Creep Properties and Related Microstructural Changes of a Lamellar Titanium Aluminide Alloy before and after Helium-Implantation

The creep properties of a cast gamma-alpha-2 lamellar titanium aluminide of type Ti-46Al-2W-0.5Si (at. %) containing beta-particles was investigated using both helium-implanted and non-implanted material. Creep tests were performed in vacuum under constant stress using small miniaturized dog-bone shaped samples with gauge dimensions of 10 × 2 × 0.2 mm3. Creep temperatures ranged from 700 °C to 1000 °C and applied stresses ranged from 75 MPa to 400 MPa. Helium implantation was carried out with 0-24 MeV helium ions homogeneously implanting the miniaturized samples from 6.3 to 1333 appm He, at temperatures of 150 °C and 630 to 1000 °C. Non-implanted samples showed very good creep resistance up to temperatures of roughly 850 °C; at temperatures above 900 °C a decrease in creep resistance was observed. For temperatures up to 900 °C the activation energy of creep was determined to 405 kJ/mol and the stress exponent of 7.2. Size and geometry effects of miniaturized samples were observed in form of large scatter in creep strain rates and lower strains at fracture. Electron microscopy showed that the initial microstructure was unstable and that the alloy had a tendency to decompose the alpha-2 phase and form beta-gamma structures, loss of initial microstructure was accelerated with increasing temperature. The creep deformation mechanism was determined to be a combination of stress-induced phase change and dislocation creep. A gradual change in fracture behavior was observed, at lower temperatures fracture was brittle with little necking, at the highest temperature the fracture was completely ductile. Material crept and implanted at 700 and 800 °C exhibited embrittlement at helium contents above 10 appm, characterized by a strong loss of ductility and creep lifetime. Samples crept and implanted at 900 °C experienced embrittlement for all tested helium contents. Investigations with transmission electron microscopy revealed a substantial clustering and growth of helium bubbles at interfaces in the alpha-2 phase. An average bubble size for embrittlement between 5.3 and 6.7 nm was determined. The behavior and evolution of helium bubbles were studied with post implantation annealing experiments. Samples were implanted at 150, 630, 800 and 1000 °C and subsequently annealed at temperatures from 600 to 900 °C. Material implanted at 150 °C showed no bubbles before annealing, only after post-implantation annealing was widespread clustering observed. Material implanted at temperatures of 630 °C or higher, exhibited bubbles before post-implantation annealing. Depending on the implantation temperature, two different branches were observed in the size and distribution of helium bubbles even after post-implantation annealing. Samples implanted at 150 or 630 °C and then annealed at higher temperatures had high bubble density with small bubbles found in the entire microstructure. Samples implanted at 800 or 1000 °C had low bubble densities and the bubbles were only observed at alpha-2/beta and alpha-2/gamma interfaces, with sizes larger than the average bubble radius for embrittlement. Based on the post-implantation annealing experiments, a temperature limit for the onset of helium embrittlement between 630 and 700 °C is predicted. The helium embrittlement, the loss of initial microstructure and decreased creep resistance at temperatures above 900 °C suggests a limited applicability of lamellar TiAl in advanced nuclear reactor environments.

On the relationship between the microstructure and the mechanical properties of an ODS ferritic/martensitic steel

Amuthan Ramar

In the quest of materials for the first wall of the future fusion reactor, it has been shown that oxide dispersion strengthened (ODS) ferritic / martensitic (F/M) steels appear to be promising candidates. The inherent good mechanical properties supported by a good thermal conductivity, swelling resistance and low radiation damage accumulation of the base material, such as EUROFER97, are further enhanced by the presence of a fine dispersion of nanometric oxide particles. They would allow in principle for a higher operating temperature of the fusion reactor, above 550°C, which improves its thermal efficiency. In effect, their strength remains higher than the base material at high temperatures. There are however limitations to their application in the reactor, which relate to the intimate link between the microstructure and the mechanical properties. These limitations, such as an increased ductile to brittle transition temperature above room temperature relative to the base material, come directly from the role played by the oxide particles but also from their indirect impact on the overall microstructure, stemming from the procedure of elaboration of these materials. In addition, there is a lack of knowledge on their response to irradiation, which, even though it appears to be promising with a reduced irradiation induced hardening, is not very well understood. In this work we have developed and evaluated ferritic/martensitic steel, EUROFER97, strengthened by a dispersion of yttria particles. (1) The powder metallurgical process used to obtain it was optimized, (2) the basic mechanisms of hardening were identified and (3) the impact of irradiation on its microstructure and mechanical properties were investigated. In order to understand the behaviour of this class of material other oxide dispersion strengthened (ODS) ferritic/martensitic steels based on EUROFER97 and model alloys were also investigated. For the elaboration of the ODS steel the powders of atomized EUROFER97 and yttria are ball milled in a planetary ball mill and the material is compacted by hot isostatic pressing (HIP). The mechanical properties are evaluated by uniaxial tensile test, Charpy test and micro-indentation while the microstructure is investigated by optical microscopy, scanning electron microscopy and transmission electron microscopy. The focused ion beam (FIB) technique is also used. In the study of the ball milling it appears that the optimal time is 20 hours, as the X-ray peak broadening, crystallite size and hardness of the particles remain unchanged at longer times. Yttria nanoparticles clusters are no more detectable by X-ray after 2 hrs of milling, which is attributed here to their incorporation in the matrix, as opposed to dissolution in the matrix as is often suggested in the open literature. HR-TEM performed on the yttria particles before and after incorporation in the steel by ball milling shows that the yttria maintain its bcc equilibrium structure throughout the process. This could suggest that they are not dissolved during ball milling. We have clarified this point by a cross sectional examination of the EUROFER97 ball milled particles with 0.3wt% of yttria in an FIB. By X-ray EDS analysis it shows nanometric regions reach in yttrium and oxygen, which are the yttria particles embedded in the EUROFER97 particles, proving that they are not dissolved by ball milling. The mechanism at the origin of the alloy compaction during HIP was elaborated. The alloy compaction was done through HIPing at 1150°C for 2.5 hrs at a pressure of 190 MPa. This condition is shown to provide the highest bulk density, to a value close to 99.9%. In the optimization process, it is found that temperature plays a larger role than pressure for alloy compaction. The compaction appears to be mainly due to diffusion. Also, it appears that the ball milled particle size plays a major role in the alloy compaction. It is shown that smaller particles size induces better compaction, provided that the densification proceeds through diffusion flow. This further suggests to terminate the ball milling process by 20 hrs as the particles size reaches its minimum. The impact of the ball milling process on the mechanical properties was evaluated. EUROFER97 with and without ball milling, with and without oxide addition presents an average density around 99.8%. In the as received state, casted EUROFER97, atomized EUROFER97, ball milled EUROFER97 and ODS Yttria present microstructure morphology typical of tempered martensite, whereas ODS Ti presents a nano sized grain microstructure with an average grain size of 200 nm. Ball milling and HIPping does have a slight influence up to 400°C, due to the higher dislocation density induced by ball milling relative to casted EUROFER97. It becomes negligible above 400°C, since both casted and ball milled EUROFER97 present similar hardness and microstructure. In ODS Yttria, yttria presents an average particle size of 25 nm. In ODS Ti, Y-Ti-O presents an average particle size of 7 nm. In the following the identification of the hardening mechanisms in ODS F/M steels. A new model was developed to quantify the collective hardening due to different obstacles, assuming that they provide independent hardening mechanisms. The model helped identifying the dominating source of hardening in ODS F/M steels. The base of the model is the dispersion barrier strengthening model, which is given by Δσsparticle or defects = Mαμb(Nd)1/2 for the hardening due small defects or oxide particles and is given by Δσd-d = Mαμb(ρ)1/2 for the hardening due to dislocations. In this approximation, the total hardening is then given by Δσtotal = {(Δσd-d)2 + (Δσparticle or defects)2}1/2 . The model is used to explain the hardening due to oxides, dislocations and irradiation induced defects. It appears that dispersed oxides is dominating hardness up to 800°C, while following a heat treatment at 1000°C it is the dislocations that then dominate the hardness. Deformation mechanisms were investigated using differential strain rate experiments. The activation parameters, i.e. the activation volume and activation energy, were calculated. It appears that there is a decrease in the activation volume from room temperature to 400°C, which is less than 3%. At about 500°C there is a sudden drop in the activation volume that reaches a minimum at 600°C. With further increase in temperature, there is a significant raise in the activation volume. This abrupt decrease between 400°C and 600°C hints at a change in dislocation-rate controlling mechanism, marking two regimes, a low temperature one and a high temperature one. In situ TEM observations show that at low temperature the deformation is mainly due to the formation of the Orowan loops around the particle. At high temperature the activation energy is about 3.4 eV for ODS, whereas it is 1.3 eV for EUROFER97. The calculated activation energy corresponds well to the one of dislocation climb mechanism. This indicates that at high temperature the yttria particles could be overcome by climb, while at low temperature they are overcome by the Orowan mechanism. The peculiar response of the yield stress to irradiation for this class of material was rationalized using the total hardening model we developed. At a dose of 2 dpa, irradiation induced defects are higher in number density than the dispersed yttria particles, which although they are weaker than the yttria particles they can explain the significant hardening observed in the material. Conversely, the shallow impact of irradiation on ODS F/M steels at low doses, relative to base F/M steels, is explained by the fact that the amount of dispersoid is found to be higher than the density of irradiation induced defects.

EPFL2008

Helium embrittlement of a lamellar titanium aluminide

Per Magnusson, Philippe Spätig

Embrittlement by helium was investigated in a lamellar TiAl alloy under two conditions: Specimens were implanted to various amounts of helium up to 762 appm at temperatures from 630 °C to 1000 °C and some of them subsequently creep-tested at the same temperature under stresses from 150 to 300 MPa. The microstructure and fracture surfaces of creep-deformed and non-creep-deformed specimens were then studied by transmission electron microscopy (TEM) and by scanning electron microscopy (SEM), respectively. Specimens were implanted to various amounts of helium at a low temperature (150 °C) and post-implantation annealed at elevated temperatures for TEM studies. Embrittlement was revealed by reduction in time- and strain-to-rupture and by a transition in fracture surface from ductile to an inter-lamellar appearance. Embrittlement occurred above a critical He concentration, which decreased from about 10 appm at 700 °C to below 6 appm at 900 °C. TEM showed that embrittlement could be associated to reaching a critical bubble diameter of about 5 nm. Bubble diameters increased with increasing temperature ranging in high-temperature implanted specimens from about 3 nm (630 °C) to 20 nm (1000 °C) and in post-implantation annealed ones from 1.2 nm (600 °C) to 2.2 nm (900 °C), respectively. With increasing temperature, the bubble distribution grew less homogenous with a lower density of larger bubbles situated preferentially at interfaces and sinks. This was ascribed to a change in bubble nucleation mode from homogeneous di-atomic nucleation at lower temperatures to multi-atomic nucleation at sinks at higher temperature. © 2012 Elsevier B.V. All rights reserved.

Elsevier2013

Embrittlement induced by the synergistic effects of radiation damage and helium in structural steels

Kun Wang

Ferritic/martensitic steels are candidate materials for fusion reactor structural components, liquid metal containers of spallation neutron sources, and accelerator driven systems, with good radiation resistance and thermo-mechanical properties. However, embrittlement resulting from the combined effects of radiation induced displacement damage (measured in dpa = displacement per atom) and transmutation products, especially helium gas, is one of the key issues. Four different steels were selected for mechanical and microstructural studies to understand the mechanisms of embrittlement induced by the combined effects of displacement damage and helium after irradiation in SINQ, the Swiss spallation source. The irradiations conditions were in the range: 10.7 ¿ 20.4 dpa with 850-1750 appm He at 160-300 °C. The evolution of the mechanical properties after irradiation was investigated by tensile and hardness tests. Radiation-induced defect clusters and helium bubbles were quantified by transmission electron microscopy (TEM). Emphasis was put on the deformation mechanisms under the different observed fracture mode, (ductile, quasi-cleavage and intergranular), whose occurrence depends on the irradiation conditions (dpa, He content and irradiation temperature). The tensile stress-strain curves and the scanning electron microscopy images of fracture surfaces showed distinct fracture mechanisms under different irradiation and test conditions. The tensile tests showed a yielding stress increase and loss of ductility of irradiated specimens. Hardness was measured on the specimens before tensile testing. The hardness results demonstrated an increasing trend with irradiation dose and helium content. TEM observations were done for all irradiated fractured specimens. Small defect clusters were observed in the 12.3 dpa specimen, but large defect clusters with loop-shape were very few. In the specimens of 17.2, 17.7 and 20.4 dpa, many large dislocation loops were detected besides small clusters. In addition, helium bubbles were observed in all specimens. The average size of defect clusters increased from 4.2 nm to 11.8 nm with dose increasing from 12.3 dpa to 20.4 dpa, whereas the number density did not change significantly. Meanwhile, the average size of visible helium bubbles increased from 1.03 nm to 1.93 nm. The microstructures in deformed area of irradiated specimens were observed and defect free channels with {110} and {112} slip planes were found in some specimens, indicating plastic flow localization. The average width of the channels is about 100 nm. Regarding the brittle samples, the TEM-lamella were extracted directly below intergranular fracture surfaces or cleavage surfaces by focused ion beam. Strikingly, deformation twinning was observed as the main feature in three irradiated specimens at high dose. Only twins with {112} planes were observed in all of these samples. The average thickness of twins is about 34 nm. Twins started from a fracture surface became gradually thinner with distance away from the fracture surface and stopped in the matrix finally. Features such as twin-precipitates interaction, twin-grain boundary and/or lath boundary interaction were observed. Twinning bands were seen to be arrested by grain boundaries or large precipitates, but could penetrate martensitic lath boundaries. Unlike the case of defect free channels, inside twins small defect-clusters, dislocation loops and dense small helium bubbles were observed.

EPFL2015