Energy-based method for determining tensile properties of structural materials in fusion reactors using small-sized three-point bending specimens

Evaluation of low-activation vanadium alloys for use as structural material in 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.

An overview is presented on the structural materials in fusion reactors such as the first wall/blanket structural materials, the divertor and limiter plate materials and the tritium breeding materials. Special emphasis is placed on the heavy irradiation effects on the blanket structural materials. The problem and the fundamental mechanisms of the irradiation effects and the current candidate materials are reviewed. A future strategy for the irradiation simulation studies is also presented.

The development of reliable fusion energy is one of the most important challenges in this century. The accelerated degradation of structural materials in fusion reactors caused by neutron irradiation would cause severe problems. Due to the lack of suitable fusion neutron testing facilities, we have to rely on ion irradiation experiments to test candidate materials in fusion reactors. Moreover, fusion neutron irradiation effects are accompanied by the simultaneous transmutation production of helium and hydrogen. One important method to study the He–H synergistic effects in materials is multiple simultaneous ion beams (MSIB) irradiation that has been studied for decades. To date, there is no convincing conclusion on these He–H synergistic effects among these experiments. Recently, a multiple ion beam in-situ transmission electron microscopy (TEM) analysis facility was developed in Xiamen University (XIAMEN facility), which is the first triple beam system and the only in-running in-situ irradiation facility with TEM in China. In this work, we conducted the first high-temperature triple simultaneous ion beams irradiation experiment with TEM observation using the XIAMEN facility. The responses to in-situ triple-ion beams irradiation in austenitic steel 304L SS and ferritic/martensitic steel CLF-1 were studied and compared with the results in dual- and single-ion beam(s) irradiated steels. Synergistic effects were observed in MSIB irradiated steels. Helium was found to be critical for cavity formation, while hydrogen has strong synergistic effect on increasing swelling.

Authors

Fusion has a practically limitless fuel supply and is attractive as an energy source. The main goal of fusion research is to construct and operate an energy generating system. Fusion researches also contains fusion structural materials used fusion reactors. Material issues are very important for development of fusion reactors. Therefore, a wide range of fusion structural materials have been considered for fusion energy applications. Zirconium (Zr), Niobium (Nb) and Tantalum (Ta) containing alloys are important structural materials for fusion reactors and many other fields. Naturally Zr includes the 90 Zr (%51.5), 91 Zr (%11.2), 92 Zr (%17.1), 94 Zr (%17.4), 96 Zr (%2.80) isotopes and 93 Nb and 181 Ta include the 93 Nb (%100) and 181 Ta (%99.98), respectively. In this study, the charge, mass, proton and neutron densities and the root-mean-square (rms) charge radii, rms nuclear mass radii, rms nuclear proton, and neutron radii have been calculated for 87-102 Zr, 93 Nb, 181 Ta target nuclei isotopes by using the Hartree–Fock method with an effective Skyrme force with SKM*. The calculated results have been compared with those of the compiled experimental taken from Atomic Data and Nuclear Data Tables and theoretical values of other studies.

High-performance structural materials (HPSMs) are needed for the successful and safe design of fission and fusion reactors. Their operation is associated with unprecedented fluxes of high-energy neutrons and thermomechanical loadings. In fission reactors, HPSMs are used, e.g., for fuel claddings, core internal structural components and reactor pressure vessels. Even stronger requirements are expected for fourth-generation supercritical water fission reactors, with a particular focus on the HPSM's corrosion resistance. The first wall and blanket structural materials in fusion reactors are subjected not only to high energy neutron irradiation, but also to strong mechanical, heat and electromagnetic loadings. This paper presents a historical and state-of-the-art summary focused on the properties and application potential of irradiation-resistant alloys predominantly strengthened by an oxide dispersion. These alloys are categorized according to their matrix as ferritic, ferritic-martensitic and austenitic. Low void swelling, high-temperature He embrittlement, thermal and irradiation hardening and creep are typical phenomena most usually studied in ferritic and ferritic martensitic oxide dispersion strengthened (ODS) alloys. In contrast, austenitic ODS alloys exhibit an increased corrosion and oxidation resistance and a higher creep resistance at elevated temperatures. This is why the advantages and drawbacks of each matrix-type ODS are discussed in this paper.

Effect of tungsten doping on the irradiation resistance of FeCrV-based refractory medium entropy alloy for potential nuclear applications

Structural changes in a copper alloy due to helium implantation

Composition optimization of high strength and ductility ODS alloy based on machine learning

The objective of this work is to provide a database on the baseline tensile properties of candidate V-Cr-Ti alloys. Vanadium-base alloys of the V-Cr-Ti system are attractive candidates for use as structural materials in fusion reactors. The current focus of the U.S. program of research on these alloys is on the V-(4-6)Cr-(3-6)Ti alloys containing 500-1000 wppm Si. In this paper, we present experimental results on baseline tensile properties of V-Cr-Ti alloys measured at 230-700{degrees}C, with an emphasis on the tensile properties of the U.S. reference alloy V-4Cr-4Ti. The reference alloy was found to exhibit excellent tensile properties up to 700{degrees}C. 9 refs., 8 figs., 1 tab.

Heterogeneous microstructure and anisotropic mechanical properties of reduced activation ferritic/martensitic steel fabricated by wire arc additive manufacturing

This paper describes the facilities of the Institute of Atomic Energy of the National Nuclear Centre of the Republic of Kazakhstan (IAE NNC RK) (Kurchatov, Kazakhstan) designed to conduct studies on the interaction of hydrogen isotopes with materials of nuclear and fusion reactors with and without neutron irradiation. Experiments with sample irradiation are carried out at the LIANA facility, which is located at the IVG.1M reactor. The VIKA and TiGrA installations are designed to conduct experiments with materials before or after irradiation using the temperature-programmed desorption spectroscopy (VIKA), thermogravimetry, and differential scanning calorimetry (TiGrA) methods. The main results of some experimental studies are also presented.

He bubble sites in implanted copper alloy

In recent years, within the fusion long-term programmes, attention has been devoted to the characterization of Chromium (Cr) alloys, in view of their elevated corrosion resistance, low activation properties and high-temperature mechanical strength.As part of the European Fusion Programme, an activity has been launched in 1999 with the aim of exploring the potential of Cr alloys as structural materials in fusion reactors, for example, as first wall or blanket materials. Recent investigations have focused attention on two commercially available materials: high-purity 99.7% Cr (DUCROPUR) and Cr alloyed with 5% Fe and 1%Y203 (DUCROLLOY), both of which have shown excellent low activation characteristics.The mechanical properties of these two alloys, in both as-received and heat-treated conditions, have been characterized at SCK•CEN by means of tensile, instrumented impact and static three-point bend tests, using standard and sub-size specimens. Tensile tests have also been carried out on samples irradiated at 300 °C in the BR2 reactor in Mol up to an accumulated dose of about 0.5 dpa.