V.P. Chakin

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.

Evolution of beryllium microstructure under high-dose neutron irradiation

Abstract Beryllium is an advantageous material from the point of its usage in some components of perspective fusion reactors, such as reactor wall, divertor and blanket. The character of the microstructure change under neutron irradiation of beryllium is a determining factor in understanding of the processes resulting in the degradation of physical–mechanical properties of the material. The performed examinations of the TE-56 beryllium grade irradiated in the SM-reactor at 70–120 °C up to fluences of 2.5×10 22 to 5.7×10 22 cm −2 ( E >0.1 MeV) and TE-400 beryllium grade irradiated in the BOR-60 reactor at 400 °C up to a fluence of 1.6×10 23 cm −2 ( E >0.1 MeV) demonstrated that low-temperature irradiation resulted in the generation of dislocation loops and high-temperature irradiation resulted in the generation of plane hexahedral voids in a basal plane. The effect of short-term high-temperature annealing on the microstructure after low-temperature irradiation is studied also. One hour anneal at 300–1200 °C gives rise to loops and results in their evolution in dislocation network as well as generation and growing of the gas bubbles.

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.

Influence of high dose neutron irradiation on thermal conductivity of beryllium

Abstract The thermal conductivity of beryllium is a very important physical characteristic, which considerably determines the serviceability of fusion reactor components where this material is supposed to be used. The neutron irradiation leads to a decrease of the thermal conductivity, the quantitative characteristic of the effect considerably depending upon the irradiation temperature and dose. So, after the irradiation in the SM reactor at a temperature of 70 °C up to neutron fluence of (1–6)×10 22 cm −2 ( E >0.1 MeV) the effect reaches hundreds of percents, after irradiation in the BOR-60 reactor at a temperature of 400 °C up to fluence of 1.6×10 23 cm −2 ( E >0.1 MeV) the effect reaches only tens of percents. One hour annealing of beryllium at 500 °C after the low temperature irradiation leads to a partial thermal conductivity recovery.

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.

Effects of neutron irradiation on physical and mechanical properties of Mo-Re alloys

The effects of neutron irradiation on mechanical strength, plasticity and electric conductivity of Mo-Re alloys are presented. Pure Mo and Mo-0.5%Re to Mo-30%Re alloys were studied. Samples were irradiated in the reactor SM-2 ( ~ 5 × 1021 n/cm2, Tirr ~ 240–330°C). Irradiation at 300°C indicates that the radiation embrittlement results in reduced strength and ductility characteristics of Mo-Re alloys. Measurement of electric resistance shows that low-temperature irradiation of Mo-alloys results in 20–30% reduced conductivity. SEM studies of fracture surfaces show that after irradiation brittle intergranular fracture takes places that correlates with an abrupt reduction of strength.

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.

Radiation damage features V2.5Zr0.35C alloy welds

Abstract We have carried out irradiation studies of base metal and welds of V2.5Zr0.35C alloy in the BOR-60 reactor in a 7Li environment at temperatures of 540 to 820°C for doses of 43–60 dpa. The welding was made by tungsten electrodes under Ar or He atmosphere. The welding zone is in the middle of the flat tensile specimens. The rupture of the specimens was in the base metal both before and after the irradiation while the radiation damage character of the welding zone differed drastically from the base metal. With the help of TEM investigation it was shown that during the irradiation, intensive precipitation of the fine phase occurred. This phase was of cubic type, presumably VC and/or ZrC, less than 5 nm in size and of 3 · 1016 cm−3 density. This should lead to significant strengthening and embrittlement of the welds under irradiation.