In situ observation of damage structure in ODS austenitic steel during electron irradiation (original) (raw)
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Effect of Oxide Particles and Pre-Implanted Helium on Defect Evolution during Electron Irradiation
MATERIALS TRANSACTIONS, 2014
In-situ observations of ferritic/martensitic steels by electron irradiation with a 1.25 MeV high voltage microscope at 573 K were carried out to study damage evolution in the steels. The development of interstitial type loops and cavities in both of the two steels, F82H-IEA and F82H-ODS, showed smaller and more numerous defects in the ODS steel. The cavities were formed preferentially at the interface between oxide particles and matrix. The results suggest that ODS particles may function to suppress the nucleation and growth of loops and cavities arising from irradiation. The effect of pre-implanted helium was also studied. The pre-implanted helium led to a homogenous distribution of black dots and cavities in the steels, and these may act as sinks for point defects arising from irradiation, causing a suppression of the subsequent growth of loops and cavities. The hardening corresponding to the microstructural evolution was estimated by assuming parameters extracted from the ion irradiation.
Irradiation-produced defects in austenitic stainless steel
1971
The microstructure of annealed AISI Type 304 and type 316 stainless steels has been characterized by transmission electron microscopy as a function of fast reactor irradiation at fluence levels from 4×1021 to 7×1022 n per sq cm (E>0.1 mev) and at irradiation temperatures from 370° to 700°C. Several irradiation produced defect types where found: voids, Frank faulted loops, perfect loops, dislocation networks, and precipitates. Void number density obeys a power law relationship to fluence, wherein the exponent increases with increasing temperature from 0.8 to 1.4 over the irradiation temperatures investigated. The void size is nearly independent of fluence and increases with increasing temperature. The upper limit irradiation temperature for void formation is about 650° to 700°C. The density and size of Frank faulted loops followed trends similar to those found for voids to temperatures of ∼550°C where unfaulted loops, perfect loops, and dislocation networks coexist. These experime...
Oxide Dispersion Strengthened steels are materials for the first wall of proposed nuclear fusion reactors including DEMO. They contain a nanoparticle dispersion of Y2O3 which restricts dislocation motion. The particles act as low energy sites capable of trapping helium. The experiments reported here focus on the mechanical effects of ion implantation on a model Fe-Cr alloy and an ODS equivalent. Samples were helium and iron ion implanted as a simulation of neutron damage and material transmutation. The main restriction of using ion implantation is the limited damage depth available in samples (<3μm), therefore nanoindentation was used to provide information on the surface hardness changes. Transmission Electron Microscopy was used to see how the dislocation structure within these test volumes changes with implantation.
Radiation induced microstructures in ODS 316 austenitic steel under dual-beam ions
Journal of Nuclear Materials, 2014
An ODS 316 austenitic steel was fabricated and irradiated using dual ion beams (1 MeV Kr + and 15 keV He + ) with in-situ transmission electron microscope (TEM) observation. Cavities formed at a low dose in samples irradiated with simultaneous helium injection. It was found that Y-Ti-O particles acted as strong traps for cavity formation at low doses. Helium exhibited a significant effect on cavity development. Cavities were also preferentially nucleated along grain boundaries, phase boundaries and twin boundaries. Irradiation induced lattice defects mainly consisted of small 1/2h1 1 0i perfect loops and 1/3h1 1 1i Frank loops. An increment of helium injection rate also greatly enhanced the Frank loop growth. Small (<10 nm) Y-Ti-O particles were found to be unstable after irradiation to high doses. M 23 C 6 precipitates were observed after irradiation and helium might play a major role in their formation.
Development of damage structure in oxide dispersion strengthened steels
Journal of Nuclear Materials, 1991
Power Reactor and Nuclear Fuel Development Corporation, Oarai, Electron irradiation experiments on Oxide Dispersion Strengthened steels (ODS ferritic steels) were performed using a high-voltage electron microscope (HVEM) with the aim of clarifying irradiation behavior. Four types of Fe-13 Cr base steels dispersed with Y,O, particles were irradiated at 673-823 K up to 30 dpa. They were made by mechanical alloying with high-energy milling.
Corrosion Science, 2019
The role of irradiation and irradiation defects on the oxidation first stages of 316 L alloy was investigated. A sample with both a proton pre-irradiated and an unirradiated area was exposed to a simulated PWR environment during 24 hours. Irradiation defects and Radiation Induced Segregation at grain boundary and on irradiation defects were characterized and quantified and their effect on the oxidation was evaluated. Irradiation affects the morphology, thickness and chemistry of the oxide layers formed. It enhances the oxidation kinetic and induces the formation of an inner oxide richer in chromium. Defects induced by irradiation act as preferential nucleation sites.
Effects of helium on radiation-induced defect microstructure in austenitic stainless steel
Journal of Nuclear Materials, 2000
In the construction materials surrounding the spallation neutron source (SNS) mercury target, considerable quantities of transmutation products, particularly hydrogen and helium, will be generated due to the exposure to a high ux of 1 GeV protons and associated neutrons. In an eort to investigate the eects of high helium, therefore, bubble formation and defect clustering processes in AISI 316 LN austenitic steel were studied as a function of helium concentration and displacement damage dose with 360 keV He and 3500 keV Fe ion beams at 200°C. Helium irradiation was less eective in producing defects such as black dots and dislocation loops than Fe ion irradiation at equivalent displacement dose. On the other hand, the formation of helium bubbles produced a strong depressive eect on the growth of loops and the evolution of line dislocations. The results indicated that the eect of helium bubbles was augmented as the bubble number density and size increased with increasing helium beyond 1 atomic percent (at.%). In such a case, the eect of helium bubbles can be more important than that of radiation-induced defects on the evolution of microstructure and the change in mechanical properties.
Nanostructure evolution in ODS steels under ion irradiation
Nuclear Materials and Energy, 2016
Excellent mechanical properties of ODS steels are directly related to the high density of homogeneously distributed, well-formed oxide particles (such as Y 2 O 3 , or Y-Ti-O). However, atom probe tomography study of ODS steels revealed that in addition they contain almost a hundred times more nanoclusters enriched in Y, O and V/Ti (if present in the alloy composition) than larger oxide particles. In this work, we carried out atom probe tomography (APT) and transmission electron microscopy (TEM) studies of three different ODS steels produced by mechanical alloying: ODS Eurofer, 13.5Cr ODS and 13.5Cr-0.3Ti ODS. These materials were investigated after irradiation with Fe (5.6 MeV) or Ti (4.8 MeV) ions up to 10 15 ion/cm 2 and part of them up to 3 × 10 15 ion/cm 2. In all cases, areas for TEM investigation were cut at a depth of ∼ 1.3 μm from the irradiated surface corresponding to the peak of the radiation damage dose. It was shown that after irradiation at RT and at 300 °С the number density of oxide particles in all the samples grew up. Meanwhile, the fraction of small particles in the size distribution has increased. APT revealed an essential increase in nanoclusters number and a change of their chemical composition at the same depth. The nanostructure was the most stable in 13.5Cr-0.3Ti ODS irradiated at 300 °С : the increase of the fraction of small oxides was minimal and no change of nanocluster chemical composition was detected.
Microstructural development under irradiation in European ODS ferritic/martensitic steels
Journal of Nuclear Materials, 2006
Oxide dispersion strengthened steels based on the ferritic/martensitic steel EUROFER97 are promising candidates for a fusion reactor because of their improved high temperature mechanical properties and their potential higher radiation resistance relative to the base material. Several EUROFER97 based ODS F/M steels are investigated in this study. There are the Plansee ODS steels containing 0.3 wt% yttria, and the CRPP ODS steels, whose production route is described in detail. The reinforcing particles represent 0.3-0.5% weight and are composed of yttria. The effect of 0.3 wt% Ti addition is studied. ODS steel samples have been irradiated with 590 MeV protons to 0.3 and 1.0 dpa at room temperature and 350°C. Microstructure is investigated by transmission electron microscopy and mechanical properties are assessed by tensile and Charpy tests. While the Plansee ODS presents a ferritic structure, the CRPP ODS material presents a tempered martensitic microstructure and a uniform distribution of the yttria particles. Both materials provide a yield stress higher than the base material, but with reduced elongation and brittle behaviour. Ti additions improve elongation at high temperatures. After irradiation, mechanical properties of the material are only slightly altered with an increase in the yield strength, but without significant decrease in the total elongation, relative to the base material. Samples irradiated at room temperature present radiation induced defects in the form of blacks dots with a size range from 2 to 3 nm, while after irradiation at 350°C irradiation induced a 0 h1 0 0i{1 0 0} dislocation loops are clearly visible along with nanocavities. The dispersed yttria particles with an average size of 6-8 nm are found to be stable for all irradiation conditions. The density of the defects and the dispersoid are measured and found to be about 2.3 • 10 22 m À3 and 6.2 • 10 22 m À3 , respectively. The weak impact of irradiation on mechanical properties of ODS F/M steel is thus explained by a lower density of irradiation induced defects relative to the density of reinforcing particles.
Morphology of oxide particles in ODS austenitic stainless steel
Journal of Nuclear Materials, 2013
In this study, identification of the crystal structure and analysis of the orientation relationship of oxide particles in an oxide dispersion strengthened austenitic stainless steel was carried out. High resolution transmission electron microscopy (HRTEM) and energy dispersive spectroscopy showed that most of the oxide particles had a faceted shape and consisted of a complex oxide, the anion-deficient fluorite structure Y 2 Hf 2 O 7. Selected area diffraction patterns and HRTEM indicated that the faceted oxide particle has a cube-on-cube orientation relationship with the surrounding matrix. In addition, strain fields were observed around the oxide particle with given reflection conditions, indicating that it surrounds the oxide particle. The observed strain fields would affect glide dislocation pinning and the migration of irradiationinduced point defects.