Yu. R. Kevorkyan

Grain-Boundary Segregation of Phosphorus in Low-Alloy Steel

Grain-boundary segregation of impurity elements, such as phosphorus, arsenic, antimony, and others, decreases the grain-boundary cohesion, which can substantially increase the temperature of the ductile-brittle transition in low-alloy structural steel. The most dangerous surface-active impurity for low-alloy steel employed for nuclear reactor vessels is phosphorus. A change of the cohesive strength of grain boundaries as a result of radiation-stimulated phosphorus segregation is considered to be one of the main mechanisms determining the radiation embrittlement of reactor-vessel materials. Since the mechanisms of embrittlement during development of reversible temper brittleness and radiation-stimulated grain-boundary segregation of phosphorus are the same, the main characteristics of the influence of the latter on the mechanical properties of steel can be determined by investigating steel treated in the range 400–600°C. The present investigation made it possible to develop a relation for determining the change in the temperature of the ductile-brittle transition in low-alloy steel as a result of the development of temper brittleness.

Radiation Embrittlement of VVÉR-1000 Vessel Materials

Radiation embrittlement of VVÉR-1000 vessel materials has been studied much less than for VVÉR-440 reactors. In the present paper the results of an investigation of the first batches of control samples of VVÉR-1000 vessel materials are discussed. The chemical composition of the materials is characterized by a low content of harmful impurities (copper and phosphorus) and a high nickel content (up to 1.9% in some weld seams). The actual rate of radiation embrittlement of the material studied is comparable to the embrittlement calculated using the Russian standards. The dependence of radiation embrittlement of VVÉR-1000 vessel materials on the metallurgical variables and the damaging dose is studied. The investigation showed that nickel greatly intensifies the radiation embrittlement. New relations were developed for determining the actual rate of radiation embrittlement of VVÉR-1000 reactor vessel materials and assessment of its conservativeness.

Embrittlement of low-alloy structural steel by neutron irradiation

The radiation embrittlement of reactor vessel materials is a complex process, which depends on the conditions of irradiation and the microstructure and chemical composition of the steel. It is universally acknowledged that phosphorus, copper, and nickel intensify the radiation embrittlement of vessel material the most. It is believed that Mn, N, C, Mo, Si, As, Sn, V, and other elements also influence radiation embrittlement, but their effect has not been definitely established and is much less than the effect of phosphorus, copper, and nickel. The presence of a synergetic interaction of elements in the irradiation process and the complex interaction of metallurgical factors and the irradiation conditions make it difficult to determine the degree to which impurities and alloying elements influence radiation embrittlement. The effect of the chemical composition of steel, as one of the most important parameters determining the radiation service life of vessel material, on radiation embrittlement is studied, 5 figures, 1 table, 20 references.

INFLUENCE OF RADIATION-STIMULATED GRAIN-BOUNDARY SEGREGATION OF PHOSPHORUS ON THE OPERATIONAL PROPERTIES OF NUCLEAR-REACTOR-VESSEL MATERIALS

The main mechanisms of radiation embrittlement of reactor vessel materials are considered to be hardening of material as a result of the formation of matrix defects, for example, micropores and second-phase precipitates – copper and others, and a change in the cohesive strength of grain boundaries as a result of the segregation of surface-active impurities, primarily, phosphorus. The question of the degree to which the latter mechanism affects the change in the properties of reactor-vessel materials under irradiation remains open. In the present paper, computational estimates of the kinetics of radiation-stimulated segregation of phosphorus on grain boundaries in reactor-vessel materials and the resulting changes in the mechanical characteristics of steel are compared with corresponding experimental data.

Experimental-statistical analysis of radiation embrittlement of VVÉR-440 vessel materials

An experimental-statistical approach, consisting in obtaining and then analyzing statistically an experimental database, is used as a basis for predicting radiation embrittlement of reactor vessel materials. A large experimental database on radiation embrittlement of materials irradiated in VVÉR-440 has now been accumulated. The purpose of the present paper is to analyze these data statistically, compare the data with currently employed standard relation, and obtain a correlation relation, which better describes these data, for weld-seam material and for the main metal of a vessel. A new, more reliable, standard relation is developed for these materials and recommended for use. 6 figures, 6 references.

Restoration of mechanical properties of irradiated steel by thermal annealing

During operation a reactor vessel is exposed to neutron irradiation fluxes, which causes degradation of the mechanical properties of the vessel material. This concerns primarily the change in the critical temperature of brittleness and other characteristics of the fracture viscosity of steel. Heating steel above the irradiation temperature, thereby increasing the diffusion mobility of point defects, is a prerequisite for the appearance of thermodynamic instability of various radiation damage of steel and, consequently, creates the conditions for restoration of the mechanical properties. In this paper the factors influencing the effectiveness of the restoration of the critical temperature of brittleness of materials of water-moderated water-cooled power reactor vessels with post-radiation annealing are examined.

On the role of replacement sequences in forming cascade region structures

Abstract A model of replacement sequence (RS) evolution in a crystal has been developed. The calculation results of the RS characteristics are compared with the data obtained by the molecular dynamics method. The estimations of the temperature effect on the RS mean lengths showed that this effect is small; at least up to 500 K.

Influence of cascade micropores on diffusion fluxes of point defects in reactor vessels materials

The main mechanism determining radiation embrittlement of the materials of reactor vessels are considered to be a change in the cohesive strength of the grain boundaries resulting from the segregation of surface-active impurities (mainly phosphorous), hardening of the material by precipitations of a second phase, segregation of impurities onto interphase surfaces of the precipitate-matrix interface and an associated increase in the hardening effect of the radiation precipitations. The materials of reactor vessels are irradiated at a temperature of 250–300°C which is insufficiently high for the thermal activation of diffusion processes. When describing the process of an accelerated diffusion of copper and phosphorous atoms by irradiation, a model has so far normally been used which neglects the cascade mechanism for generating defects and also the formation and evolution of point defect complexes. Individual studies taking account of the reduction in the efficiency of generating point defects in cascades due to the formation of micropores contain certain deficiencies. The present work is devoted to developing and analyzing a satisfactory model of radiation damage which is to a considerable extent free of these deficiencies.