N. Baluc

The microstructure and associated tensile properties of irradiated fcc and bcc metals

The differences and similarities of behaviour between fee and bcc metals after irradiation have been investigated. For this purpose, fee Cu, Pd and 304 stainless steel and bcc Fe, Mo and Mo-5% Re were irradiated with either neutrons or 590 MeV protons at temperatures below recovery stage V. It is shown that a dense population of defect clusters (up to 10(22)-10(24) m(-3)) develops, the type of cluster formed depending apparently on the stacking fault energy. In the case of stacking fault tetrahedra formed in Cu, their size is independent of dose, while interstitial loops in stainless steel grow at neutron doses higher than 1 dpa. The defect microstructure is found to be independent of the recoil energy spectra in this temperature region, but shows a very strong dependence on the type of crystalline structure. The results of tensile testing indicate the presence of radiation hardening, starting at very low doses as an upper yield point develops followed by a (serrated) yield region. The main deformation mode observed is dislocation channeling. The hardening is modelled in terms of the initial dislocation locking by the irradiation-induced defects followed by the dispersed hardening induced by the global distribution of clusters in the matrix

Status of R&D activities on materials for fusion power reactors

Current R&D activities on materials for fusion power reactors are mainly focused on plasma facing, structural and tritium breeding materials for plasma facing (first wall, divertor) and breeding blanket components. Most of these activities are being performed in Europe, Japan, the Peoples Republic of China, Russia and the USA. They relate to the development of new high temperature, radiation resistant materials, the development of coatings that will act as erosion, corrosion, permeation and/or electrical/MHD barriers, characterization of candidate materials in terms of mechanical and physical properties, assessment of irradiation effects, compatibility experiments, development of reliable joints, and development and/or validation of design rules. Priorities defined worldwide in the field of materials for fusion power reactors are summarized, as well as the main achievements obtained during the last few years and the near-term perspectives in the different investigation areas.

On the potentiality of using ferritic/martensitic steels as structural materials for fusion reactors

Reduced activation ferritic/martensitic (RAFM) steels are the reference structural materials for future fusion reactors. They have proven to be a good alternative to austenitic steels for their higher swelling resistance, lower damage accumulation and improved thermal properties. However, irradiated RAFM steels exhibit a low temperature hardening and an increase in the ductile-to-brittle transition temperature, which imposes a severe restriction on reactor applications at temperatures below about 350°C. Furthermore, a high density of small cavities (voids or helium bubbles) has been recently evidenced in specimens irradiated with a mixed spectrum of neutrons and protons at about 300°C at a dose of 10 dpa, which could affect their fracture properties at intermediate temperatures. The upper temperature for the use of RAFM steels is presently limited by a drop in mechanical strength at about 500°C. New variants that can withstand higher temperatures are currently being developed, mainly using a stable oxide dispersion. This paper reviews European activity in the development of RAFM steels.

Microstructure and mechanical properties of two ODS ferritic/martensitic steels

A microstructural analysis and tensile tests were performed on two oxide dispersion strengthened ferritic/martensitic steels. Dispersion hardening represents an interesting approach to improve the mechanical properties at elevated temperatures, as they are foreseen in the future fusion reactor. while maintaining the inherent advantages of the ferritic/martensitic steel in an irradiation environment (high thermal conductivity and low swelling rate). The base material is the ferritic/martensitic steel EUROFER 97 with the chemical composition Fe, 8.9 wt%,) Cr, 1.1 wt%,) W, 0.47 wt% Mn, 0.2 wt% V, 0.14 wt%) Ta and 0.11 wt%) C. In one steel the strengthening material Y2O3 represents 0.3 wt% while in the second it represents 0.5 wt%. It appears that the ODS with 0.3 wt%) yttria presents, in terms of critical stress and uniform elongation, a better mechanical behaviour than the base material up to 500 degreesC and still maintains fair properties up to 700 degreesC

Irradiation-induced stacking fault tetrahedra in fcc metals

Irradiation induces the formation of stacking fault tetrahedra (SFTs) in a number of fcc metals, such as stainless steel and pure copper. In order to understand the role of the materials parameters on this formation, pure Cu, Ni, Pd and Al, having a respective stacking fault energy of 45, 125, 180 and 166 mJ m−2, have been irradiated with high energy protons to a dose of about 10−2 dpa at room temperature. The irradiation-induced microstructure has been investigated using transmission electron microscopy. All irradiated metals but Al present SFTs. The proportion of perfect, truncated and grouped SFTs has been determined. The SFT energy as a function of size has been calculated using elasticity of the continuum, with respect to the energy of a number of other possible defect configurations. It appears that the key parameters are the stacking fault energy and the shear modulus. Their implication on the formation and stability of the SFTs is discussed.

Structural materials for fusion reactors

Note: 2nd Int. Conference on Advanced Materials and Structures, Politehnica Univ. of Timisoara, Timisoare, Romania, September 2002 Reference CRPP-CONF-2002-015 Record created on 2008-05-13, modified on 2017-05-12

Microindentation of Al-Cu-Fe icosahedral quasicrystal

Reference CRPP-ARTICLE-1999-033doi:10.1016/S1359-6462(99)00242-0View record in Web of Science Record created on 2008-04-16, modified on 2017-05-12

Weak beam transmission electron microscopy imaging of superdislocations in ordered Ni3Al

Abstract In order to provide quantitative information about dissociation modes of superdislocations in ordered Ni3Al, weak-beam transmission electron microscopy observations of dislocation core structures have been combined with image simulations. The use of image simulations allowed us to interpret intensity peaks on weak-beam micrographs in terms of actual partial dislocations and splitting distances, and corresponding fault plane energies could be subsequently inferred. A description of the main parameters to be considered in the calculations is given below and the validity of such calculations is discussed in a critical way. The impact of these results on interpretation of the anomalous flow stress behaviour in Ni3Al is also addressed.

The effects of irradiation and testing temperature on tensile behaviour of stainless steels

Abstract Irradiated 304 and 316 stainless steel samples were investigated. The steels were neutron irradiated to 1.5 and 7.5 dpa at 550 K. Similar types of steels were irradiated at 550 K with 590 MeV protons in the PIREX facility at PSI, Switzerland. The doses reached in this case were 0.15 and 0.3 dpa. The stress–strain relationships at different temperatures were measured. The irradiation and deformation microstructures were investigated using transmission electron microscopy (TEM). Two modes of deformation were found, twinning and channelling, depending on the testing temperature. These results are discussed in terms of deformation mechanisms at different temperatures correlated with radiation hardening. Finally, possible correlations between deformation modes and previous irradiation assisted stress corrosion cracking (IASCC) studies are discussed.

Optimization of the chemical composition and manufacturing route for ODS RAF steels for fusion reactor application

As the upper temperature for use of reduced activation ferritic/martensitic steels is presently limited by a drop in mechanical strength at about 550 degrees C, Europe, Japan and the US are actively researching steels with high strength at higher operating temperatures, mainly using stable oxide dispersion. In addition, the numerous interfaces between matrix and oxide particles are expected to act as sinks for the irradiation-induced defects. The main RD activities aim at finding a compromise between good tensile and creep strength and sufficient ductility, especially in terms of fracture toughness. Oxide dispersion strengthened (ODS) reduced activation ferritic (RAF) steels appear as promising materials for application in fusion power reactors up to about 750 degrees C. Six different ODS RAF steels, with compositions of Fe-(12-14)Cr-2W-(0.1-0.3-0.5)Ti-0.3Y(2)O(3) (in wt%), were produced by powder metallurgy techniques, including mechanical alloying, canning and degassing of the milled powders and compaction of the powders by hot isostatic pressing, using various devices and conditions. The materials have been characterized in terms of microstructure and mechanical properties. The results have been analysed in terms of optimal chemical composition and manufacturing conditions. In particular, it was found that the composition of the materials should lie in the range Fe-14Cr-2W-(0.3-0.4)Ti-(0.25-0.3)Y2O3, as 14Cr ODS RAF steels exhibit higher tensile strength and better Charpy impact properties and are more stable than 12Cr materials (no risk of martensitic transformation), while materials with 0.5% Ti or more should not be further investigated, due to potential embrittlement by large TiO2 particles.