E. A. Krasikov

Intergranular and intragranular phosphorus segregation in Russian pressure vessel steels due to neutron irradiation

Russian reactor pressure vessel steels have been studied in three conditions: initial, irradiated and annealed. It has been established that irradiation induces both intergranular as well as intragranular phosphorus segregation. Fractographic studies demonstrated that brittle intergranular and ductile intergranular fracture surfaces of Charpy specimens appear as a result of intergranular and intragranular segregation, respectively. Transmission electron microscope (TEM) studies have revealed radiation-induced precipitates on interface boundaries to which intragranular phosphorus segregation occurs. Auger electron spectroscopy (AES) has been applied to detect phosphorus enrichment of fracture surfaces in the regions of brittle and ductile intergranular fractures.

Hydrogen-irradiated steel interaction during alternating hydrogenation and annealing

Abstract At present when notable progress in plasma physics has been obtained, practical energy utilization from fusion reactors is determined by the state of material science problems. The last includes not only the routine problems of nuclear engineering but also a number of entirely new problems connected with extreme conditions of materials operation – irradiation environment, hydrogenation, thermocycling, etc. Limited data suggest that the combined effect of these factors is more severe than any one of them alone. To study the possible significance of in-service synergistic phenomena, we studied hydrogen–irradiated steel interaction during alternating hydrogenation and heat treatment (annealing). These studies suggest the existence of residual (unrecovered by annealing) degradation effects caused by hydrogen–irradiated steel interaction during alternating hydrogenation and annealing. We propose that this phenomenon must be taken into account when in-service properties of the candidate structural materials are analyzed.

EFFECTIVE AVERAGE INTEGRATED γ-RAY FLUX STRIKING RADIATION DEFECTS DURING IRRADIATION IN A REACTOR

The results of experiments on the irradiation of diamond which attest to the fruitfulness of the zone theory of damage to materials are presented. The theory is used to derive an expression for the effective average integrated γ-ray flux striking radiation defects during irradiation in a reactor. It is shown by statistical analysis of the results on the critical neutron fluence for reactor graphite that when the average γ-ray flux is taken into account, other conditions being equal, the statistical error in the estimate of the critical neutron fluence for graphite decreases by a factor of 1.5. This shows that the effective average integrated γ-ray flux is another important parameter of the irradiation conditions, and this expands the physical meaning of the radiation composition factor. 3 figures, 1 table, 9 references.

Influence of isothermal and cyclic annealing on structure and swelling of neutron-irradiated beryllium

The influence of cyclic and isothermal annealing on structure and swelling of neutron-irradiated beryllium was examined. At cyclic annealing the base temperature was equal to 150°C, the maximum cycle temperature was 600°C (700°C in the second experimental set), the number of heating-cooling cycles was from 10 to 100. Isothermal annealing of beryllium was performed at 600°C, 700°C and 800°C. The annealing time in the isothermal regime was the same as well as the total annealing time of beryllium samples at appropriate temperatures in the cyclic regime. It was shown that there are no significant differences in the influence of cyclic and isothermal annealing on structure and swelling of irradiated beryllium.

Development of miniaturized specimens for the study of neutron irradiation/plasma exposure synergistic effects on candidate fusion reactor materials

Abstract The aim of this work is to choose a miniaturized specimen version relevant for testing candidate fusion reactor materials including mechanical testing after combined neutron irradiation/plasma exposure in a fission reactor. The material examined was reactor pressure vessel type steel in irradiated and aged (unirradiated) conditions. Comparative standard impact, three point bend and small punch tests were conducted. It is established that there is a possibility of miniaturization of irradiated steel experimental specimens by means of proper specimens type choice with mass reducing from ≈ 40 (Charpy) to 0.4 g (small plates).

Distribution of properties in nuclear reactor vessel shells in the unirradiated state

The distributions of the chemical composition, the strength characteristics, and critical ductile-brittle transition temperature Tcr are studied in the axial, radial, and tangential directions of the material of a test ring cut from a standard forging used for a VVER-1000 reactor vessel shell. The values of Tcr of specimens cut from the test ring are shown to be well below those of the internal volume of the shell, which can explain the substantial scatter of the results obtained on reference specimens cut from the base metal.