[en] The present paper was devoted to investigation of the stress effect on swelling and microstructure evolution of the Fe-15.8Cr-15.3Ni-2.8Mo-0.6Nb steel irradiated in the BOR-60 reactor at temperatures from 395 to 410 deg. C and damage doses from 79 to 98 dpa. Was found out that the stress increase leads to an increase of swelling, that can be associated with a decrease in incubation period with a practically constant swelling rate. Voids concentration increases at the first stage of irradiation when the void sizes are practically constant, and then the concentration reaches some saturation and swelling increase is caused by void growth

[en] The necessity of prolongation of operating lifetime of the WWER reactors in Russia and Ukraine, PWR reactors in Europe, Japan and America evoked a great number of examinations of radiation phenomena in austenitic steels under irradiation conditions close to those of operation of power reactor in-vessel devices. This paper is devoted to analysis of the examination results of hardening and changing of microstructure characteristic Fe-18Cr-10Ni-Ti steel irradiated in the WWER-1000 reactors at relatively low irradiation temperatures

[en] At present, work in justification of the lifetime prolongation of the operated VVER-440 and VVER-1000 internals as well as of the operation of new VVER reactor internals up to 60 years is the most urgent. Fe-0.08C-18Cr-10Ni-Ti austenitic steel, being the material of operated and new VVER internals, was selected for the experiment. As the design of internals is very complicated and there are many holes for cooling, areas with compressive and tensile stresses may appear, so it is important to investigate the effect of stresses on the properties and structure of the material. Experiments to investigate the effect of tensile stress on the properties and structure of the material have been carried out both at 'SSC RIAR', Russia and abroad, but the effect of compressive stress has not been practically studied. Besides, we had to check if the known mechanisms and dependence of creep strain on stress type would remain. This paper presents the effect of compressive and tensile stresses on swelling, microstructure and creep strain of Fe-0.08C-18Cr-10Ni-Ti steel. It appears that: -) creep strain of the specimens is in the proportion to damage dose and tensile stress, and -) hardening induced by irradiation is the same for both stressed and non-stressed specimens

[en] Fuel cladding samples from OCr16Ni15Mo3Nb stainless steel have been examined after irradiation in the Reactor BOR-60 to the fluences 4.2 x 10²² and 6.6 x 10²² n/cm², E > or = 0.1 MeV. Voids, dislocation loops and precipitates are found. It is established that maximum swelling observed near approximately 500⁰C is equal to 6.5%. Intergranular M₂₃C6 precipitates and intragranular fine dispersed Nb(C,N) and Laves phase type Fe₂(Mo, Nb) precipitates due to radiation-induced diffusion are found

[en] 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 deg. C and doses of 50 dpa. Steel 0.1C-13Cr-Mo, V, Nb, B was chosen for comparison. At irradiation temperatures of 330-340 deg. 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 deg. 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 ε-bar/σ-bar=B₀+DS, when B₀ and D have the values typical for steels of FM type.

[en] Beryllium is one of the main candidate materials both for the neutron multiplier in a solid breeding blanket and for the plasma facing components. That is why its behaviour under the typical for fusion reactor loading, in particular, under the neutron irradiation is of a great importance. This paper presents mechanical properties, swelling and microstructure of six beryllium grades (DshG-200, TR-30, TshG-56, TRR, TE-30, TIP-30) fabricated by VNIINM, Russia and also one - (S-65) fabricated by Brush Wellman, USA. The average grain size of the beryllium grades varied from 8 to 25 μm, beryllium oxide content was 0.8-3.2 wt. %, initial tensile strength was 250-680 MPa. All the samples were irradiated in active zone of SM-3 reactor up to the fast neutron fluence (5.5-6.2) · 10²¹ cm^-2 (2.7-3.0 dpa, helium content up to 1150 appm), E > 0.1 MeV at two temperature ranges: T₁ = 130-180degC and T₂ = 650-700degC. After irradiation at 130-180degC no changes in samples dimensions were revealed. After irradiation at 650-700degC swelling of the materials was found to be in the range 0.1-2.1 %. Beryllium grades TR-30 and TRR, having the smallest grain size and highest beryllium oxide content, demonstrated minimal swelling, which was no more than 0.1 % at 650-700degC and fluence 5.5 · 10²¹ cm^-2. Tensile and compression test results and microstructure parameters measured before and after irradiation are also presented. (author)

[en] The understanding of processes happening in fuel element materials under irradiation is connected with application of the modern instruments and techniques providing the largest body of information on changes of chemical and isotopic composition, crystalline structure, micro- and macrodefects, phase composition and distribution. It is the practice study by sampling techniques providing the representativity of the results obtained. Last years at the FSC RIAR hot laboratory the technique of integrated analysis is extensively used for grain a large information body by the same sample. It is achieved due to application of mutually co-ordinated instruments providing for examination of irradiated specimens of the same shape and dimension, quick fulfillment of transport operations between the instruments, specimen surface being without contact with the air in certain cases. The detailed non-destructive and traditional destructive techniques make it possible to choose the most representative and informative part of an irradiated fuel element for sampling and examinations. As an example of the WWER fuel, fuel element cladding investigations it was shown the main advantage of the integrated microscopy application, Auger-spectroscopy of X-ray microanalysis for examination of fuel element cladding materials and corrosion effects observed in fuel and coolant, secondary ionic and X-ray microanalysis for study of fuel composition. (author). 5 figs, 1 tab

[en] Evidence has been provided to support the 'temperature-shift' concept of swelling. The swelling of annealed X18H10T stainless steel at PWR-relevant dpa rates was found to be higher than predicted by a swelling equation developed from data obtained at dpa rates an order of magnitude higher. These data were derived from identical hexagonal wrappers irradiated in both in-core and reflector positions. Based on these results, irradiation in PWRs would yield higher swelling to lower temperatures than would occur in core regions of fast reactors. (authors)

[en] The structural changes of structural elements of FA of WWER-1000 (claddings of fuel elements, process channels, central tube, stiffening angles), influence of mechanical stresses on the processes of nucleation and formation of radiation-induced fine-dispersed particles under irradiation are investigated

[en] The results of irradiation influence at 120-160 C up to 3.6-6.0 x 10²¹ n/cm² (E > 0.1 MeV) in the SM reactor on mechanical properties and microstructure of the electron-beam welds of Mo-Re alloys with a content of 15%, 20%, 30%, and 41% Re are presented in the paper. Severe radiation embrittlement of these materials due to formation of dislocation loops is observed. The welds of Mo-Re alloys with the higher Re content are comparatively less susceptible to the radiation embrittlement. The density of dislocation loops is reduced and the fracture type is changed from intergranular to transgranular with an increase of Re content in the alloys. There is no formation of dislocation loops at all in the weld-fusion zone of Mo-41Re alloy after the initial annealing at 1400 C for 1 h. No radiation-induced formation of the second phases is detected in the investigated Mo-Re alloys. (orig.)