V.V. Bryk

Microstructural evolution and radiation stability of steels and alloys

Abstract Results of a systematic investigation of structure–phase transformations in the main FSU (Former Soviet Union) construction steels and alloys of ferritic and austenitic classes irradiated in reactors and in heavy ion accelerators are presented. Features in the dislocation structure evolution for these steels related to differences in stacking fault energy are considered and the role of cold deformation in swelling behaviour is investigated. The mechanisms of infiltration and compulsory alloying of second phase precipitates during irradiation as a result of mutual recombination at structural defects in interphase boundaries are discussed. The role of second phase formation and evolution on swelling behavior processes is investigated. The influence of undersized and oversized alloying elements (B, Si, Ti, Nb, Mo, Sc) on different aspects of structure–phase transformations, which are generally defined as radiation stability over a wide interval of irradiation temperature and dose, is considered.

On the effect of radiation-induced segregation on void shape and growth rate

Abstract Microstructural changes have been studied in Cr18Ni10Ti steel irradiated in BOR-60 up to 40 dpa at 580–600 °C. It was found that in the vicinity of the voids associated with G-phase particles the composition of steel is practically the same as the average matrix one, whereas the isolated voids are surrounded by Ni and Si-enriched zones. The shape and size of a void depend on local composition near it — the mean size of octahedral voids (associated with G-phase particles) is greater than that of isolated cubic voids. A theoretical model is developed to explain the observed difference between growth rates of free and precipitate-attached voids.

Microstructure investigation of Cr and Cr alloys irradiated with heavy ions

Abstract The results of microstructure investigation of pure Cr and low- and high-doped Cr alloys (Cr with low additions of La, Ta, Re; Cr doped with low additions of La, Ta, V; Cr alloyed with low additions of Fe, La; and 66Fe33CrAl, 33Fe66CrAl) irradiated by Cr ions in the temperature range 550 to 800°C up to dose of 180 dpa are presented. Dislocation structure evolution in low-doped Cr alloys differs from that in α-Fe alloys: loop faulting disappears in the beginning of irradiation; perfect loops have mainly Burgers vector ( a /3)〈112〉 on {111} planes and appreciably less have Burgers vector ( a /2)〈111〉 on {110} planes. High stability for dislocation loops in Cr alloys irradiated to high dose levels was discovered. The temperature and dose dependences obtained allowed evaluation of swelling resistance in Cr alloys. It was established that the swelling behavior is similar to that for other refractory metals having a bcc lattice. The maximum value of swelling is 9% at an irradiation dose level of 180 dpa in the temperature range from 730 to 750°C. Voids in Cr alloys are formed at dose levels of 1–2 dpa, similar to other refractory metals having a bcc lattice. The voidage is characterized by a homogeneous distributions of fine voids. It was established that a preliminary heat treatment, CW or alloying with low additions do not have significant effects on either the void nucleation process or swelling response. Void lattice formation was observed in Cr alloyed with low additions of La, Ta and V. The homogeneous distribution of fine voids through the matrix may be the cause of loss in ductility of the material. Highly-alloyed Cr alloys undergo intensive solid solution decomposition leading to heterogeneous swelling.

Simulation of radiation effects on reflectors using heavy ion beams

Abstract The results on modelling the behaviour of metal mirrors under neutron irradiation in a fusion reactor are presented. Mirror-like specimens were irradiated with beams of ions: Cr + (stainless steel), Cu + (copper) and Ne + and/or He + (beryllium). The energy of ions was in the range of a few keV (He + ) to 3 MeV (Cr + and Cu + ). The reflectivity as a function of irradiation dose was measured in the interval of spectrum 0.2–1.0 μm. There was no strong degradation of the mirror reflectivity up to doses ∼ 10 dpa in a near surface layer. In contrast, all used methods showed definite change of surface characteristics.

Problem of Radiation Resistance of Structural Materials of Nuclear Power

Mechanisms of radiation damage of construction materials in nuclear engineering and progress in developing radiation-resistant materials for the present and future generation nuclear reactors are reported. The analysis of the present state and of the perspectives toward a solution of the problem show that, in spite of the considerable research efforts throughout the world, economically acceptable operation standards for presently operating reactors still have not been reached. The reason is insufficient radiation resistance of the basic ingredients of the existing nuclear devices — of various classes of stainless steel and zirconium alloys. A key problem in the material sciences provision of the modern and future nuclear engineering is the study of the microstructure evolution and its impact on the degradation of the input physical and chemical characteristics.