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Thermal fatigue behaviour of the ferritic–martensitic steels MANET-II, 12Cr–1.5NiMo and F82H-mod. have been investigated in the temperature range from 50°C to 350°C and total strain range ⩽0.33%. Crack appearance has been checked after 3×103, 6×103 and 104 cycles and has been successively detected in these steels. The thermal fatigue cracks have a transgranular character; sometimes, intergranular cracks are observed in the F82H-mod. steel. A certain correlation of grain size and ferrite content with the thermal fatigue crack peculiarities has been noted. Specimens of MANET-II and 12Cr–1.5NiMo have been irradiated in a WWR-M reactor with a fluence of 1×1025 n m−2 at a temperature of 300°C and then subjected to thermocyclic loading. It has been established that the neutron irradiation does not significantly affect fatigue crack nucleation in both materials.

Effects of molybdenum replacement by tungsten in steels of the bainitic, martensitic, and martensitic–ferritic classes containing 2.5%, 8% and 11% Cr, respectively, were investigated. The phase composition and structure of the bainitic steels were varied by changing the cooling rates from the austenitization temperature (from values typical for normalization up to V=3.3 × 10−2°C/s) and then tempering. The steels were irradiated to a fluence of 4×10 23 n / m 2 (⩾0.5 MeV) at 270°C and to fluences of 1.3×1023 and 1.2×10 24 n / m 2 (⩾0.5 MeV) at 70°C. The 2.5Cr–1.4WV and 8Cr–1.5WV steels have shown lower values of the shifts in ductile–brittle transition temperature (DBTT) under irradiation in comparison with corresponding Cr–Mo steels. Radiation embrittlement at elevated irradiation temperature was lowest in bainitic 2.5Cr–1.4WV steel and martensitic–ferritic 11Cr–1.5WV steel. The positive effect of molybdenum replacement by tungsten at irradiation temperature ∼300°C is reversed at Tirr=70∘C.

Thermonuclear fusion reactors of the Tokamak-type will be unique power engineering plants to operate in thermocyclic mode only. Ferritic–martensitic stainless steels are prime candidate structural materials for test blankets of the ITER fusion reactor. Beyond the radiation damage, thermomechanical cyclic loading is considered as the most detrimental lifetime limiting phenomenon for the above structure. With a Russian and a German facility for thermal fatigue testing of neutron irradiated materials a cooperation has been undertaken. Ampule devices to irradiate specimens for postirradiation thermal fatigue tests have been developed by the Russian partner. The irradiation of these ampule devices loaded with specimens of ferritic–martensitic steels, like the European MANET-II, the Russian 05K12N2M and the Japanese Low Activation Material F82H-mod, in a WWR-M-type reactor just started. A description of the irradiation facility, the qualification of the ampule device and the modification of the German thermal fatigue facility will be presented.

The deformability and fracture appearance of neutron irradiated Ti–Zr alloy were investigated. It was shown that irradiation at 200–250°C resulted in a drastic decrease of uniform elongation. The material irradiated at 350°C and over was characterized by high deformability. The dimple ductile transgranular fracture was found to be typical for the examined material both in initial state and after the low temperature irradiation. Elements of ductile intergranular fracture were observed in the material after neutron irradiation at 400°C.

Hollow specimens of a Cr-steel have been subjected to triangle temperature cycles and different temperatures. Out-of-phase thermal fatigue lifetime is analysed and compared to isothermal fatigue data close to the mean cycle temperature. Segments associated with different temperature ranges and isothermal fatigue lines display in logarithmic scale only a slight slope difference. The thermomechanical fatigue degradation model suggests that the strain-related degradation per cycle is multiplied by a factor reflecting the particular effect of temperature oscillation.

Effects of neutron irradiation to fluence of 2.0 × 1024 n/m2 (E > 0.5 MeV) in temperature range 70–300°C on mechanical properties and structure of the experimental reduced-activation ferritic 0.1%C–(2.5–12)%Cr–(1–2)%W–(0.2–0.7)%V alloys were investigated. The steels were studied in different initial structural conditions obtained by changing the modes of heat treatments. Effect of neutron irradiation estimated by a shift in ductile-brittle transition temperature (ΔDBTT) and reduction of upper shelf energy (ΔUSE) highly depends on both irradiation condition and steel chemical composition and structure. For the steel with optimum chemical composition (9Cr–1.5WV) after irradiation to 2 × 1024 n/m2 (E ⩾ 0.5 MeV) at 280°C the ΔDBTT does not exceed 25°C. The shift in DBTT increased from 35°C to 110°C for the 8Cr–1.5WV steel at a decrease in irradiation temperature from 300°C to 70°C. The CCT diagrams are presented for several reduced-activated steels.

Thermocyclic behaviour of austenitic materials: 16%Cr11%Ni, 16%Cr15%Ni, 20%Cr45%Ni has been investigated after being irradiated in WWR and BOR-60 reactors at a temperature of about 300°C up to fluences of approximately 10 25 and 10 26 n/m 2 . Strength and strain characteristics of irradiated materials are recovered during thermocycling (50–300°C) and reached initial unirradiated values. Simultaneously to the recovery processes an intensive accumulation of thermocyclic damage takes place with early material fracture. Thermal fatigue failure has shown an intergranular mode. It is assumed that two opposite processes take place during thermocycling: recovery in grain volume and the accumulation of damage along grain boundaries, which do not compensate each other.

The interaction of P impurity atoms with primary radiation defects is investigated by the molecular dynamics method with Morse-like potentials. The substitution of a P atom has a considerable region in which a self-interstitial dumb-bell can be captured, and this event can result in the P transition from the low mobile substitution state into a mixed P-Fe dumb-bell. The latter may rearrange into an energetically more efficient tetrahedral interstitial. The above reactions promote an essential growth of P mobility under low-temperature irradiation, when the vacancy mechanism of the impurity migration is not valid.

Transmission electron microscopy was used to compare the microstructure and particle distributions of two commercial oxide dispersion-strengthened copper alloys, GlidCop A125 and MAGT 0.2. Measurements were made on specimens in their as-wrought condition, after thermal annealing for l h at 900°C, and after 3 MeV Ar + -ion irradiation at 180 and 350°C to damage levels of 20 to 30 displacements per atom (dpa). All of the annealed and ion-irradiated specimens were found to be resistant to recrystallization. In addition, void formation was not observed in any of the irradiated specimens. The GlidCop oxide particle geometry was transformed from triangular platelets to circular disks by the ion irradiation. The MAGT particle geometry consisted of circular disks and spheres before and after irradiation. The oxide particle edge length in the unirradiated GlidCop alloy was about 10 nm, whereas the mean diameter of the particles in the unirradiated MAGT alloy was about 6 nm. After irradiation, the mean particle diameter in both the MAGT and GlidCop alloys was about 6 nm.

Tensile properties have been determined on specimens of four precipitation-hardened copper alloys irradiated in the SM-2 reactor between 100 and 500°C ( φ ≈ 10 21 n/cm 2 ). The effect of the irradiation temperature on the mechanical properties of the alloys has shown that some strengthening occur at low-temperature irradiations (100–200°C). Irradiations at 400–500°C result in the decrease of alloy strength. Growth and coagulation of secondary dispersed phases, accompanied by a decrease in their density, are proposed to cause the observed reduction in strength.

The effects of neutron irradiation on strength and ductility properties of precipitation hardened (PH) copper alloys are discussed. The analysis is based on the experimental study of radiation damage of PH alloys irradiated in the mixed spectrum reactor SM-2 to fluences of 3.7–5.5 × 1025 n/m2 (E > 0.1 MeV), corresponding to NRT displacement dose levels of 2.6–3.8 dpa. At irradiation temperatures of 100–285°C the processes of radiation hardening and reduction in the uniform elongation are the major effects. Irradiation at temperatures higher than 300°C causes a dramatic softening and improvement in uniform elongation of the CuCrZr and CuCrZrMg alloys. The threshold softening temperature for the PH alloys is shown to be about 300°C at a dose of 4.5 × 1025 n/m2 (E > 0.1 MeV). The effect of the irradiation dose and temperature on the shift of the threshold temperature of PH copper-alloys softening is also considered.

The paper presents the results of studies of copper alloys, conducted in the USSR to support ITER design. The following materials were investigated (precipitation-hardened (PH) alloys Cu-Be, Cu-Cr-Zr, Cu-Cr-Zr-Mg and dispersion-strengthened (DS) alloys Cu-Mo, Cu-Al 2 O 3 (MAGT-0.2 and MAGT-0.05). Materials were irradiated under the following conditions: reactor BOR-60: Ft ~ 10 22 n/cm 2 , T irr = 350–450°C; reactor SM-2: Ft ~ l−5 × 10 21 n/cm 2 , T irr = 100–500°C, where Ft is fluence (flux × time) and T irr is the irradiation temperature. Electrical resistivity, swelling, strength, ductility, fatigue, creep and corrosion in water have been studied. The investigation of the effect of the irradiation temperature on mechanical properties of copper alloys has shown that at low-temperature irradiation (100–200°C) some strengthening and slight change in ductility occur. Irradiation at 350–500°C results in decrease of PH-alloy strength. DS alloys maintain a high level of strength characteristics under all irradiation conditions.

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.

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.

The structural materials for the ITER (International Thermonuclear Experimental Reactor) divertor must perform reliably under complex and diverse operating requirements. Only a limited number of materials offer a potential for meeting these requirements for the wide temperature range of interest. The candidate materials considered in the ITER design activity include copper, molybdenum, and niobium alloys. Molybdenum alloys being considered include dilute alloys of the TZM type and the Mo-Re system. Niobium alloys under consideration include Nb-V-Zr and Nb-Zr systems. Copper alloys being considered include dispersion strengthened alloys of the Glidcop and MAGT type, alloys of Cu-Mo system and precipitation strengthened bronzes. The projected operating conditions for the ITER divertor and the criteria for evaluating the candidate materials are reviewed. This paper summarizes the data base and presents recent experimental results on these candidate divertor structural alloys.

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.

The microstructural evolution and swelling of pure copper, a precipitation-hardened alloy Cu-Cr-Zr-Mg, and the dispersion strengthened (DS) alloys Cu-Mo, MAGT-0.2 and MAGT-0.05 have been investigated after ion irradiation to doses of 3–150 dpa at irradiation temperatures of 260–600°C. It was shown that for pure copper the maximum swelling temperature was approximately 500°C and the swelling rate was 0.3–0.5% (dpa) −1 . In the Cu-Cr-Zr-Mg alloy voids have not been found for doses up to 150 dpa, but great changes in dislocation structure and phase composition have been observed. In DS-type alloys void formation has been found to depend on the dispersion particle size and concentration. In MAGT-0,2 alloy voids have not been found up to a dose of 150 dpa. However in MAGT-0.05 alloy voids have been observed at 100 dpa. Irradiation produces a redistribution of the dislocation structure and subgrain formation.

The effects of neutron irradiation on fine structure and mechanical properties of the precipitation-hardened alloy Cu-Cr-Zr-Mg and dispersion-strengthened alloys Cu-Mo and MAGT-0.2 have been investigated, Irradiations were performed in the BOR-60 ( T irr = 420°C, 1.5 × 10 22 n/cm 2 ), SM-2 ( T irr = 300°C, 10 21 n/cm 2 ) and WWR ( T irr = 80°C, 9.3 × 10 19 n/cm 2 ) reactors. It was found that the precipitation-hardened alloy Cu-Cr-Zr-Mg exhibits high resistance to swelling. At T irr ≈ 420°C, however, polygonization and precipitate coarsening processes go on at a high rate, which results in a drastic decrease in strength. Dispersion-strengthened alloys Cu-Mo and MAGT-0.2 exibit high resistance to radiation swelling and have high stability of strengthening structure during neutron irradiation over the temperature and dose range corresponding to ITER physics and technological stages.