure were studied in a (Co_0.78Fe_0.22)₃V long-range-ordered alloy after bombardment with 4 MeV nickel ions at temperatures from 843 to 1023 K and damage levels from 10 to 70 dpa. Small cavities and Frank loops were developed, and particles of pre-existing VC phase were redistributed by the irradiation. Long-range order was retained under all irradiation conditions, but domain size was effectively reduced by the Frank loops that served as antiphase boundaries. Simultaneous injection of helium at a rate of 3 at.ppm/dpa and deuterium at a rate of 12 at.ppm/dpa during irradiation produced gas bubbles in the grain boundaries, increased the dislocation and cavity densities, reduced the mean cavity size, but had little effect on the ordered domain size or the redistribution of VC particles. Even with injected gases, swelling remained below 0.25%. It is deduced that swelling did not reach the rapid bias-driven state because the critical cavity size was large.     

Irradiation induced dislocation loop and its influence on the hardening behavior of Fe-Cr alloys by an Fe ion irradiation

Nano indentation analysis and transmission electron microscopy observation were performed to investigate a microstructural evolution and its influence on the hardening behavior in Fe-Cr alloys after an irradiation with 8 MeV Fe⁴⁺ ions at room temperature. Nano indentation analysis shows that an irradiation induced hardening is generated more considerably in the Fe-15Cr alloy than in the Fe-5Cr alloy by the ion irradiation. TEM observation reveals a significant population of the a₀<1 0 0> dislocation loops in the Fe-15Cr alloy and an agglomeration of the 1/2a₀<1 1 1> dislocation loops in the Fe-5Cr alloy. The results indicate that the a₀<1 0 0> dislocation loops will act as stronger obstacles to a dislocation motion than 1/2a₀<1 1 1> dislocation loops.     

Confinement of motion of interstitial clusters and dislocation loops in BCC Fe-Cr alloys

This work is devoted to the study of the effect of Cr solutes on the mobility of self interstitial atom (SIA) clusters and small interstitial dislocation loops (of size up to a few nanometers) in concentrated Fe-Cr alloys. Atomistic simulations have been performed to characterize the variation of the free energy of interstitial loops in the Fe-15Cr alloy using the experimentally determined profile of Cr distribution along the path of a loop. It is shown that the presence of randomly distributed Cr in Fe leads to the creation of local trapping configurations for small SIA clusters. The strength (trapping energy) and density of these configurations depend on the Cr content. On the contrary, large SIA clusters (which can be described as 1/2〈1 1 1〉 dislocation loops) are strongly affected by the presence Cr-Cr pairs and larger Cr clusters, which act as barriers to their motion.     

Mechanical properties changes of Fe-Cr alloys by fast neutron irradiation

The changes in the mechanical properties of Fe-Cr ferritic, single α-phase alloys, containing 0–15 wt% Cr, by fast neutron irradiation were studied by means of tensile tests in the temperature range between 77 and 283 K, mainly with regard to the effects of the Cr content. The temperature dependence of the yield stress decreased with increasing Cr content. The irradiation raised the thermal activated component of the yield stress in low Cr alloys severely, but not in high Cr alloys. The fracture stress in the brittle fracture mode increased with Cr content up to 10 wt%, beyond which it decreased. The frequency of twinning deformation showed an opposite tendency to the fracture stress with or without irradiation. The transition temperature of the unstable plastic flow, which occurred intensively after irradiation, decreased with increasing Cr content. The last condition mainly governed the irradiation embrittlement of Fe-Cr alloys determined from absorbed energy-temperature curves in tensile testing. Thus, high Cr alloying can make the steel more resistant to irradiation embrittlement.     

Object Kinetic Monte Carlo calculations of irradiated Fe-Cr dilute alloys: The effect of the interaction radius between substitutional Cr and self-interstitial Fe

Object Kinetic Monte Carlo models allow for the study of the evolution of the damage created by irradiation to time scales that are comparable to those achieved experimentally. Therefore, the essential Object Kinetic Monte Carlo parameters can be validated through comparison with experiments. However, this validation is not trivial since a large number of parameters is necessary, including migration energies of point defects and their clusters, binding energies of point defects in clusters, as well as the interaction radii. This is particularly cumbersome when describing an alloy, such as the Fe-Cr system, which is of interest for fusion energy applications. In this work we describe an Object Kinetic Monte Carlo model for Fe-Cr alloys in the dilute limit. The parameters used in the model come either from density functional theory calculations or from empirical interatomic potentials. This model is used to reproduce isochronal resistivity recovery experiments of electron irradiated dilute Fe-Cr alloys performed by Abe and Kuramoto. The comparison between the calculated results and the experiments reveal that an important parameter is the capture radius between substitutional Cr and self-interstitial Fe atoms. A parametric study is presented on the effect of the capture radius on the simulated recovery curves.     

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.     

Effect of Cr on the mechanical properties and microstructure of Fe-Cr model alloys after n-irradiation

High-chromium ferritic-martensitic steels are candidate structural materials for high-temperature applications in fusion reactors and accelerator driven systems (ADS). Cr concentration has been shown to be a key parameter which needs to be optimized in order to guarantee the best corrosion and swelling resistance, together with the minimum embrittlement. The behavior of Fe-Cr model alloys with different Cr concentrations (0, 2.5, 5, 9 and 12 wt%Cr) has been studied. Tensile tests have been performed in order to characterize the flow properties in the temperature range from −160 °C to 300 °C. The trend of the yield strength with temperature shows that the strain hardening is the same for all materials at low temperatures, even though they have different microstructures. The same materials have been neutron-irradiated at 300 °C in the BR2 reactor of SCK·CEN, up to three different doses (0.06, 0.6 and 1.5 dpa). The results obtained so far indicate that even at these low doses, the Cr content affects the hardening behavior of Fe-Cr binary alloys. Using the Orowan mechanism, the TEM observed microstructure provides an explanation of the obtained hardening but only at the very low dose, 0.06 dpa. At higher doses, other hardening mechanisms are needed.     

Radiation damage in non-fissile metallic alloys

A general survey is presented of radiation-induced displacement damage in non-fissile metallic alloys. The importance of the spatial arrangement of the vacancies and interstitials so produced is highlighted, especially as a guide to formulating an appropriate gauge for the various radiation-induced phenomena considered—i.e. hardening, embrittlement, growth, creep, swelling and fracture. The present level of theoretical understanding and the technological import of these phenomena are also assessed.     

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.     

Ion beam induced damage in InP during heavy-ion elastic recoil detection analysis

Heavy-Ion Elastic Recoil Detection Analysis (HIERDA) is an analytical technique which has undergone rapid development in the past few years with the availability of high-energy Tandem accelerators for materials science applications. HIERDA has found application in the study of various semiconductor systems, particularly III–V compounds. The technique employs a high-energy heavy-ion analysing beam to knock constituent nuclei from the target material and a time of flight and energy (ToF-E) detector system to extract mass and depth of origin information from these recoiling nuclei. Present work examines the sample damage produced in InP under typical analysis conditions. The depth distribution of damage induced by an ¹²⁷I analysing beam of varying energy (54–98 MeV) and dose (10¹³−2 × 10¹⁴ ions/cm²) in InP has been examined using RBS channelling, and cross-sectional TEM.     

Effect of Fe content on positron annihilation in neutron irradiated VFe alloys

Neutron irradiated VxFe alloys (with x from 0 to 5 at.%) have been studied by the conventional positron annihilation technique. A remarkable narrowing of angular correlation of annihilation radiation (ACAR) curves was observed for all alloys investigated. A specific feature of ACAR curves in pure vanadium is the presence of a narrow component attributed usually to the positronium (Ps) formation in voids, with inner surfaces covered by gaseous impurities such as oxygen. Significant changes in the ACAR curve component intensities with increase of iron content has been observed. At Fe concentration of about 1 at.% the narrow component disappears completely and the intensity of the middle one decreases significantly. It was concluded that the increase of Fe concentration in VFe alloys suppresses the void surface contamination by oxygen atoms and changes the positron work function from bulk materials into voids. Such behavior of the ACAR curve component intensities can be explained in terms of radiation-induced segregation of iron atoms at point defect sinks.     

Isochronal recovery and damage rate measurements in dilute NiSi alloys

The interaction of Ni-selfinterstitials with Si atoms in dilute NiSi alloys has been investigated by residual resistivity recovery after 5 K low dose electron irradiation and by damage rate measurements in the temperature range of stage II. In stage II the recovery spectra not only show the well known 250 K recovery stage but also a strong recovery peak at 105 K. After annealing to temperatures above 400 K a smaller residual resistivity is found as before irradiation, indicating a Si segregation during the recovery. The damage rates at 86 K show that single Si atoms act as traps for migrating Ni-selfinterstitials. At 105 K the initial damage rate becomes strongly dependent on the Si concentration but trapping at Si is still observed. This behavior as well as the dependence of the recovery from the Si and the defect concentration has been explained by the migration of the Ni-selfinterstitial-Si complexes at 105 K combined with a strong absorption of these migrating complexes at sinks.     

The improvement of irradiation-enhanced copper embrittlement in FeCu alloys

The effect of alloying elements on neutron irradiated FeCu alloys has been investigated in order to obtain the fundamental information on the irradiation-enhanced copper embrittlement for power reactor vessel steels. The mechanism of copper-induced irradiation embrittlement in the copper-containing iron alloys was proved to be due to both the interaction of copper atoms with irradiation-produced complex defects within grains, and the preferred grain boundary segregation of copper atoms existing near grain boundaries. The former effect causes the increase of yield strength, and the latter results in the ductility loss and grain boundary crackings. The addition of titanium up to 0.4 wt% to the Fe-0.1 wt% Cu alloy was found to be extremely effective in the improvement of both the irradiation-induced ductility loss and strength. Aluminum and silicon were not as effective as titanium.