Yu. A. Nikolaev

Assessment of relative contributions from different mechanisms to radiation embrittlement of reactor pressure vessel steels

Abstract Experimental data on radiation embrittlement in pressure vessel steels of both Russian and American grades, obtained by the authors and also taken from the literature, have been analyzed to assess the relative contributions from the following mechanisms: radiation-induced hardening, inter- and intragranular segregation of impurities at precipitate/matrix interfaces. It is demonstrated that radiation-induced intragranular segregation of impurities frequently provides a significant contribution to radiation embrittlement of pressure vessel steels.

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.

Radiation Embrittlement of Cr–Ni–Mo and Cr–Mo RPV Steels

Radiation embrittlement kinetics of Cr–Mo and Cr–Ni–Mo steels, used for manufacturing VVER-440 and VVER-1000 RPV steels, respectively, are under consideration. Radiation embrittlement of VVER-440 and VVER-1000 RPV steels exposed to irradiation for 17–22 years is analyzed. Extremely high levels of thermal aging of VVER-1000 RPV steels at 320°C is pointed out. The neutron flux effect problem and the problem of accelerated radiation embrittlement of welds with high nickel contents are emphasized. The effect of phosphorus, copper, silicon, manganese, and nickel contents on radiation embrittlement of VVER-440 and VVER-1000 RPV steels is studied using the Russian surveillance database and representative results of research programs. The trend curves for VVER-440 and VVER-1000 RPV steels are proposed. The mechanisms of radiation embrittlement of Cr–Ni–Mo and Cr–Mo RPV steels are supposed.

Radiation embrittlement kinetics of the first generation of VVER-440 RVPs after post-irradiation annealing

Abstract A new approach for estimation of re-irradiation kinetics of VVER-440 reactor pressure vessel steels is presented. Effect of gamma flux on re-irradiation kinetics is discussed. Comparision of predicted and observed values of re-irradiation ductile-to-brittle transition termperature shifts for VVER-440 reactor pressure vessel core welds is provided.

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.