A. V. Nikolaeva

The contribution of grain boundary effects to low-alloy steel irradiation embrittlement

Abstract The contribution of irradiation-induced enrichment of grain boundaries by impurities to irradiation embrittlement of reactor pressure vessel materials is discussed. Possible mechanisms of impurities and the effect of alloying elements on irradiation embrittlement of reactor pressure vessel steels are considered. Nickel has been found to influence greatly the tendency to irradiation embrittlement of nickel-containing steels with Ni wt% > 0.9. Irradiation resistance of nickel-containing steels has been shown to decrease significantly with the increase of silicon concentration from 0.24–0.28 to 0.3–0.4 wt%. The model for irradiation-induced enrichment of grain boundaries by impurities is used in order to explain the effect of silicon and nickel on irradiation embrittlement. In terms of the model, Si and Ni themselves do not prove the embrittlement, but they only influence thermodynamic and kinetic parameters of the phosphorus gain boundary adsorption. The embrittlement process itself is a result of decreasing of grain boundary cohesion with formation of phosphorus irradiation-induced grain boundary segregation.

Radiation embrittlement and thermal annealing behavior of CrNiMo reactor pressure vessel materials

The last generation of Russian type of reactor vessels (WWER-1000) is made of low alloy chromium-nickel-molybdenum steel. In order to study the radiation behavior of that steel, fourteen different materials, i.e. eight base metals and six weld metals, have been irradiated to different fluences at 290°C. The results of the corresponding Charpy V-notch impact tests are represented in this article. Some results of tensile tests are also given. Emphasis is given to the roles of metallurgical variables and dose effect. The results indicate anomalous dose dependence of irradiation-induced impact transition temperature shift. The corresponding trend curve has been proposed. Some of the irradiated materials have been subsequently annealed. It has been shown that the restoration effectiveness of anneal increases with increasing annealing temperature from 400 to 490°C, and nickel enhances residual shift after postirradiation annealing at 460°C.

Radiation embrittlement of low-alloy steels

Abstract Results of phosphorus, copper and nickel effect on radiation induced yield stress increase and ductile-to-brittle transition temperature (DBTT) shift are presented. The synergetic interaction between phosphorus and nickel is observed. The results of Russian VVER-440 and VVER-1000 surveillance programs and results of research programs on reactor pressure vessel (RPV) steel irradiation in surveillance channels of power reactors are discussed. The basic regularities of VVER-440 and VVER-1000 RPV steel are discussed. Trend curves for VVER-440 and VVER-1000 RPV steels are developed. The annealing effectiveness for VVER-440 and VVER-1000 RPV steel grades was compared. DBTT recovery of VVER-1000 RPV steels was found to be much lower than for VVER-440 RPV steels. Nickel was supposed to increase the post-irradiation residual DBTT shift of VVER-1000 type steels. Models for prediction of the post-irradiation residual DBTT shift of VVER-440 and VVER-1000 type steels were suggested.

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.

Behavior of mechanical properties of nickel-alloyed reactor pressure vessel steel under neutron irradiation and post-irradiation annealing

Abstract The effect of neutron irradiation and post-irradiation thermal annealing on tensile and impact properties of Cr–Ni–Mo steel used for WWER-1000 reactor pressure vessel (RPV) manufacturing was studied. A gap in yield stress and ultimate tensile stress fluence dependence at the fluence range of 0–3×1023 neutrons m−2 was observed while ductile-to-brittle transition temperature (DBTT) was continuously increasing with damage dose. The post-irradiation annealing recovery of tensile properties was found to be higher than the one of impact properties. Over-recovery of tensile properties due to 460 and 490°C post-irradiation annealings were observed. The annealing effectiveness of WWER-440 and WWER-1000 grades was compared. Nickel was supposed to affect both the radiation sensitivity and the post-irradiation residual DBTT shift of WWER-1000 type steel.

Grain boundary embrittlement due to reactor pressure vessel annealing

Abstract The tendency of the Cr-Ni-Mo low-alloyed steel to brittle fracture as a function of the sizes of austenitic grains and the phosphorus concentration at the grain boundary has been studied. A simple analytical dependence connecting the temperature of ductile-to-brittle transition of steel with the boundary phosphorus concentration and the austenitic grain size has been found. In estimating the kinetics of the development of intergranular embrittlement the decrease in the diffusion coefficient of phosphorus in α-Fe in the presence of molybdenum was taken into account. The effect of the mutually increasing grain boundary adsorption of phosphorus and nickel was considered as well. The possibility to predict the tendency of the Cr-Ni-Mo low-alloyed steel to temper embrittlement is shown. The technique proposed was successfully used to estimate the degree of recovery of the Ni-containing materials of the nuclear reactor vessels after annealing radiation defects.

Mechanism of the drop in the dependence of yield stress on neutron irradiation dose for low-alloy steel

In the present work, influence of irradiation on tensile and impact properties of Cr-Ni-Mo steel using for WWER-1000 type reactor pressure vessel (RPV) manufacturing was studied. An abnormal behavior of tensile properties of the investigated materials under irradiation was observed. Decrease in yield stress and ultimate tensile stress at the first stage of irradiation was revealed while ductile-to-brittle transition temperature (DBTT) was continuously increasing with damage dose. A model of the phenomenon considering influence of residual impurities and alloying elements on α-Fe lattice tetragonality under irradiation was proposed. Some experimental evidences of the model were found. Nickel was supposed to govern the radiation sensitivity of WWER-1000 type steel.

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.