This paper presents the results of atom probe tomography studies on radiation-induced phase formation in light-water reactor pressure vessel steels after neutron irradiation under various conditions in comparison with the literature data. The given irradiation conditions are fluence (10–100) × 1022 m−2, flux (5–2700) × 1014 m−2s−1 and irradiation temperature (50–400) °C. The composition of the studied steels varies in a wide range for the elements significantly affecting radiation and thermal resistance of steels: Ni in the range of 0.2–6.0 wt.%, Mn–0.03–1.1 wt.%, Cu–0.01–0.16 wt.%, and P–0.01–0.03 wt.%. The number density, volume fraction, size, composition, and nucleation sites of precipitates are determined. The regularities of the effect of various operational factors on the phase formation in these steels have been analyzed and revealed. The study shows that in materials with high copper content, Cu-rich precipitates are formed by a radiation-enhanced mechanism. In materials with low copper content, their formation upon irradiation at 300 °C occurs by a radiation-induced mechanism since the main nucleation sites are point defect clusters formed in cascades. At the same time, the density, volume fraction, and composition of the precipitates depend on the steel composition (Ni and Mn content). In the steel with increased Ni content up to 5 wt.% but with ultra-low Mn content ≤ 0.03 wt.%, it is possible to suppress the formation of Ni-Si-Mn precipitates under irradiation.

The paper considers the results of structural studies and mechanical tests after a long-term thermal exposure of laboratory heats of the metallurgically improved 15Kh2NMFA steel and steel with an increased content of nickel considered as materials for the pressure vessels of advanced VVER-type reactors of various designs. It has been shown that, both for the improved 15Kh2NMFA steel and the high-nickel steel, there are no signs of grain boundary embrittlement after an segregation provoking embrittlement heat treatment. This is explained by the extremely low grain boundary segregation of phosphorus in the initial state caused by a high degree of the structure dispersity as well as by rather a low content of impurities. Besides, no changes have been found in the yield strength value for the improved 15Kh2NMFA steel, which agrees with the structure investigation results. For the high-nickel steel, a tendency towards a minor yield strength decrease by 5 to 10% and a regular reduction of the critical brittleness temperature has been revealed. A decrease in the mechanical properties has been caused by a relatively low temperature of tempering for the high-nickel steel and, accordingly, by the potential occurrence of the structure recovery during long-term thermal exposure, as evidenced by the results of an X-ray diffraction analysis. Despite the structure recovery in the high-nickel steel under the long-term thermal exposure, the main strengthening carbide phases remain stable. Due to this, the yield strength value remains at a relatively high level that exceeds the values for the modern VVER-type vessel steels, even in the case of a thermal exposure much in excess of the expected operating conditions for advanced VVER reactors. The observed decrease of critical brittleness temperature during the long-term thermal exposure contributes to an increase in the steel resistance to brittle fracture.

In this paper the phase formation and mechanical properties of VVER-type reactor pressure vessel (RPV) steels with various Ni (1.57–5.95 wt.%) and Mn (0.03–0.76 wt.%) content after neutron irradiation up to fluences in the range of (53–120) × 1022 n/m2 at 400 °C were studied. The possibility of carbonitride formation under these irradiation conditions is shown. In case of sufficient Ni (>1.5 wt.%) and Mn (>0.3 wt.%) content formation of Ni-Si-Mn precipitates is observed. Their chemical composition is close to G-phase and Γ2-phase and differs from that of radiation-induced precipitates in VVER-1000 RPV steels. This indicates the prerequisites for thermally conditioned mechanism of Ni-Si-Mn precipitates formation and growth at 400 °C enhanced by irradiation. It is also shown that the optimized steel manufacturing technology coupled with an ultralow Mn content (≤0.03 wt.%) in steel with increased up to 5.26 wt.% Ni content facilitates suppressing the Ni-Si-Mn precipitates and carbonitrides formation. This, in turn, reduces the contribution of the hardening embrittlement mechanism and, correspondingly, facilitates high radiation resistance of the steels with ultralow Mn content at the increased irradiation temperature (400 °C).

Comparative studies of the radiation-induced structure of austenitic steels with a nickel content of 10, 20 and 25 wt.%, irradiated sequentially in the SM-3 and BOR-60 reactors, as well as to higher damaging doses in the BOR-60 reactor, have been carried out. The phase composition, dislocation structure, pores, and radiation-induced segregations at grain boundaries were studied by high-resolution analytical methods of transmission electron microscopy, scanning electron microscopy, and atomic probe tomography. The formation of radiation-induced phase precipitates based on nickel has been established, and its volume fraction correlates with the level of radiation-induced segregations, and increases, the higher the nickel content in the steel. The values of barrier strength factors for radiation-induced structural elements in the studied steels are adjusted by calculation and experiment, which makes it possible to determine their contribution to radiation hardening.It is shown that the largest contribution to radiation hardening as a result of neutron irradiation in BOR-60 at high irradiation temperature up to 29 dpa is made by large radiation-induced precipitates of (G + γ') phases. It is shown that with an increase in the damaging dose, the main factor limiting the performance of internal devices will be radiation swelling, since the contribution to the change in properties from radiation-induced phases and radiation defects will not increase due to their density reaching saturation. Steel with 25 wt.% Ni exhibits the lowest level of swelling at high radiation doses, which makes it possible to consider it as a material-candidate for internals for promising VVER reactors with higher temperatures and longer service life.

Nickel is an essential alloying element in steels used as structural materials in the most common nuclear power reactors of the VVER type. The paper considers reviews the results of structural studies of traditional and advanced materials of the vessels and internals of VVER-type reactors with high nickel contents in their compositions. It is shown that an increased nickel content (up to 5 wt.%) in the steels of VVER pressure vessels contributes to the formation of a more dispersed structure with a smaller size of substructural elements and an increased density of dislocations, as well as a higher volume density of carbide phases. The revealed features of the structure of the reactor pressure vessel steel with high nickel content have the prerequisites for improving the strength and viscoplastic properties due to the increased number of barriers both for the dislocation motion and brittle crack propagation. Using the example of materials for VVER internals, it is shown that the nickel content increased in them up to 25 wt.% contributes to an increase in the volume density of radiation defects (dislocation loops of various types) and radiation-induced phase precipitates (G-phase). As nickel increases from 10 to 25 wt.%, there is a tendency to reduce swelling, which contributes to less shape change of the components of the reactor vessel internals. At the same time, in the steel with the highest nickel content, the highest nickel content was found in the near-boundary regions of the matrix, which contributes to greater austenite stability and a lower probability of the formation of an embrittling α-phase. The data obtained in the work on the effect of nickel alloying on the steel structural phase state and service characteristics were used in the development of new materials for the vessels and internals of advanced reactors.

The paper presents the results of structural studies of ring specimens made of the 42XNM alloy after irradiation as part of the control and protection system of the VVER-1000 reactor to a damaging dose of ~12 dpa at a temperature of ~350°C and subsequent isothermal annealings in the temperature range of 400– 1150°C (heating and holding for ~2 h). It is shown that during long-term isothermal annealing, a change in the phase composition of the alloy is observed, dislocation structures and grain-boundary segregations are annealed, and porosity evolves. It has been confirmed that decrease in the plastic properties of the 42XNM alloy after irradiation and subsequent isothermal annealing in the temperature range of 400–1000°C could be explained by the formation of precipitates of the second phases (zones of discontinuous decomposition of the solid solution with the release of α-Cr particles along the grain boundaries) and pores at the grain boundaries.

The paper presents the investigation results using TEM and SEM methods of alloy 42XNM after irradiation as a part of the absorbing element (the control rod cladding) of the VVER-1000 reactor control system (to the damaging dose of similar to 12 dpa at a temperature of similar to 350 degrees C) and after isothermal annealing in the temperature range of (500-1100)degrees C simulating LOCA-parameters and corresponding to a sharp decrease in plastic properties. It was shown that during such annealing, a change in the alloy's phase composition and porosity evolution was observed, dislocation loops and grain boundary segregations were completely annealed. The probable reasons for the decrease in the plasticity of the 42XNM alloy in the specified temperature range were found: the formation of alpha-Cr particles along grain boundaries and the pore formation on interfacial and grain boundaries. (c) 2022 Elsevier B.V. All rights reserved.

The paper presents the results of electron microscopic studies of ferrite-martensitic steel samples hardened with Y-O oxides, EP-450 DUO in the initial state and after neutron irradiation in the BN-600 reactor at 1000°C to a damaging dose of 77.5 dpa. These studies showed that the main types of oxide phases were Y2(Si, Ti)2O7 and Y2(Si, Ti)O5. These precipitates at sizes less than 10-20 nm were semi-coherent with a ferritic matrix of steel EP-450 DUO with the ratio (110)malrix//(221)palticle. Some of the Y-Ti-O oxides in the initial state were Y2Ti2O7-type with some deviations from the stoichiometric composition.However, after neutron irradiation under BN-600 conditions at temperature ~ 1000°C, oxide particles could not be described by the indicated stoichiometry. Besides, after irradiation, silicon and aluminum were found in the oxide's composition. In the case of taking these elements into account during the construction of a triple composition diagram, it was shown that the oxide phases had Y2(Ti, Si, Al)2O7 and Y2(Ti, Si, Al)O5 types. It was established that in samples of EP-450 DUO steel in the initial state with oxide particles up to 20 nm in size, the yttrium content is generally lower than the titanium concentration. The titanium and yttrium concentrations corresponded to the stoichiometric composition Y2Ti2O7 (1:1) with a further increase in the average diameter of these phases. After irradiation, the situation changed somewhat: the yttrium content in most oxide phases exceeds the total concentration of titanium, silicon, and aluminum.The paper also presents the analysis of porosity and evolution of grain structure in EP-450 DUO steel after neutron irradiation.

The influence of the grain structure and phase composition of welded seams of WWER-type reactors on their operational characteristics in the initial state is analyzed. Comprehensive studies of welds with various structures formed as a result of welding using various technologies have been carried out. It is shown that differences in the phase composition and morphology of the grain structure of the weld metal in the initial state, which is characterized by an increased proportion of columnar grains and a higher density of carbide phases, lead to an increase in the yield strength and critical brittleness temperature of welded joints.

Complex structure studies and mechanical tests of low carbon Ni–Cr–Mo–V steels with low and high nickel content (low-Ni and high-Ni steels) were performed. Mechanical tests showed that high-Ni steels in initial state are characterized by higher strength properties and lower values of critical temperature of brittleness in comparison with low-Ni steels. Estimation of input of various structural elements in ensuring yield strength and brittle fracture stress (which is connected with critical temperature of brittleness) was performed. It was shown that increased strength characteristics of high-Ni steels are determined mostly by higher density of nanoscale carbide precipitations and dislocations. It was also shown that lower critical temperature of brittleness of high-Ni steels is associated with lower width of substructural blocks and, consequently, with higher number of barriers to brittle transcrystalline crack propagation.

Comparative studies of porosity and calculation of the swelling profile in samples of austenitic stainless steel with a nickel content of 10 and 20 wt.%. The samples were irradiated at the Tandem-3M accelerator to the same doses of 300 dpa with Ni ions with an ion energy of 11.5 MeV at a temperature of 550°C with preliminary implantation of He. To calculate the swelling profile, digitally processed images were obtained by scanning transmission electron microscopy (STEM). In addition, comparative studies of the phase composition and radiation-induced segregations at grain boundaries, pore/matrix interfacial boundaries, and on the surface of phase precipitates were carried out on irradiated samples with varying nickel contents.

One of the key embrittlement mechanisms in reactor pressure vessel (RPV) steels is the hardening produced by nanometer features. In this paper low Cu high Ni WER-1000 RPV steels with a wide range of Ni (1.1-1.94wt.%) and Mn (0.38-1.1wt.%) contents irradiated at 290 degrees C to (6-101) 10(22) n/m(22) (E>0.5MeV) at both low (surveillance specimens) and high (test reactor) fluxes were analyzed using atom probe tomography and transmission electron microscopy. Formation of high number density of NMS precipitates with the average composition close to at.%: 45Ni-33Si-22Mn regardless of the irradiation conditions was observed. Its number density increases with the increase of the total Ni and Mn concentration in steel. High flux irradiation demonstrates lower size and volume fraction of NMS precipitates compared to the low flux irradiation in case of welds. For the base metal specimens with low Ni (1.1-1.3 wt.%) and Mn (0.38-0.51) content no flux effect was observed. The contribution of NMS phases and dislocation loops into radiation hardening of RPV steels was established by the dispersed barrier strength model (Orowan equation). (C) 2021 Elsevier B.V. All rights reserved.

The paper presents the results of microstructural studies by transmission electron microscopy and atom probe tomography of the Ni-Cr-Mo (42XNM) alloy in the initial state, as well as after neutron irradiation, in the range of fast neutron fluence (0.9–1.5)•1026n•m−2 (E>0.1 MeV) (5–12 dpa) at ~ 300°C. It was shown that during neutron irradiation, various defects were formed – dislocation loops and pores, the bulk density and size of the second phase precipitates (TiN and α-Cr) remained stable. Quantitative analysis of chemical composition for regions near the boundaries of austenitic grains was made, which showed typical effects for materials with an FCC lattice (which contains Ni and Cr) after neutron irradiation – nickel depletion of grain boundaries and enrichment in chromium. Similar studies were carried out for segregation processes on the Frank loop edges.

Reorientation of hydrides in the course of dry storage of spent nuclear fuel is a possible mechanism for the degradation of the properties of fuel claddings made of alloy E110 (Zr–1% Nb) based on spongy zirconium in VVER reactors. To determine the degree of reorientation of hydride precipitates in fuel claddings under conditions that simulate the conditions of dry storage, a series of tests have been performed for unirradiated samples with a hydrogen content of 100 and 200 wppm at various levels of circumferential stresses up to 100 MPa for 40 days. The orientation of hydride precipitates is determined using metallographic measurements. To assess the mechanical characteristics, the ring specimens have been tested for static uniaxial tension. The hydrides are analyzed using transmission electron microscopy and X-ray diffraction analysis. It is shown that test conditions lead to partial reorientation of hydride precipitates, which contributes, in turn, to a decrease in the plasticity.

The paper presents results of microstructural studies of E110 alloy specimens in fuel claddings based on sponge and electrolytic zirconium after operation in the fuel elements in VVER-1000. During the creep tests with axial loading no changes were observed in the studied specimens referring the chemical composition, average size and bulk density of the second phases, including radiation-induced ones. It was found that during creep tests, dislocation loops are annealed, i.e. an increase occurs in their average size with a  simultaneous decrease in bulk density. It was shown that the specimens of fuel elements claddings from an alloy based on electrolytic zirconium demonstrate greater creep resistance compared with sponge based zirconium specimens, which is apparently linked with a higher density of globular β-Nb precipitates in the irradiated electrolytic zirconium specimens.

TEM, SEM, and APT techniques have been used to analyze radiation-induced components of metal structure of fragments cut from the pressure vessel internals of Novovoronezh NPP Unit No 3 after 45 years of operation. The fragments differed in the neutron damaging doses (from 14 to 43 dpa) and the irradiation temperature (from 285 to 315°C). The density and dimensions of titanium carbides and carbonitrides, dislocation loops, radiation-induced voids, segregations, and nanoscale precipitates were determined. The contributions of structural components to the radiation hardening of the investigated fragments of 18Cr-10Ni-Ti stainless steel were estimated.