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Dive into the research topics where M.S. Veshchunov is active.

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Featured researches published by M.S. Veshchunov.


Journal of Nuclear Materials | 2000

On the theory of fission gas bubble evolution in irradiated UO2 fuel

M.S. Veshchunov

The standard approaches for modelling of the intra- and intergranular bubbles evolution in the UO2 fuel are critically analysed on the basis of available experimental data. It is demonstrated that the main disadvantage of the simplified treatment of the problem by the standard models can be associated with underestimation of the radiation effects at low temperatures (below 1500°C) and thermal effects at high temperatures (above 1500°C). The presented analysis allows a quantitative description of the bubble nucleation mechanism, adequate modelling of the bubble diffusion growth, and evaluation of the intragranular bubble number density and stable size attained under steady irradiation conditions.


Journal of Nuclear Materials | 1996

Critical evaluation of uranium oxide dissolution by molten Zircaloy in different crucible tests

M.S. Veshchunov; P. Hofmann; A.V. Berdyshev

Abstract A critical analysis is presented of the experimental data on UO 2 dissolution by molten Zircaloy at high temperatures (⩾ 2000°C), obtained in the different UO 2 crucible tests by various investigators. On this basis, improvement and further development of the theoretical model proposed in a previous paper of the authors are performed. Resolution of apparent disagreements amongst the results obtained in the different tests and explanation of all the available data in the framework of the newly developed model are proposed.


Journal of Nuclear Materials | 1998

Modelling of hydrogen absorption by zirconium alloys during high temperature oxidation in steam

M.S. Veshchunov; A.V. Berdyshev

Abstract A model is developed for description of hydrogen absorption by Zr alloys during high temperature oxidation in steam, based on the detailed experimental results of recent tests on the kinetics of hydrogen absorption by Zr–1Nb cladding during steam oxidation in the temperature range from 900 to 1200°C. The standard consideration of the steam oxidation process in the framework of the coupled anodic/cathodic reactions at the two oxide interfaces (gas/oxide and oxide/metal) is modified, taking into account that hydrogen may intrude into oxide in the form of positively charged protons. For quantitative description of hydrogen behaviour by this mechanism, mass transfer in the three layers, gas, oxide and metal, and at corresponding interfaces, gas/oxide and oxide/metal, is self-consistently considered. Numerical solution of the model generally confirms the main conclusions of simplified analytical treatment and furnishes a satisfactory fitting between measured kinetic curves and calculations.


Journal of Nuclear Materials | 1998

Modeling of chemical interactions of fuel rod materials at high temperatures I. Simultaneous dissolution of UO2 and ZrO2 by molten Zr in an oxidizing atmosphere

M.S. Veshchunov; A.V. Berdyshev

Abstract Investigations and modeling of high-temperature processes associated with the oxidation of UZrO molten mixtures under various conditions of severe accidents (intact heated fuel rods or relocating melt) are presented on the basis of a thorough analysis of metallographic post-test examination data obtained in the fuel bundle CORA experiments. In Part I a model of simultaneous dissolution of solid UO2 and ZrO2 phases by molten Zr in an oxidizing atmosphere (steam) is presented. This modeling is performed on the basis of the generalization of a previously developed model of UO2 dissolution by molten Zr taking into account the chemical dissolution by convectively stirring melt of a ZrO2 layer interacting with steam. The model describes the complicated kinetics of these interactions accompanied by precipitation of a ceramic (U, Zr)O2−x phase in the bulk of the liquid. This process finally leads to complete conversion of the melt into the (U, Zr)O2−x layer located between UO2 fuel and ZrO2 shell. Such a three-layer structure was regularly observed in the CORA post-test metallographic examinations of fuel rods heated above the melting point of Zr cladding (T ≥ 1950°C).


Journal of Nuclear Materials | 1992

On the kinetics of UO2 interaction with molten zircaloy at high temperatures

M.S. Veshchunov; A.M. Volchek

Abstract The theoretical study of kinetic processes of UO2/Zr reaction at high temperatures is presented. Possible reasons for the plane interphase boundaries destruction are suggested for two temperature ranges studied experimentally, 1200–1700°C (solid Zr) and 1900–2200°C (liquid Zr). On the basis of the proposed mechanisms of moving boundaries instability, two different approaches are presented to describe mass transfer through the two-phase region, that appears as a result of interphase boundary instability.


Journal of Nuclear Materials | 1998

Modeling of chemical interactions of fuel rod materials at high temperatures II. Investigation of downward relocation of molten materials

M.S. Veshchunov; A.V. Palagin

Abstract In Part II of the modeling of chemical interactions of fuel rod materials at high temperatures, qualitative results on the nature of Zr-rich melt oxidation and interactions with fuel rods allow further interpretation of the post-test examinations of structures (debris) formed in the CORA tests under more complicated conditions, namely during downward relocation of the melt. In this situation, the molten mass extensively oxidizes and simultaneously dissolves UO2 pellets and ZrO2 scales of the cladding. The analysis of these simultaneous physico-chemical processes on the basis of the kinetic oxidation/dissolution model developed in Part I of the paper, allows a new interpretation and explanation of the CORA tests results concerning relocation dynamics of the major part of the melt (slow relocation of melt in the form of massive slug rather than quick relocations of droplets and rivulets), formation of local blockages (debris) in the interrod space and accumulation of the melt in the core region in the form of molten pool.


Journal of Nuclear Materials | 1992

On the theory of high temperature transition in fluorite-type oxides

M.S. Veshchunov

From the general symmetry point of view the transition into superionic state in the crystals with fluorite structure is studied within the framework of phenomenological theory. The Landau Hamiltonian defined in the paper restricts significantly the options for microscopic models that describe this transition. The nature of the observed anomalies of the heat capacity and diffusion coefficient could be explained within the proposed phenomenological consideration.


Journal of Nuclear Materials | 2007

Mechanistic modelling of urania fuel evolution and fission product migration during irradiation and heating

M.S. Veshchunov; R. Dubourg; V.D. Ozrin; V.E. Shestak; V.I. Tarasov


Journal of Nuclear Materials | 2008

Modelling of grain face bubbles coalescence in irradiated UO2 fuel

M.S. Veshchunov


Journal of Nuclear Materials | 2008

An advanced model for intragranular bubble diffusivity in irradiated UO2 fuel

M.S. Veshchunov; V.E. Shestak

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V.E. Shestak

Russian Academy of Sciences

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V.I. Tarasov

Russian Academy of Sciences

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A.V. Berdyshev

Russian Academy of Sciences

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A.V. Boldyrev

Russian Academy of Sciences

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V.D. Ozrin

Russian Academy of Sciences

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A.M. Volchek

Russian Academy of Sciences

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A.V. Palagin

Russian Academy of Sciences

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R. Dubourg

Institut de radioprotection et de sûreté nucléaire

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