V.A. Kazakov

Research and development on vanadium alloys for fusion applications

The current status of research and development on unirradiated and irradiated V-Cr-Ti alloys intended for fusion reactor structural applications is reviewed, with particular emphasis on the flow and fracture behavior of neutron-irradiated vanadium alloys. Recent progress on fabrication, joining, oxidation behavior, and the development of insulator coatings is also summarized. Fabrication of large (>500 kg) heats of V-4Cr-4Ti with properties similar to previous small laboratory heats has now been demonstrated. Impressive advances in the joining of thick sections of vanadium alloys using GTA and electron beam welds have been achieved in the past two years, although further improvements are still needed.

Critical issues and current status of vanadium alloys for fusion energy applications

Vanadium alloys are widely regarded as possessing desirable mechanical and physical properties for application as structural materials in fusion power systems. The bulk of the recent research on vanadium is focussed on ternaries containing 4–5% Cr and 4–10% Ti. The aim of this paper is to review significant results generated by the international research and development community on this alloy system and to highlight the critical issues that must be resolved before alloy development can proceed to the next stage. Recent progress on understanding the physical metallurgy, fabrication and joining behavior, and compatibility with hydrogen and oxygen containing environments of unirradiated vanadium alloys is discussed. The effect of low-temperature neutron irradiation on mechanical properties and their relationship to the observed microstructure are briefly summarized. Current efforts to characterize the high-temperature mechanical properties, develop constitutive equations describing flow and fracture, and understand and mitigate the effects of non-metallic impurities on properties are presented.

Effects of neutron irradiation on properties of refractory metals

In the framework of ITER-supporting R&D, the USSR team studied the refractory materials W, Mo, Ta. Research reactors with different neutron spectra are used for materials irradiation: SM-2, BOR-60. Special attention is paid to electric resistance, density mechanical strength and ductility, fatigue life, destruction mechanisms and material structure. Materials were irradiated in a wide range of doses and temperatures SM-2 reactor: 1−5 × 10 21 n/cm 2 , BOR-60 reactor: 0.8−1.6 × 10 22 n/cm 2 . The investigation of the effect of the irradiation temperature on the mechanical properties of W and Mo alloys has shown that at T irr ~ 250–550°C irradiation embrittlement and intercrystalline fracture take place.

Prospects and problems using vanadium alloys as a structural material of the first wall and blanket of fusion reactors

Abstract Vanadium-based alloys are most promising as low activation structural materials for DEMO. It was previously established that high priority is to be given to V-alloys of the VTiCr system as structural materials of a tritium breeding blanket and the first wall of a fusion reactor. However, there is some uncertainty in selecting a specific element ratio between the alloy components in this system. This is primarily explained by the fact that the properties of V-alloys are dictated not only by the ratio between the main alloying elements (here Ti and Cr), but also by impurities, both metallic and oxygen interstitials. Based on a number of papers today one can say that VTiCr alloys with insignificant variations in the contents of the main constituents within 5–10 mass% Ti and 4–6 mass% Cr must be taken as a base for subsequent optimization of chemical composition and thermomechanical working. However, the database is obviously insufficient to assess the ecological acceptability (activation), physical and mechanical properties, corrosion and irradiation resistance and, particularly, the commercial production of alloys. Therefore, there is a need for comprehensive studies of promising V-alloys, namely V4Ti4Cr and V10Ti5Cr.

Effects of neutron irradiation at 70 200 °C in beryllium

Abstract At present beryllium is considered one of the metals to be used as a plasma facing and blanket material. This paper presents the investigations of several Russian beryllium grades fabricated by HE and HIP technologies. Beryllium specimens were irradiated in the SM reactor at 70–200 °C up to a neutron fluence (0.6–3.9)×1022 cm−2 (E>0.1 MeV). It is shown that the relative mass decrease of beryllium specimens that were in contact with the water coolant during irradiation achieved the value >1.5% at the maximum dose. Swelling was in the range of 0.2–1.5% and monotonically increasing with the neutron dose. During mechanical tensile and compression tests one could observe the absolute brittle destruction of the irradiated specimens at the reduced strength level in comparison to the initial state. A comparatively higher level of brittle strength was observed on beryllium specimens irradiated at 200 °C. The basic type of destruction of the irradiated beryllium specimens is brittle and intergranular with some fraction of transgranular chip.

Radiation embrittlement of low-alloyed Mo alloys

Abstract The influence of irradiation in the BOR-60 reactor at 250–1020°C up to fluences 0.013–7.6 × 10 26 n/m 2 ( E > 0.1 MeV) on the mechanical properties of a number of low-alloyed Mo alloys has been studied. Zr, Ti, Ru, B, C were used as alloying elements. Significant radiation strengthening and embrittlement of all the alloys investigated took place during the course of irradiation. According to tension test results, the brittle—ductile transition temperature, T k , was 600–700°C. Under comparable irradiation conditions, alloying had no substantial effect on the extent of irradiation embrittlement, since it did not prevent the radiation defect formation causing strengthening and, therefore, embrittlement. It was assumed that improvement of the Mo radiation resistance could be only obtained by high alloying with such elements as rhenium.

Tensile properties and fracture behaviour of V–Cr–Ti alloys after neutron irradiation at 330°C

Results of irradiation effects in the fast neutron spectrum of the BOR-60 reactor on the mechanical properties and fracture behaviour of vanadium as well as on binary and ternary alloys (including V–4Cr–4Ti and V–5Cr–10Ti) are presented. The irradiation was carried out at 330°C in a 7Li isotope environment to a damage dose of 18 dpa. It is shown that all alloys without exception experienced severe radiation embrittlement. Most specimens failed without discernible traces of ductile strain at room temperature. At a test temperature of 350°C, ductility was observed but the total elongation, as a rule, did not exceed 1%. The fracture was mixed: transgranular brittle cleavage and ductile dimple rupture. As the test temperature was increased from 20°C to 350°C, the brittle fracture fraction decreased. At 600°C only ductile fracture was observed. The behaviour of V–4Cr–4Ti alloys of American and Russian production is analyzed in the 275–530°C temperature range, where a shift in behaviour from brittle to ductile failure is observed.

Formation of the σ-phase in CrFe alloys under irradiation

Abstract Chromium alloys are perspective structural materials for use at high neutron irradiation doses. The low chromium alloy and alloys with 10% and 35% Fe content were investigated after irradiation at 600–750°C up to 2.8–4.7 × 1026 n/m2 (E > 0.1 MeV). As a result of irradiation, considerable embrittlement occured in all alloys. The brittle—ductile transition temperature, Tk, increases up to 200–600°C. Radiation embrittlement is caused by radiation strengthening due to formation of vacant voids or by formation of the brittle σ-phase. In CrFe alloys, there is radiation-induced segregation leading to the enrichment of the sinks with iron.

Experimental facilities for investigation of structural material properties for fusion reactor under irradiation

Abstract The study of structural and breeding materials for fusion reactors covers a wide range of investigations including the effect of different operating factors; irradiation is the main factor. This paper presents basic reactor characteristics, the types of investigations on structural and breeding materials carried out at these reactors, and the reactor irradiation conditions. The design of equipment used for parameter control during the irradiation is also discussed. CM-2 and BOR-60 reactors are primarily used to irradiate structural materials for the blanket, first wall and divertor at temperatures of 80 and 350°C and fluences up to 5 × 10 22 n/cm 2 . The IVV-2 reactor is used to investigate breeding blanket materials and to study the problems of hydrogen/tritium permeability and recovery from Li-Pb eutectic and through 0.4C-16Cr-11Ni-3Mo-Ti steel. In addition, there are facilities for carrying out irradiation experiments at cryogenic temperatures as well as in different media.

Features of radiation damage of vanadium and its alloys at a temperature of 330–340°C

Abstract Microstructural changes of vanadium alloys after irradiation at 340°C to 12 dpa in the BOR-60 reactor in 7Li environment is analyzed. Materials are vanadium and its alloys V–3Ti, V–3Fe, V–6Cr, V–4Cr–4Ti, V–5Cr–10Ti, V–6Cr–1Zr–0.1C. Void formation was observed in the binary alloys V–3Fe, V–3Ti and V–6Cr. It is shown that three–four-fold increase in V–4Cr–4Ti yield stress is produced by the formation of dislocation loops (DLs) and fine radiation-induced precipitates (RIPs) with a density of 1.7×10 17 cm −3 . It is expected that embrittlement of the welds will be worse because density of DLs and RIPs is 1.4–1.6 times higher. Besides, invisible coherent or semi-coherent RIPs are formed in the fusion zone. Elemental maps of the rupture surface of irradiated V–4Cr–4Ti are presented.