A.M. Kryukov

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.

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.

Assessment of irradiation response of WWER-440 welds using samples taken from Novovoronezh unit 3 and 4 reactor pressure vessels

Abstract The results of the study on Novovoronezh unit 3 and 4 (NV NPP-3 and 4) reactor pressure vessel (RPV) radiation embrittlement measured using subsize impact specimens (5×5×27.5 mm 3 ) fabricated from samples taken from the corresponding RPV walls are presented. The post-irradiation annealing effectiveness and the embrittlement kinetics of Novovoronezh unit 3 and 4 RPV welds under re-irradiation are discussed. Ductile-to-brittle transition temperatures (DBTT) obtained using standard Charpy (TT 10×10 ) and subsize impact (TT 5×5 ) specimens of trepans cut out from Novovoronezh unit 2 RPV are compared. A new relation between TT 10×10 and TT 5×5 has been developed.

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.

Basic results of the Russian WWER-1000 surveillance program

The surveillance test results of the reactor pressure vessels (RPV) of three Russian WWER-1000 type units designated Units 1, 2 and 3 are given and the embrittlement rates compared to those predicted by the Russian Regulatory Guide. The surveillance materials properties measured by manufacturers of the RPVs are reviewed. The chemical compositions indicate low impurity contents (copper and phosphorus) but nickel contents up to 1.9 wt.% in some welds. The Charpy test results were available for the surveillance base and weld metals and the heat-affected zone (HAZ) of the three units. Dependence of the radiation behavior of WWER-1000 RPV steels on metallurgical variables and the damage dose is considered. The trend curves for the steels under investigation are proposed.

The properties of WWER-1000 type materials obtained on the basis of a surveillance program

The surveillance test results of the reactor pressure vessels (RPV) of three Russian WWER-1000 units designated unit-1, -2 and -3 are given and the embrittlement rates compared to those predicted by the Russian Regulatory Guide. Dependence of the radiation behavior of WWER-1000 type RPV steels on metallurgical variables and the damage dose is considered. The trend curves for the steels under investigation are proposed.

The properties of WWER-440 type reactor pressure vessel steels cut out from operated units

This paper presents the results of study on radiation degradation occurring in WWER-440 reactor pressure vessel (RPV) steel, using subsize impact specimens (5×5×27.5 mm3). The results of testing trepans and templates cut out from WWER-440 reactor pressure vessels are considered. Ductile-to-brittle transition temperatures (DBTT) obtained using standard Charpy and subsize impact specimens are compared. The relation between these two values is established.