Yu. D. Goncharenko

Mechanical properties and microstructure of advanced ferritic–martensitic steels used under high dose neutron irradiation

Abstract Some results of the study of mechanical properties and structure of ferritic–martensitic chromium steels with 13% and 9% chromium, irradiated in the BOR-60 reactor up to different damage doses are presented in this report. Results concerning the behaviour of commercial steels, containing to molybdenum, vanadium and niobium, and developed for the use in fusion reactors, are compared to low-activation steels in which W and Ta replaced Mo and Nb. It is shown that after irradiation to the dose of ∼10 dpa at 400°C 0.1C–9Cr–1W, V, Ta steels are prone to lower embrittlement as deduced from fracture surface observations of tensile specimens. Peculiarities of fine structure and fracture mode, composition and precipitation reactions in steels during irradiation are discussed.

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.

Evolution of the mechanical properties and microstructure of ferritic–martensitic steels irradiated in the BOR-60 reactor

Abstract The effect of neutron irradiation on mechanical properties of low-activation ferritic–martensitic (FM) steels 0.1C–9Cr–1W, V, Ta, B and 0.1C–12Cr–2W, V, Ti, B is studied under tension at temperatures of 330–540 °C and doses of 50 dpa. Steel 0.1C–13Cr–Mo, V, Nb, B was chosen for comparison. At irradiation temperatures of 330–340 °C, the radiation hardening of steel with 9%Cr achieves saturation at a dose of 10 dpa. In this case as compared to steels with 12%Cr, the fracture surface is characterized as ductile without cleavage traces. At irradiation temperatures higher than 420 °C, there is no difference in the behavior of the materials under investigation. The data on radiation creep obtained by direct measurement and from the profilometry data satisfy a model e / σ =B 0 +D S , when B0 and D have the values typical for steels of FM type.

High-temperature beryllium embrittlement

Abstract The neutron irradiation effect on the mechanical properties, swelling and fracture surface structure of various beryllium grades was studied in the BOR-60 reactor at 340 to 350°C up to a fluence of 7.2 × 10 21 n/cm 2 . At a mechanical testing temperature of 400°C there was observed a strong anisotropy of plastic beryllium deformation depending on the direction of sample cutting relative to the pressing direction. An increase of the testing temperature up to 700°C resulted in an abrupt embrittlement of all irradiated samples. In the most part of the surface structure the intercrystallite fracture along the grain boundaries was covered entirely with large pores, 1 to 4 μm in size. It was suggested that the increase rate of pore formation along the grain boundaries resulted from a high-temperature embrittlement under irradiation.

Anodic dissolution of Tc metal in HNO3 solutions

Corrosion and dissolution of Tc metal in 0.5–6.0 M HNO3 without external sources of the redox potential and under the conditions of constant-current electrolysis were studied. The dissolution rates of Tc metal without external potential in relation to the HNO3 concentration and the rates of anodic dissolution of Tc in relation to the anodic current density and HNO3 concentration were determined. The Tc(VII) current efficiencies in 0.5–6.0 M HNO3 solutions were measured. The anodic dissolution of Tc is characterized by the linear growth of the concentration of the dissolved Tc species in solution with time. Examination of the sample surfaces by scanning electron microscopy allows a conclusion that the corrosion degradation has intercrystallite character.

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.

Electrochemical dissolution of Тс–Ru alloys in nitric acid solutions

Electrochemical dissolution of Тс–19 wt % Ru, Тс–45 wt % Ru, and Тс–70 wt % Ru in 1 to 6 M HNO3 solutions in the amperostatic mode was studied. In the solutions formed from anodic dissolution of Тс–Ru alloys, Ru is present in the form of Ru(IV), and Тс, in the form of Тс(VII). A kinetic study of the electrochemical dissolution of the Тс–70 wt % Ru alloy in the examined interval of HNO3 concentrations showed that the alloy dissolved congruently and the Тс(VII) and Ru(IV) accumulation in the solution linearly depended on time. Anodic dissolution of Tc–Ru alloys with lower Ru content was also characterized by a linear increase in the Тс(VII) concentration in the solution with time. On the other hand, formation of poorly soluble hydrated Ru(IV) oxide on the electrode surface was observed simultaneously with accumulation of soluble Ru(IV) species in solution. The rate of electrochemical dissolution of the Тс–70 wt % Ru alloy linearly increased with increasing HNO3 concentration. The rates of electrochemical dissolution of Tc, determined for alloys with lower Ru content, were independent of the HNO3 concentration in the electrolyte. Oxidation of water with the evolution of oxygen on the surface of Tc–Ru alloys was observed simultaneously with the anodic dissolution of Tc and Ru. This process leads to a decrease in the current efficiency of the Tc and Ru dissolution. Examination of the corrosion damage of the working electrode surfaces by scanning electron microscopy shows that the electrochemical dissolution of Tc–Ru alloys leads to uniform corrosion of their surface.

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.