Development and characterisation of a new ODS ferritic steel for fusion reactor application

This paper describes the microstructure, tensile properties and Charpy impact resistance of a reduced activation oxide dispersion strengthened ferritic steel Fe-14Cr-2W-0.3Ti-0.3Y₂O₃ produced by mechanical alloying of a pre-alloyed, gas atomised steel powder with Y₂O₃ particles, compaction by hot extrusion at 1100 °C, hot rolling at 700 °C and heat treatment at 1050 °C for 1 h. At room temperature the material exhibits a high ultimate tensile strength of about 1420 MPa and high yield strength of about 1340 MPa in the transverse direction. In the longitudinal direction the values are about 10% lower, due to the anisotropy of the microstructure (elongated grains in the rolling direction). At 750 °C the material still exhibits relatively high yield strengths of about 325 MPa and 305 MPa in the longitudinal and transverse directions, respectively. The material exhibits reasonable uniform and total elongation values over the temperature range 23-750 °C, in both transverse and longitudinal directions. The material exhibits weak Charpy impact properties in the transverse direction. Charpy impact properties are slightly better in the longitudinal direction, with upper shelf energy of about 4.2 J and a ductile-to-brittle transition temperature of about 8.8 °C.     

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

The effects of neutron irradiation on the mechanical properties of welded joints made of austenitic stainless steels have been investigated. The materials are welded AISI 304 and AISI 347, so-called test weld materials, irradiated with neutrons at 573 K to doses of 0.3 and 1.0 dpa. In addition, an AISI 304 from a decommissioned pressurised water reactor, so-called in-service material, which had accumulated a maximum dose of 0.35 dpa at about 573 K, was investigated. The mechanical properties of heat-affected zones and base materials were analysed before and after irradiation. Tensile parameters were determined at room temperature and at 573 K, for all materials and irradiation conditions. In the test weld materials it is found that radiation hardening is lower and loss of ductility is higher in the heat-affected zone than in the base material. In the in-service material radiation hardening is about the same in heat-affected zone and base material. After irradiation, deformation takes place by stacking faults and twins, at both room temperature and high temperature, contrary to unirradiated materials, where deformation takes place by twinning at room temperature and by dislocation cells at high temperature. No defect free channels are observed.     

Influence of hot rolling and high speed hydrostatic extrusion on the microstructure and mechanical properties of an ODS RAF steel

An argon gas atomized, pre-alloyed Fe-14Cr-2W-0.3Ti (wt.%) reduced activation ferritic (RAF) steel powder was mechanically alloyed with 0.3wt.% Y₂O₃ nano-particles in an attritor ball mill and consolidated by hot isostatic pressing at 1150 °C under a pressure of 200 MPa for 3 h. In the aim to improve its mechanical properties the ODS steel was then submitted to a thermo-mechanical treatment (TMT): hot rolling (HR) at 850 °C or high speed hydrostatic extrusion (HSHE) at 900 °C, followed by heat treatment (HT).Transmission electron microscopy (TEM) observations of the ODS alloys after TMT and heat treatment revealed the presence of elongated grains in the longitudinal direction, with an average width of 8 μm and an average length of 75 μm, and equiaxed grains, a few microns in diameter, in the transverse direction. Two populations of oxide particles were observed by TEM: large Ti-Al-O particles, up to 250 nm in diameter, usually located at the grain boundaries and small Y-Ti-O nanoclusters, about 2.5 nm in diameter, uniformly distributed in the matrix. Charpy impact tests revealed that the HSHE material exhibits a larger upper shelf energy (5.8 J) than the HR material (2.9 J). The ductile-to-brittle transition temperature of both alloys is relatively high, in the range of 55-72 °C. Tensile mechanical properties of both ODS alloys were found satisfactory over the full range of investigated temperatures (23-750 °C). The HSHE material exhibits better tensile strength and ductility than the HR material. These results indicate that HSHE can be considered as a promising TMT method for improving the mechanical properties of ODS RAF steels.     

The formation and evolution of oxide particles in oxide-dispersion-strengthened ferritic steels during processing

The fabrication of oxide-dispersion-strengthened (ODS) steels is a multi-stage process involving powder ball milling, degassing and consolidation by hot isostatic pressing. Y is introduced by mechanical alloying (MA) with either Y₂O₃ or Fe₂Y so a high density of Y–Ti–O-based oxide nanoparticles is formed. The evolution of ～2 nm oxide nanoparticles and larger >5 nm grain boundary oxides has been studied at each step of the processing. The nanoparticle dispersions produced in material MA with Fe₂Y were comparable to those produced by alloying with Y₂O₃. Hence the majority of oxygen which forms the nanoparticles must be incorporated from the atmosphere during MA, rather than being introduced via the alloying additions. During mechanical alloying, a high density of subnanometer particles are formed (2.5 ± 0.5 × 10²⁴ m^?3). The oxygen content of the nanoparticles decreases slightly on annealing, and then the composition of the nanoparticles remains constant throughout subsequent processing stages. The nanoparticle size increases during processing up to ～2 nm radius, while the number density decreases to 4 ± 0.5 × 10²³ m^?3. Grain boundary oxides (>5 nm) have a Ti–Cr–O-rich shell, and contain no Y before consolidation, but have similar core composition to the matrix nanoparticles after consolidation. The formation of the larger grain boundary oxides is shown to take place during the degassing and consolidation stages, and this occurs at the expense of the nanoparticles in the matrix. This work provides a mechanistic understanding of the importance of controlling the oxygen content in the powder during MA, and the resulting impact on the formation of the ODS microstructure.     

Progress in IFMIF Engineering Validation and Engineering Design Activities

The International Fusion Materials Irradiation Facility (IFMIF) Engineering Design and Engineering Validation Activities (EVEDA) are being developed in a joint project in the framework of the Broader Approach (BA) Agreement between EU and Japan. This project has now entered into a crucial phase as the engineering design of IFMIF is now being formulated in a series of 3 subsequent phases for delivering an Interim IFMIF Engineering Design Report (IIEDR) by mid of 2013. Content of these phases is explained, including the plant configuration detailing the 5 IFMIF facilities and their systems. Together with the Engineering Design Activities, prototyping sub-projects are pursued in the Engineering Validation Activities which consist of the design, manufacturing and testing of the following prototypical systems: Linear IFMIF Prototype Accelerator (LIPAc), EVEDA Lithium Test Loop (ELTL), and High Flux Test Module (HFTM) with the prototypical helium cooling loop (HELOKA). Highlights are described from recent experiments in the Engineering Validation Activities.     

Effect of thermo-mechanical treatments on the microstructure and mechanical properties of an ODS ferritic steel

The Fe-14Cr-2W-0.3Ti-0.3Y₂O₃ oxide dispersion strengthened (ODS) reduced activation ferritic (RAF) steel was fabricated by mechanical alloying of a pre-alloyed, gas atomised powder with yttria nano-particles, followed by hot isostatic pressing and thermo-mechanical treatments (TMTs). Two kinds of TMT were applied: (i) hot pressing, or (ii) hot rolling, both followed by annealing in vacuum at 850 °C.The use of a thermo-mechanical treatment was found to yield strong improvement in the microstructure and mechanical properties of the ODS RAF steel. In particular, hot pressing leads to microstructure refinement, equiaxed grains without texture, and an improvement in Charpy impact properties, especially in terms of the upper shelf energy (about 4.5 J). Hot rolling leads to elongated grains in the rolling direction, with a grain size ratio of 6:1, higher tensile strength and reasonable ductility up to 750 °C, and better Charpy impact properties, especially in terms of the ductile-to-brittle transition temperature (about 55 °C).     

HRTEM of dislocation cores and thin-foil effects in metals and intermetallic compounds

Examples of the observation and analysis of dislocation cores and dislocation fine structure in metals and intermetallics using high resolution transmission electron microscopy are discussed. Specific examples include the 60 degrees dislocations in aluminum, a011 edge dislocations in NiAl, and screw dislocations in Ni3Al. The effect of the thin TEM foils on the structure and imaging of these dislocations is discussed in light of embedded atom method calculations for several configurations and coupled with image simulations. Some generalizations based on these calculations are discussed. These analyses enables determination of the spreading or decomposition of the edge component of the cores, both in and out of the glide plane, which can have significant implications for the modeling of macroscopic behavior.     

Effect of ion-irradiation on the microstructure and microhardness of the W-2Y2O3 composite materials fabricated by sintering and hot forging

A W-2Y₂O₃ material was developed in collaboration with the Plansee Company (Austria). An ingot of the material having approximate dimension of 95 mm × 20 mm was fabricated by mixing the elemental powders followed by pressing, sintering and hot forging. The microstructure of the W-2Y₂O₃ composite was investigated using transmission electron microscopy (TEM). The microhardness was studied using nano-indentation technique. We observed that the W-grains having a mean size of about 1 μm already formed and these grains contain very low density of dislocations. The size of the yttria particles was between 300 nm and 1 μm and the Berkovich hardness was about 4.8 GPa. The specimens were irradiated/implanted with Fe and He ions at JANNuS facility located at Orsay/Saclay, France. The TEM disks kept were irradiated/implanted at 300 and 700 °C using Fe and He ions with an energy of 24 and 2 MeV, respectively. The calculated radiation dose was about 5 dpa produced by Fe ions and total He content is 75 appm at both 300 and 700 °C. From the TEM investigation of irradiated samples, few radiation loops are present on the W grains, whereas on yttria particles, the radiation induced damages appear as voids. Berkovich hardness of the irradiated sample is higher than that of the non-irradiated sample. Results on the microstructure and microhardness of the ion-irradiated W-2Y₂O₃ composites are presented in detail.     

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.     

Recent progress in research on tungsten materials for nuclear fusion applications in Europe

The current magnetic confinement nuclear fusion power reactor concepts going beyond ITER are based on assumptions about the availability of materials with extreme mechanical, heat, and neutron load capacity. In Europe, the development of such structural and armour materials together with the necessary production, machining, and fabrication technologies is pursued within the EFDA long-term fusion materials programme. This paper reviews the progress of work within the programme in the area of tungsten and tungsten alloys. Results, conclusions, and future projections are summarized for each of the programme’s main subtopics, which are: (1) fabrication, (2) structural W materials, (3) W armour materials, and (4) materials science and modelling. It gives a detailed overview of the latest results on materials research, fabrication processes, joining options, high heat flux testing, plasticity studies, modelling, and validation experiments.