P.A. Platonov

The cascade mechanism of nucleation of vacancy loops and stacking fault tetrahedra in FCC metals

Abstract The mechanism of nucleation of vacancy loops (VL) and stacking fault tetrahedra (SFT) in cascade regions of damages in copper were investigated by a molecular dynamic simulation. The nucleation of VL or SFT was shown to take place in depleted zones at cascade thermal stage for ~10 −11 s. The type of an incipient vacancy cluster (loop or SFT) depends on the volume concentration of vacancies in the depleted zone. The data obtained make it possible to explain qualitatively the experimental results on the microstructure of copper irradiated at a temperature of

Radiation damage of graphite and carbon-graphite materials

Abstract The results of the study of graphite and carbon-graphite materials obtained in the Russian Research Center “Kurchatov Institute” during a period of fifteen years are presented. The main structural effects and changes in pyrographites, nuclear (polycrystalline) graphites and carbon-graphite materials under irradiation are investigated. Irradiation temperatures and neutron fluences have been taken from wide ranges: ≈ 500–1200°C, and ≈ (0.01–2) × 10 26 n/m 2 respectively, ( E > 0.18 MeV). The following characteristics of materials are discussed: linear and volume changes in size, lattice parameters, radiation defect parameters, pore structure evolution, Youngs modulus change, etc. Particular attention is given to mechanisms, governing material radiation stability as a function of processing and irradiation conditions. The widely known effect of radiation dimensional change in irradiated graphite has been observed in the experiment. It has been demonstrated that the radiation defect parameters and magnitudes of radiation dimensional change are defined by initial sizes of graphite crystallites (for irradiation temperatures ≥ 500°C). It has been shown also that this effect determines the radiation stability in polycrystalline nuclear graphites. Distinctions in behavior of carbon-graphite materials under irradiation are shown.

Post-irradiation studies of beryllium reflector of fission reactor examination of gas release, swelling and structure of beryllium under annealing

Abstract Hot-pressed high-density (TShG-type) beryllium was irradiated at 100°C up to the fast neutron fluence of 1 × 1026 n/m2. Transmutation tritium and helium contents were 652 and 4400 appm, respectively. Post-irradiation studies of beryllium consist of optical and electron microscopy, density measurements before and after isochronal annealing at the temperature range of 300–1100°C and thermodesorption gas spectrometry. Investigation shows the following: (1) Slight swelling of beryllium after neutron irradiation. (2) Spatial non-uniformity in the distribution of the pores. (3) Complicated dependence of swelling on annealing temperature caused by formation of gas porosity. In the temperature range from 500 to 800°C, swelling of beryllium was probably caused by growth of bubbles because of tritium mobility. At the temperature above 900°C swelling of beryllium was probably caused by growth of bubbles because of helium mobility. (4) Full degassing of the irradiated beryllium took place below its melting temperature.

Investigation of the mechanism of vacancy loop nucleation in the depleted zones in α-iron

Abstract The method of molecular dynamics has been used for simulation of the mechanism of vacancy loop (VL) nucleation in the depleted zones (DZ) of the cascades in α-iron. It is shown that in α-iron either “friable” three-dimentional clusters or VL nuclei in the {110} plane can be nucleated depending on the irradiation conditions, the number of vacancies in the cluster being the critical parameter determining transition from one type to another. Comparison is made of the kinetics of depleted zones relaxation in α-iron and copper. Explanation of the reasons of low efficiency of VL nucleation in DZ of α-iron is given.

Computer simulation of vacancy loops and stacking faults in zirconium

Abstract Stacking fault (SF) in the basal and prism planes, formed in “collapsing” the vacancy platelets have been investigated by computer simulation using the long-range pair potential for the hcp zirconium. Small vacancy clusters (VC) of various configurations and large vacancy loops (VL) in the [1010] and [0001] planes were simulated. In comparison with the other types of the vacancy clusters platelets in [1010] and [0001] planes were found to be the most energy favorable configurations. The results of simulating large edge VL in the basal and prism planes were compared with the continuum theory. This permitted the dislocation core radii and the factors of the elastic energy of dislocations in the vacancy loops [0001] 1 2 [0001] and [1010] 1 2 [10 1 0] to be determined. Simulating the unfaulting process prism loops [1010] 1 3 [11 2 0] and basal loops (0001) 1 2 [0001] were obtained.

Simulation of defect cascade collapse in hcp zirconium

Abstract Using the molecular dynamics method, the mechanism of vacancy loop (VL) nucleation in depleted zones (DZ) of displacement cascades in hcp zirconium has been investigated in terms of the thermal spike model. It is shown that the type of nucleation VL (basal or prism) depends on the average volume concentration of vacancies C v ∗ in the depleted zone at the moment of its crystallization at the thermal stage of development of the displacement cascade. At the vacancy concentration C v ∗ = 10 at% in DZ of hep zirconium prism loops (1010) with Burgers vectors b = 1 2 〈10 1 0〉 and b = 1 3 〈11 2 0〉 predominantly nucleate. At C v ∗ = 15 at%, loops in the basal plane with b = 1 6 〈20 2 3〉 were observed to nucleate. Simulation of various conditions of depleted zone cooling has revealed that in the case of accelerated DZ cooling conditions from the high-temperature region, formation of three-dimensional precipitate of the metastable bcc phase is possible.

Investigation of samples taken from Kozloduy unit 2 reactor pressure vessel

Abstract Within the framework of the 6 month WANO program, small samples were cut from the inside surface of the Kozloduy NPP unit 2 reactor pressure vessel to assess the actual condition of the pressure vessel material before and after annealing. The actual values of the weld metal characteristics required for estimating radiation-limited lifetime—the ductile-to-brittle transition temperature (DBTT) in the initial state ( T ko ) and the phosphorus and copper contents which affect the radiation stability of steel—were not determined during manufacturing. The Kozloduy unit 2 pressure vessel had no surveillance program. Radiation stability was evaluated using dependencies based on analysis results for surveillance samples taken from other VVER-440 reactors. For this reason, the actual pressure vessel characteristics and their changes in the course of reactor operation, as well as comparison of experimental with calculated data were the principle objectives of the study. Instrumented impact tests were carried out on sub-size specimens of base and weld metal. Correlation dependencies were used with standard tests to determine DBTTs for the base and weld metal (in accordance with Russian standards): base metal before annealing 40 °C, after annealing 16 °C; weld metal before annealing 212 °C, after annealing 70 °C. The estimated value of T ko , for the initial, unirradiated weld metal, was 50 °C. The experimental results were compared with a prediction of the extent of radiation-induced embrittlement of Kozloduy unit 2 pressure vessel materials. It was confirmed that radiation-induced embrittlement of the base metal does not impose any limits on the radiation-limited lifetime of the pressure vessel. The predicted increase in the DBTT of the weld metal as a result of irradiation (about 165 °C) is practically equal to the experimental result (162 °C). However, the value of T f obtained from tests before annealing (212 °C) is about 40 °C higher that the estimated value, i.e. the calculation does not produce a conservative estimate. This was explained by a low estimate of T ko (10 °C), which had been calculated using data from chemical analysis of the weld metal, performed by the manufacturer. The investigations on the samples, however, yielded an estimated value of T ko = 50 °C. The effectiveness of annealing in restoring the mechanical properties of irradiated VVER-440 reactor pressure vessels was confirmed. Recovery annealing lowered the DBTT of the weld metal by 85% or more of its radiation-induced shift.

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.

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.

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.