Acier | EPFL Graph Search

Un acier est un alliage métallique constitué principalement de fer et de carbone. Il se distingue des fontes et des ferroalliages par sa teneur en carbone comprise entre 0,02 % et 2 % en masse. C’est essentiellement cette teneur en carbone qui confère à l'acier ses propriétés. Histoire Histoire de la production de l'acier L’Âge du fer se caractérise par l’adaptation du bas fourneau à la réduction du fer. Ce bas fourneau produit une loupe, un mélange hétérogène de fer, d’acier et de laitier, dont les meilleurs morceaux doivent être sélectionnés, puis cinglés pour en chasser le laitier. En poussant le vent, on attise la combustion et la température de fusion du métal est atteinte. On extrait le métal par vidange du creuset : c’est la production au haut fourneau. On obtient alors de la fonte, le fer liquide se chargeant de carbone au contact du charbon de bois. En effet, deux phénomènes complémentaires se déroulent dans le creuset du haut fourneau : le fer se charge de ca

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

Liquid lead-bismuth embrittlement effects on unirradiated and irradiated ferritic/martensitic steels for nuclear applications

Bin Long

In liquid metal spallation targets such as the MEGAPIE (the megawatt pilot experiment) target and the future ADS (accelerator driven system) spallation targets, which utilize liquid lead-bismuth eutectic (LBE) as the target material, liquid metal embrittlement (LME) effects on the target structural materials are considered as one of the critical issues that determine the lifetime of the targets. However, nowadays, the LBE embrittlement effects are not yet well understood, particularly under irradiation conditions. The aim of this work was: (a) to investigate the effects of LBE embrittlement on the mechanical properties of ferritic-martensitic (FM) steels (candidate materials for applications in spallation targets) in various states, and (b) to study the mechanisms of LBE embrittlement effects on the mechanical properties of FM steels. These two goals have been reached by studying LBE embrittlement effects on the mechanical properties of the T91 steel in various conditions: (i) the standard normalized and tempered condition, (ii) hardened by tempering at lower temperatures, and (iii) after irradiation under high energy proton and spallation neutron mixed spectrum. Mechanical tests such as slow-strain-rate tensile (SSRT) tests and 3-point bending tests have been performed to characterize the mechanical properties of the T91 steel and to determine the effects of LBE embrittlement on these mechanical properties. Microstructural analyses such as transition electron microscopy (TEM) and scanning electron mirocopy (SEM) observations have been conducted to obtain the microstructural information needed for understanding the embrittlement mechanisms. SSRT tests on the T91 steel in the standard metallurgical (SM) state, tempered at 760°C and denoted as HT760, revealed that it may encounter LBE embrittlement in the temperature range of 300 to 425°C. For the first time, a well defined "ductility trough" for a FM steel – LBE system was evidenced from the reduction of the fracture strain (or total elongation) of the T91 FM steel in this temperature range. SEM observations of the fracture surfaces showed that the fracture mode of the specimens suffering embrittlement effects is brittle transgranular cleavage. SSRT tests on the hardened T91 steel, tempered at 500 or 600°C and denoted as HT500 and HT600, respectively, showed more pronounced LBE embrittlement effects. In comparison to that of HT760, the "ductility troughs" of HT600 and HT500 cover a wider temperature range. The LME onset temperature TE is not exactly known but most probably lower than 150°C, which means close to the melting point temperature (125°C) of LBE. The ductility recovery temperature TR of HT500 is higher than that of HT760 and HT600. The results of SSRT tests on the HT600 and HT500 materials demonstrate clearly that the LBE embrittlement effects on the tensile properties of FM steels are strongly enhanced by hardening. Both the influenced temperature range and the degradation level increase with the hardening amount. The fracture mode of HT600 and HT500 materials tested in liquid LBE at T ≤ 450°C showed mainly brittle transgranular cleavage. SSRT tests on the T91 and F82H FM steels irradiated at temperatures in the range of 140-500°C to doses in the range of 0.048 to 20 dpa revealed significant irradiation-induced hardening and embrittlement effects (loss of ductility) as compared to the unirradiated ones. The SSRT tests performed on irradiated specimens in liquid LBE revealed that the specimens undergo a further reduction in ductility. The ductility of some irradiated specimens is reduced to a very low level of about 2-3%. 3-point bending tests on the T91 steel showed that the load-displacement curves of specimens tested in liquid LBE are different from those of specimens tested in Ar. The evaluated J-integral values of specimens tested in liquid LBE are always lower that of the specimens tested in Ar. It means that the crack propagation in a specimen tested in liquid LBE needs less energy as compared to a specimen tested in Ar. These results suggest that the LBE embrittlement is so-called "crack propagation controlled" rather than "crack nucleation controlled". SEM observations revealed that the crack propagation mode undergoes a ductile-to-brittle transition. In liquid LBE the fracture toughness of FM steels can be reduced by about 20% as compared to that in inert Ar environment. 3-point bending tests on the hardened T91 materials HT500 and HT600 demonstrated that the LBE embrittlement effects are much more drastic for hardened materials, similarly to what was observed as a result from SSRT tests. The JQ value is reduced by as much as 95% for both HT600 and HT500 materials tested in liquid LBE. The hardened materials exhibit unstable crack propagation behavior in liquid LBE in a wide temperature range of 150 to 500°C. 3-point bending tests on the irradiated T91 steel revealed not only irradiation-induced embrittlement effects but also LBE embrittlement effects. The fracture toughness of the irradiated specimens was found to decrease further as a result from testing in liquid LBE. SEM observations of the fracture surface showed a mixture of intergranular and transgranular fracture. The LME phenomena evidenced for the T91 steel – LBE couple could be interpreted using the model of adsorption-induced reduction in cohesion of atomic bonds combined with the Kelly-Tyson-Cottrell (KTC) criterion (σ/τ ≥ σmax/τmax). According to the KTC criterion, the tendency to fail by cleavage increases as (σmax/τmax) decreases. The σmax value can be greatly reduced by adsorption of liquid Pb or Bi atoms at crack tips, while the τmax value can be increased by hardening the steel by means of either deformation, or tempering, or irradiation. Practically, the present results indicate that a great attention should be paid to application of FM steels in liquid LBE, since the mechanical properties of FM steels, such as the ductility and the fracture toughness, can be substantially degraded by LBE embrittlement effects, particularly under irradiation conditions as the LBE embrittlement are enhanced by irradiation-induced hardening. Fundamentally, the observations made in this work suggest that the mechanism responsible for liquid LBE-induced embrittlement effects on FM steels can be essentially interpreted by using the adsorption-induced reduction in cohesion of atomic bonds model combined with the KTC criterion for brittle cleavage fracture.

EPFL2009

APT/TEM characterization of the element segregation in ferritic/martensitic steels after irradiation in SINQ

Lijuan Cui

Ferritic /martensitic (FM) steels are among the most important materials in various nuclear applications. As radiation-induced segregation (RIS) is one of the major factors for embrittlement of nuclear materials, thereofre it is essential to study the RIS in FM steels. A comprehensive study has been performed on F82H, Eurofer 97, and ODS Eurofer steels after irradiation at SINQ by combining APT and TEM techniques. The study reveals that after irradiation, the RIS at low-angle GBs in F82H is less pronounced as compared to the random high-angle GBs, while in ODS Eurofer the RIS is similar. It was also found that the element segregation extent is not influenced by the GB CSL structures. The results confirmed that, as reported in literature, the small interstitial elements, such as C, N, P, exhibit GB segregation behaviour. Whereas, the GB segregation of substitutional elements depends on their atomic size. The undersized elements are enriched, and the oversized elements V and W are generally depleted at GBs. In some special cases, oversized elements are also enriched at GBs after irradiation, which may be caused by the âsolute drag effectâ of O or N. For the elements with giant sizes (with a volume over 2 times of Fe) such as Ca, Sc and K, their RIS behaviour was not known in literature, and in this work they showed strong enrichment at GBs. Further, this study also revealed that the segregation of these elements at interfaces between phases of chromium carbides, vanadium nitrides, and the matrix is essentially similar to that at GBs, except for Ti. The radiation-induced clustering of solutes was also studied. Two kinds of clusters were investigated in this study, including oxide clusters in ODS Eurofer and nano-clusters (N-clusters) with Si, Mn, Ti, Sc, Ca and K. In F82H, with increasing irradiation temperature, the N-clusters are changed from dilute, nonuniform-size and -shape to more uniform-size, compacted, and sphericial shape. Comparing with that of F82H, N-clusters have a lower number density and bigger size in Eurofer 97 irradiated under the same condition ~20 dpa/ ~300 Â°C. In ODS Eurofer, N-clusters were only observed in specimen irradiated at 300 Â°C. Some oxide clusters in the specimen irradiated at 480 Â°C contain a high portion of Si, Mn, Ti, Sc, Ca, and K. The number density of oxide clusters increases and their size decreases with irradiation temperature. Last but not least, the concentrations of spallation transmutants were quantified in detail. The results agree well with that of the neutronic calculations. This comprehensive study of irradiation effects on GB segregation and solute clustering revealed a number of novel features in FM steels. The investigation on the transmutation elements not only greatly contributes to spallation materials technology, but also broadens the basic understanding of atomic size effect on GB segregation. The comparison between the RIS behaviors in F82H / Eurofer 97 and ODS Eurofer demonstrated that the oxide clusters or oxygen could promote the segregation of some elements such as V, W and Ti, which is of significance for the application of ODS steels in intensive irradiation environments.

EPFL2021