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.     

Embrittlement of irradiated ferritic/martensitic steels in the absence of irradiation hardening

Irradiation damage caused by neutron irradiation below 425-450 °C of 9-12% Cr ferritic/martensitic steels produces microstructural defects that cause an increase in yield stress. This irradiation hardening causes embrittlement observed in a Charpy impact test as an increase in the ductile-brittle transition temperature. Little or no change in strength is observed in steels irradiated above 425-450 °C. Therefore, the general conclusion has been that no embrittlement occurs above these temperatures. In a recent study, significant embrittlement was observed in F82H steel irradiated at 500 °C to 5 and 20 dpa without any change in strength. Earlier studies on several conventional steels also showed embrittlement effects above the irradiation-hardening temperature regime. Indications are that this embrittlement is caused by irradiation-accelerated or irradiation-induced precipitation. Observations of embrittlement in the absence of irradiation hardening that were previously reported in the literature have been examined and analyzed with computational thermodynamics calculations to illuminate and understand the effect.     

Welding-induced microstructure in austenitic stainless steels before and after neutron irradiation

The effects of neutron irradiation on the microstructure of welded joints made of austenitic stainless steels have been investigated. The materials were welded AISI 304 and AISI 347, so-called test weld materials, and irradiated with neutrons at 300 °C to 0.3 and 1.0 dpa. In addition, an AISI 304 type from a decommissioned pressurised water reactor, so-called in-service material, which had accumulated a maximum dose of 0.35 dpa at about 300 °C, was investigated. The microstructure of heat-affected zones and base materials was analysed before and after irradiation, using transmission electron microscopy. Neutron diffraction was performed for internal stress measurements. It was found that the heat-affected zone contains, relative to the base material, a higher dislocation density, which relates well to a higher residual stress level and, after irradiation, a higher irradiation-induced defect density. In both materials, the irradiation-induced defects are of the same type, consisting in black dots and Frank dislocation loops. Careful analysis of the irradiation-induced defect contrast was performed and it is explained why no stacking fault tetrahedra could be identified.     

A SANS investigation of the irradiation-enhanced α–α′ phases separation in 7–12 Cr martensitic steels

Five reduced activation (RA) and four conventional martensitic steels, with chromium contents ranging from 7 to 12 wt%, were investigated by small angle neutron scattering (SANS) under magnetic field after neutron irradiation (0.7–2.9 dpa between 250 and 400 °C). It was shown that when the Cr content of the b.c.c. ferritic matrix is larger than a critical threshold value (∼7.2 at.% at 325 °C), the ferrite separates under neutron irradiation into two isomorphous phases, Fe-rich (α) and Cr-rich (α^′). The kinetics of phase separation are much faster than under thermal aging. The quantity of precipitated α^′ phase increases with the Cr content, the irradiation dose, and as the irradiation temperature is reduced. The influence of Ta and W added to the RA steels seems negligible. Cold-work pre-treatment increases slightly the coarsening of irradiation-induced precipitates in the 9Cr–1Mo (EM10) steel. In the case of the low Cr content F82H steel irradiated 2.9 dpa at 325 °C, where α^′ phase does not form, a small irradiation-induced SANS intensity is detected, which is probably due to point defect clusters. The α^′ precipitates contribute significantly to the irradiation-induced hardening of 9–12 wt% Cr content steels.     

Ab initio calculations of self-interstitial interaction and migration with solute atoms in bcc Fe

The embrittlement of pressure vessel steels under radiation has been long ago correlated with the presence of solute Cu. Indeed the atom probe and the small angle neutron scattering, principally, have revealed the formation of Cu clusters under neutron flux in reactor pressure vessel (RPV) steels and dilute FeCu alloys. Other solutes such as Ni, Mn and Si which are also found within the clusters, are now suspected to contribute to the embrittlement. The interactions of these solutes with radiation induced point defects need thus to be characterized properly in order to understand the elementary mechanisms behind the formation of these clusters. We have investigated by ab initio calculations based on the density functional theory the interactions of self-interstitials with solute atoms in dilute FeX alloys (X = Cu, Mn, Ni or Si). Different possible configurations of solute-dumbbell complexes have been studied. Their binding energies are discussed, as well as their relative stability. The migration of dumbbells with a solute atom in their vicinity was also investigated. All these results are compared to some experimental ones obtained on dilute FeX model alloys. Our results indicate that for Mn solute atoms, diffusion via an interstitial mechanism is very likely.     

Hardening and microstructural evolution in A533B steels under neutron irradiation and a direct comparison with electron irradiation

A533B steels irradiated at 290 °C up to 10 mdpa in the Kyoto University Reactor were examined by hardness, positron annihilation and atom probe measurements. Dose dependent irradiation hardening and formation of Cu-rich clusters were confirmed in medium Cu (0.12% and 0.16%Cu) steels whereas neither hardening nor cluster formation was detected in low Cu (0.03%Cu) steel. No microvoids were formed in any of the steels. Post-irradiation annealing in medium Cu steels revealed that the hardening recovery at temperatures above 350-400 °C could be attributed to compositional changes and dissociation of the Cu-rich clusters. Compared to electron irradiation at almost the same dose and dose rate, KUR irradiation caused almost the same hardening and produced Cu-rich clusters, more solute-enriched with larger size and lower density. Considering lower production of freely-migrating vacancies in neutron irradiation, the results suggested that cascades enhance the formation of Cu-rich clusters.     

Understanding silicon-rich phase precipitation under irradiation in austenitic stainless steels

Atom probe samples have been Fe⁺ ion irradiated at different doses (from 0.5 to 10 dpa) and different temperatures (between 300 and 400 °C) in order to understand the mechanism of formation, under irradiation, of Si-rich phases in austenitic stainless steels. Atom probe results show the presence of Si-enriched clusters which can also be enriched in Ni and depleted in Cr. Number densities of solute clusters can be linked to number densities of dislocation loops already observed by transmission electron microscopy in a previous work. This suggests that solute clusters are formed by heterogeneous precipitation on dislocation loops. Furthermore, the evolution of the composition of solute clusters as a function of the irradiation temperature is consistent with a radiation-induced mechanism. Results are also compared with previous results obtained after neutron irradiation at lower dose rate (in term of dpa s⁻¹). The comparison is, here again, consistent with the radiation-induced mechanism. Thus, Si-rich clusters may be formed by radiation-induced segregation to dislocation loops. Results also show that Si is probably dragged to sinks via the interstitial mechanism.     

Response of nanoclusters in a 9Cr ODS steel to 1 dpa, 525 °C proton irradiation

Ferritic-martensitic (F/M) alloys are expected to play an important role as cladding or structural components in Generation IV and other advanced nuclear systems operating in the temperature range 350-700 °C and to doses up to 200 displacements per atom (dpa). Oxide dispersion strengthened (ODS) F/M steels have been developed to operate at higher temperatures than traditional F/M steels. These steels contain nanometer-sized Y-Ti-O nanoclusters for additional strengthening. A proton irradiation to 1 dpa at 525 °C has been performed on a 9Cr ODS steel to determine the nanocluster stability at low dose. The evolution of the nanocluster population and the composition at the nanocluster-matrix interface were studied using electron microscopy and atom probe tomography. The data from this study are contrasted to those from a previous study on heavy-ion irradiated 9Cr ODS steel.     

Small-angle neutron scattering study on the effect of hydrogen in irradiated reactor pressure vessel steels

Hydrogen uptake can enhance the neutron embrittlement of reactor pressure vessel (RPV) steels. This suggests that irradiation defects act as hydrogen traps. The evidence of hydrogen trapping was investigated using the small-angle neutron scattering (SANS) method on four RPV steels. The samples were examined in the unirradiated and irradiated states and both in the as-received condition and after hydrogen charging. Despite the low bulk content of hydrogen achieved after charging with low current densities, an enrichment of hydrogen in small microstructural defects could be identified. Preferential traps were microstructural defects in the size range of ≈ > 10 nm in the unirradiated and irradiated samples. However, the results do not show any evidence for hydrogen trapping in irradiation defects.     

Recent applications of small-angle neutron scattering in the characterization of irradiated steels for nuclear technologies

Small-angle neutron scattering (SANS) is a powerful experimental tool to investigate the microstructural evolution under irradiation in steels for fission and future fusion reactor systems. We present recent SANS results concerning the modelling of helium bubble growth in F82H-mod. steel implanted with α-particles and the dose dependence of microstructural radiation damage in Eurofer-97 steel for fusion reactors irradiated at 250 °C. The discussion of these results is focussed on the quality of the metallurgical information obtained by such SANS measurements and consequently on their usefulness also for engineering and design purposes.     

Prediction of irradiation embrittlement of vanadium alloyed low nickel steel for future reactors

A comparison between pearlitic 2CrMoV steel (WWER-440) and 9% Cr based ferritic-martensitic steels (EUROFER 97 and LA12TaLC) is presented as regards irradiation induced ductile-brittle transition temperature shifts. For neutron doses of 1.5-2 dpa and irradiation temperatures around 300 °C the transition temperature shifts for WWER-440 steel and EUROFER 97 welds are comparable. In the temperature range 350-500 °C the radiation embrittlement levels of both steels are low. Moreover, post-irradiation annealing is proposed as a promising method to predict results of high temperature irradiation embrittlement from existing lower temperature irradiation embrittlement data.     

Dislocation loop evolution under ion irradiation in austenitic stainless steels

A solution annealed 304 and a cold worked 316 austenitic stainless steels were irradiated from 0.36 to 5 dpa at 350 °C using 160 keV Fe ions. Irradiated microstructures were characterized by transmission electron microscopy (TEM). Observations after irradiation revealed the presence of a high number density of Frank loops. Size and number density of Frank loops have been measured. Results are in good agreement with those observed in the literature and show that ion irradiation is able to simulate dislocation loop microstructure obtained after neutron irradiation.Experimental results and data from literature were compared with predictions from the cluster dynamic model, MFVIC (Mean Field Vacancy and Interstitial Clustering). It is able to reproduce dislocation loop population for neutron irradiation. Effects of dose rate and temperature on the loop number density are simulated by the model. Calculations for ion irradiations show that simulation results are consistent with experimental observations. However, results also show the model limitations due to the lack of accurate parameters.     

Atom probe characterization of nano-scaled features in irradiated ODS Eurofer steel

Our previous investigations of unirradiated ODS Eurofer by tomographic atom probe (TAP) revealed numerous nano-scaled features (nanoclusters) enriched in vanadium, yttrium and oxygen. In this work the effect of neutron irradiation on nanostructure behaviour of ODS Eurofer (9%-CrWVTa) was investigated. The irradiation was performed in the research reactor BOR-60 (Dimitrovgrad, Russia) where materials were irradiated at 330 °С to 32 dpa. TAP studies were performed on the needles prepared from parts of broken Charpy specimens. For all specimens except one, which was tested at 500 °C, the Charpy tests were performed at temperatures not exceeding the irradiation temperature. A high number density 2-4 × 10²⁴ m⁻³ of ultra fine 1-3 nm diameter nanoclusters enriched in yttrium, oxygen, manganese and chromium was observed in the irradiated state. The composition of detected clusters differs from that for unirradiated ODS Eurofer. It was observed in this work that after neutron irradiation vanadium atoms had left the clusters, moving from the core into solid solution. The concentrations of yttrium and oxygen in the matrix, as it was detected, increase several times under irradiation. In the samples tested at 500 °C both the number density of clusters and the yttrium concentration in the matrix decrease by a factor of two.     

Behavior of ferritic/martensitic steels after n-irradiation at 200 and 300 °C

High chromium ferritic/martensitic (F/M) steels are considered as the most promising structural materials for accelerator driven systems (ADS). One drawback that needs to be quantified is the significant hardening and embrittlement caused by neutron irradiation at low temperatures with production of spallation elements. In this paper irradiation effects on the mechanical properties of F/M steels have been studied and comparisons are provided between two ferritic/martensitic steels, namely T91 and EUROFER97. Both materials have been irradiated in the BR2 reactor of SCK-CEN/Mol at 300 °C up to doses ranging from 0.06 to 1.5 dpa. Tensile tests results obtained between −160 °C and 300 °C clearly show irradiation hardening (increase of yield and ultimate tensile strengths), as well as reduction of uniform and total elongation. Irradiation effects for EUROFER97 starting from 0.6 dpa are more pronounced compared to T91, showing a significant decrease in work hardening. The results are compared to our latest data that were obtained within a previous program (SPIRE), where T91 had also been irradiated in BR2 at 200 °C (up to 2.6 dpa), and tested between −170 °C and 300 °C. Irradiation effects at lower irradiation temperatures are more significant.     

Materials researches at the Paul Scherrer Institute for developing high power spallation targets

For the R&D of high power spallation targets, one of the key issues is understanding the behavior of structural materials in the severe irradiation environments in spallation targets. At PSI, several experiments have been conducted using the targets of the Swiss spallation neutron source (SINQ) for studying radiation damage effects induced by high energy protons and spallation neutrons. As well, experiments have been performed to investigate liquid lead-bismuth eutectic (LBE) corrosion and embrittlement effects on T91 steel under irradiation with 72 MeV protons. In this paper, an overview will be given showing a selection of results from these experiments, which include the mechanical properties and microstructure of ferritic/maretensitic (FM) steels (T91, F82H, Optifer etc.) and austenitic steels (EC316LN, SS 316L, JPCA etc.) irradiated to doses higher than ever attained by irradiation in a spallation environment, and the behaviors of T91 irradiated with 72 MeV protons in contact with flowing LBE.     

Monitoring the embrittlement of reactor pressure vessel steels by using the Seebeck coefficient

The degree of embrittlement of the reactor pressure vessel (RPV) limits the lifetime of nuclear power plants. Therefore, neutron irradiation-induced embrittlement of RPV steels demands accurate monitoring. Current federal legislation requires a surveillance program in which specimens are placed inside the RPV for several years before their fracture toughness is determined by destructive Charpy impact testing. Measuring the changes in the thermoelectric properties of the material due to irradiation, is an alternative and non-destructive method for the diagnostics of material embrittlement. In this paper, the measurement of the Seebeck coefficient () of several Charpy specimens, made from two different grades of 22 NiMoCr 37 low-alloy steels, irradiated by neutrons with energies greater than 1 MeV, and fluencies ranging from 0 up to 4.5 × 10¹⁹ neutrons per cm², are presented. Within this range, it was observed that increased by ≈500 nV/°C and a linear dependency was noted between and the temperature shift ΔT_41 J of the Charpy energy vs. temperature curve, which is a measure for the embrittlement. We conclude that the change of the Seebeck coefficient has the potential for non-destructive monitoring of the neutron embrittlement of RPV steels if very precise measurements of the Seebeck coefficient are possible.     

Radiation-damage in molybdenum-rhenium alloys for space reactor applications

Various Mo-Re alloys are attractive candidates for use as fuel cladding and core structural materials in spacecraft reactor applications. Molybdenum alloys with rhenium contents of 41-47.5% (wt%), in particular, have good creep resistance and ductility in both base metal and weldments. However, irradiation-induced changes such as transmutation and radiation-induced segregation could lead to precipitation and, ultimately, radiation-induced embrittlement. The objective of this work is to evaluate the performance of Mo-41Re and Mo-47.5Re after irradiation at space reactor relevant temperatures. Tensile specimens of Mo-41Re and Mo-47.5Re alloys were irradiated to ∼0.7 displacements per atom (dpa) at 1073, 1223, and 1373 K and ∼1.4 dpa at 1073 K in the High Flux Isotope Reactor at Oak Ridge National Laboratory. Following irradiation, the specimens were strained to failure at a rate of 1 × 10⁻³ s⁻¹ in vacuum at the irradiation temperature. In addition, unirradiated specimens and specimens aged for 1100 h at each irradiation temperature were also tested. Fracture mode of the tensile specimens was determined. The tensile tests and fractography showed severe embrittlement and IG failure with increasing temperatures above 1100 K, even at the lowest fluence. This high temperature embrittlement is likely the result of irradiation-induced changes such as transmutation and radiation-induced segregation. These factors could lead to precipitation and, ultimately, radiation-induced embrittlement. The objective of this work is to examine the irradiation-induced degradation for these Mo-Re alloys under neutron irradiation.     

Stability of nanometer-sized oxide clusters in mechanically-alloyed steel under ion-induced displacement cascade damage conditions

Oxide dispersion strengthened ferritic steels are being considered for a number of advanced nuclear reactor applications because of their high strength and potential for high temperature application. Since these properties are attributed to the presence of a high density of very small (nanometer-sized) oxide clusters, there is interest in examining the radiation stability of such clusters. A novel experiment has been carried out to examine oxide nanocluster stability in a mechanically alloyed, oxide dispersion strengthened ferritic steel designated 12YWT. Pre-polished specimens were ion irradiated and the resulting microstructure was examined by atom probe tomography. After ion irradiation to ∼0.7 dpa with 150 keV Fe ions at 300 °C, a high number density of ∼4 nm-diameter nanoclusters was observed in the ferritic matrix. The nanoclusters are enriched in yttrium, titanium and oxygen, depleted in tungsten and chromium, and have a stoichiometry close to (Ti + Y):O. A similar cluster population was observed in the unirradiated materials, indicating that the ultrafine oxide nanoclusters are resistant to coarsening and dissolution under displacement cascade damage for the ion irradiation conditions used.     

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.     

Effects of solute atoms on evolution of vacancy defects in electron-irradiated Fe-Cr-based alloys

The evolution of vacancy-type defects in Fe-Cr alloys (13-16 at.% Cr) undoped and doped with C, N, Au, or Sb and in conventional ferritic-martensitic steel (∼13% Cr) has been investigated using positron annihilation spectroscopy under electron irradiation at room temperature and subsequent stepwise annealing. Small vacancy clusters are formed in the undoped Fe-16Cr alloy, which anneal out between 320 and 550 K. It is shown that oversized substitutional solute atoms (Sb, Au) in the Fe-Cr alloy interact with vacancies and form complexes, which are stable up to 600 and 420 K, respectively. It is found that the accumulation of vacancy defects considerably increases in the alloys and the steel with an enhanced content of interstitial impurities. It is shown that this effect is related to the formation of vacancy-carbon complexes. It is known that chromium in iron decreases the diffusion mobility of carbon. Therefore, the structure of vacancy-carbon complexes and the kinetics of their annealing in Fe-Cr alloys differ from those in the Fe-C system.