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O. O. Zabusov

Bio: O. O. Zabusov is an academic researcher from Kurchatov Institute. The author has contributed to research in topics: Embrittlement & Reactor pressure vessel. The author has an hindex of 8, co-authored 19 publications receiving 291 citations.

Papers
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TL;DR: In this paper, high number densities of 2-nm-diameter Ni-, Si- and Mn-enriched nanoclusters were found in the neutron irradiated base and weld metals.

104 citations

Journal ArticleDOI
TL;DR: In this article, it has been established that irradiation induces both intergranular as well as intragranular phosphorus segregation in Russian reactor pressure vessel steels and that brittle and ductile fracture surfaces of Charpy specimens appear as a result of inter-granular and intraggranular segregation, respectively.

46 citations

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TL;DR: In this paper, the influence of structural parameters on the tendency of steels to reversible temper embrittlement was studied for assessment of performance properties of reactor pressure vessel steels with extended service life.

35 citations

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TL;DR: In this paper, the fine structure and mechanical properties of VVER-1000 pressure vessel steels were studied using methods of transmission electron microscopy, fractographic analysis, and Auger electron spectroscopy.

34 citations

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TL;DR: In this paper, the results of VVER-440 steel Sv-10KhMFT and SV-1000 steel SV-10khGNMAA investigations by transmission electron microscopy, scanning electron microscopes, Auger-electron spectroscopy and mechanical tests are presented.

34 citations


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Journal ArticleDOI
TL;DR: In this article, the authors reviewed current phenomenological knowledge and understanding of mechanisms for radiation embrittlement of reactor pressure vessel low alloy steels and irradiation assisted stress corrosion cracking of core internals of stainless steels.
Abstract: Current phenomenological knowledge and understanding of mechanisms are reviewed for radiation embrittlement of reactor pressure vessel low alloy steels and irradiation assisted stress corrosion cracking of core internals of stainless steels. Accumulated test data of irradiated materials in light water reactors and microscopic analyses by using state-of-the-art techniques such as a three-dimensional atom probe and electron backscatter diffraction have significantly increased knowledge about microstructural features. Characteristics of solute clusters and deformation microstructures and their contributions to macroscopic material property changes have been clarified to a large extent, which provide keys to understand in the degradation mechanisms. However, there are still fundamental research issues that merit study for long-term operation of reactors that requires reliable quantitative prediction of radiation-induced degradation of component materials in low-dose rate high-dose conditions.

134 citations

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TL;DR: In this article, Atom probe tomography was used to study the formation of Mn-Ni-Si-dominated precipitates in irradiated Cu-free and Cu-bearing reactor pressure vessel steels.

114 citations

Journal ArticleDOI
09 Jul 2009-JOM
TL;DR: In this paper, the authors developed a model to predict the ductile-brittle transition temperature (ΔT) of reactor pressure vessel steels at high fluence conditions that are far outside the existing database.
Abstract: Nuclear plant life extension to 80 years will require accurate predictions of neutron irradiation-induced increases in the ductile-brittle transition temperature (ΔT) of reactor pressure vessel steels at high fluence conditions that are far outside the existing database. Remarkable progress in mechanistic understanding of irradiation embrittlement has led to physically motivated ΔT correlation models that provide excellent statistical fits to the existing surveillance database. However, an important challenge is developing advanced embrittlement models for low flux-high fluence conditions pertinent to extended life. These new models must also provide better treatment of key variables and variable combinations and account for possible delayed formation of “late blooming phases” in low copper steels. Other issues include uncertainties in the compositions of actual vessel steels, methods to predict ΔT attenuation away from the reactor core, verification of the master curve method to directly measure the fracture toughness with small specimens and predicting ΔT for vessel annealing remediation and re-irradiation cycles.

105 citations

Journal ArticleDOI
TL;DR: The Ringhals Units 3 and 4 PWRs in Sweden are pressurized water reactors (PWRs) designed and supplied by Westinghouse Electric Company, with commercial operation in 1981 and 1983, respectively as discussed by the authors.

94 citations

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TL;DR: In this article, nonlinear ultrasound was used to monitor radiation damage in two reactor pressure vessel (RPV) steels, and the results showed a clear increase in the measured acoustic nonlinearity from the unirradiated state to the medium dose, and then a decrease from medium dose to high dose.
Abstract: Nonlinear ultrasound was used to monitor radiation damage in two reactor pressure vessel (RPV) steels. The microstructural changes associated with radiation damage include changes in dislocation density and the formation of precipitates, and nonlinear ultrasonic waves are known to be sensitive to such changes. Six samples each of two different RPV steels were previously irradiated in the Rheinsberg power reactor to two fluence levels, up to 1020 n/cm2 (E > 1 MeV). Longitudinal waves were used to measure the acoustic nonlinearity in these samples, and the results show a clear increase in the measured acoustic nonlinearity from the unirradiated state to the medium dose, and then a decrease from medium dose to high dose.

90 citations