The Experts below are selected from a list of 258 Experts worldwide ranked by ideXlab platform
E. V. Usov - One of the best experts on this subject based on the ideXlab platform.
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The EUCLID/V2 Code Physical Models for Calculating Fuel Rod and Core Failures in a Liquid Metal Cooled Reactor
Thermal Engineering, 2019Co-Authors: A. A. Butov, N A Mosunova, V. S. Zhdanov, I. A. Klimonov, I. G. Kudashov, A. E. Kutlimetov, V. F. Strizhov, A. A. Sorokin, S. A. Frolov, E. V. UsovAbstract:— The article describes the basic models laid down in the second version of the EUCLID/V2 integrated code developed for carrying out end-to-end analysis of severe accidents in liquid metal cooled reactors. Brief information about the basic analogs of the code is given. Unlike the first version of the code, its second version includes additional tools for analyzing design-basis and beyond-design-basis accidents involving Fuel pin, Fuel assembly, and reactor core failures. To this end, the code is supplemented with additional modules using which it is possible to calculate Fuel Rod tightness failure as a consequence of its melting, escape of fission pRoducts into the coolant, their transport over the circuit, and release into the nuclear power plant rooms. The code also incorporates modules for calculating the core failure processes. Special attention is paid to the physical models for calculating the core materials' melting processes, motion of the pRoduced melt, its interaction with the coolant and with other materials, and propagation of fission materials. For calculating the core failure processes, a multicomponent 3D model has been implemented. The methods used for calculating heat transfer and friction between the components are based on well-proven analytical and empirical relations for determining the heat transfer and friction coefficients. The coefficients presented in the article also depend on the obtained multicomponent flow motion regime and the type of components (metal and ceramics). The algorithms governing joint operation of the thermomechanical, thermal-hydraulic, neutronics, and the Fuel Rod thermal failure module are described. Emphasis is placed on data exchange methods in the course of an accident in the reactor. The approaches used for calculating the transport of fission pRoducts in the coolant and in the NPP rooms are presented.
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a step in the verification of the hydra ibrae lm v1 thermohydraulic code for calculating sodium coolant flow in Fuel Rod assemblies
Atomic Energy, 2015Co-Authors: E. V. Usov, N A Mosunova, A. A. Butov, I. G. Kudashov, V. F. Strizhov, N. A. Pribaturin, G. A. Dugarov, E. N. IvanovAbstract:The basic relations used in the HYDRA-IBRAE/LM/V1 code for calculating the flow and heat-exchange of sodium coolant in Fuel-Rod assemblies in regimes with and without boiling and in the presence of a crisis of heat exchange are presented. The experimental data suitable for verifying the thermohydraulic code for calculating friction in one- and two-phase regimes as well as heat exchange in regimes with and without boiling are picked on the basis of a review of the literature. The computational error of different parameters is obtained on the basis of the verification results. It is shown that the thermohydraulic code HYDRA-IBRAE/LV/V1 can be used to calculate correctly the primary processes in design basis and beyond-design basis accidents in sodium-cooled reactor facilities.
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A Step in the Verification of the Hydra-Ibrae/LM/V1 Thermohydraulic Code for Calculating Sodium Coolant Flow in Fuel-Rod Assemblies
Atomic Energy, 2015Co-Authors: E. V. Usov, N A Mosunova, A. A. Butov, I. G. Kudashov, V. F. Strizhov, N. A. Pribaturin, G. A. Dugarov, E. N. IvanovAbstract:The basic relations used in the HYDRA-IBRAE/LM/V1 code for calculating the flow and heat-exchange of sodium coolant in Fuel-Rod assemblies in regimes with and without boiling and in the presence of a crisis of heat exchange are presented. The experimental data suitable for verifying the thermohydraulic code for calculating friction in one- and two-phase regimes as well as heat exchange in regimes with and without boiling are picked on the basis of a review of the literature. The computational error of different parameters is obtained on the basis of the verification results. It is shown that the thermohydraulic code HYDRA-IBRAE/LV/V1 can be used to calculate correctly the primary processes in design basis and beyond-design basis accidents in sodium-cooled reactor facilities.
A. A. Butov - One of the best experts on this subject based on the ideXlab platform.
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The EUCLID/V2 Code Physical Models for Calculating Fuel Rod and Core Failures in a Liquid Metal Cooled Reactor
Thermal Engineering, 2019Co-Authors: A. A. Butov, N A Mosunova, V. S. Zhdanov, I. A. Klimonov, I. G. Kudashov, A. E. Kutlimetov, V. F. Strizhov, A. A. Sorokin, S. A. Frolov, E. V. UsovAbstract:— The article describes the basic models laid down in the second version of the EUCLID/V2 integrated code developed for carrying out end-to-end analysis of severe accidents in liquid metal cooled reactors. Brief information about the basic analogs of the code is given. Unlike the first version of the code, its second version includes additional tools for analyzing design-basis and beyond-design-basis accidents involving Fuel pin, Fuel assembly, and reactor core failures. To this end, the code is supplemented with additional modules using which it is possible to calculate Fuel Rod tightness failure as a consequence of its melting, escape of fission pRoducts into the coolant, their transport over the circuit, and release into the nuclear power plant rooms. The code also incorporates modules for calculating the core failure processes. Special attention is paid to the physical models for calculating the core materials' melting processes, motion of the pRoduced melt, its interaction with the coolant and with other materials, and propagation of fission materials. For calculating the core failure processes, a multicomponent 3D model has been implemented. The methods used for calculating heat transfer and friction between the components are based on well-proven analytical and empirical relations for determining the heat transfer and friction coefficients. The coefficients presented in the article also depend on the obtained multicomponent flow motion regime and the type of components (metal and ceramics). The algorithms governing joint operation of the thermomechanical, thermal-hydraulic, neutronics, and the Fuel Rod thermal failure module are described. Emphasis is placed on data exchange methods in the course of an accident in the reactor. The approaches used for calculating the transport of fission pRoducts in the coolant and in the NPP rooms are presented.
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a step in the verification of the hydra ibrae lm v1 thermohydraulic code for calculating sodium coolant flow in Fuel Rod assemblies
Atomic Energy, 2015Co-Authors: E. V. Usov, N A Mosunova, A. A. Butov, I. G. Kudashov, V. F. Strizhov, N. A. Pribaturin, G. A. Dugarov, E. N. IvanovAbstract:The basic relations used in the HYDRA-IBRAE/LM/V1 code for calculating the flow and heat-exchange of sodium coolant in Fuel-Rod assemblies in regimes with and without boiling and in the presence of a crisis of heat exchange are presented. The experimental data suitable for verifying the thermohydraulic code for calculating friction in one- and two-phase regimes as well as heat exchange in regimes with and without boiling are picked on the basis of a review of the literature. The computational error of different parameters is obtained on the basis of the verification results. It is shown that the thermohydraulic code HYDRA-IBRAE/LV/V1 can be used to calculate correctly the primary processes in design basis and beyond-design basis accidents in sodium-cooled reactor facilities.
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A Step in the Verification of the Hydra-Ibrae/LM/V1 Thermohydraulic Code for Calculating Sodium Coolant Flow in Fuel-Rod Assemblies
Atomic Energy, 2015Co-Authors: E. V. Usov, N A Mosunova, A. A. Butov, I. G. Kudashov, V. F. Strizhov, N. A. Pribaturin, G. A. Dugarov, E. N. IvanovAbstract:The basic relations used in the HYDRA-IBRAE/LM/V1 code for calculating the flow and heat-exchange of sodium coolant in Fuel-Rod assemblies in regimes with and without boiling and in the presence of a crisis of heat exchange are presented. The experimental data suitable for verifying the thermohydraulic code for calculating friction in one- and two-phase regimes as well as heat exchange in regimes with and without boiling are picked on the basis of a review of the literature. The computational error of different parameters is obtained on the basis of the verification results. It is shown that the thermohydraulic code HYDRA-IBRAE/LV/V1 can be used to calculate correctly the primary processes in design basis and beyond-design basis accidents in sodium-cooled reactor facilities.
N A Mosunova - One of the best experts on this subject based on the ideXlab platform.
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The EUCLID/V2 Code Physical Models for Calculating Fuel Rod and Core Failures in a Liquid Metal Cooled Reactor
Thermal Engineering, 2019Co-Authors: A. A. Butov, N A Mosunova, V. S. Zhdanov, I. A. Klimonov, I. G. Kudashov, A. E. Kutlimetov, V. F. Strizhov, A. A. Sorokin, S. A. Frolov, E. V. UsovAbstract:— The article describes the basic models laid down in the second version of the EUCLID/V2 integrated code developed for carrying out end-to-end analysis of severe accidents in liquid metal cooled reactors. Brief information about the basic analogs of the code is given. Unlike the first version of the code, its second version includes additional tools for analyzing design-basis and beyond-design-basis accidents involving Fuel pin, Fuel assembly, and reactor core failures. To this end, the code is supplemented with additional modules using which it is possible to calculate Fuel Rod tightness failure as a consequence of its melting, escape of fission pRoducts into the coolant, their transport over the circuit, and release into the nuclear power plant rooms. The code also incorporates modules for calculating the core failure processes. Special attention is paid to the physical models for calculating the core materials' melting processes, motion of the pRoduced melt, its interaction with the coolant and with other materials, and propagation of fission materials. For calculating the core failure processes, a multicomponent 3D model has been implemented. The methods used for calculating heat transfer and friction between the components are based on well-proven analytical and empirical relations for determining the heat transfer and friction coefficients. The coefficients presented in the article also depend on the obtained multicomponent flow motion regime and the type of components (metal and ceramics). The algorithms governing joint operation of the thermomechanical, thermal-hydraulic, neutronics, and the Fuel Rod thermal failure module are described. Emphasis is placed on data exchange methods in the course of an accident in the reactor. The approaches used for calculating the transport of fission pRoducts in the coolant and in the NPP rooms are presented.
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development and validation of the berkut Fuel Rod module of the euclid v1 integrated computer code
Annals of Nuclear Energy, 2018Co-Authors: D P Veprev, A V Boldyrev, Yu S Chernov, N A MosunovaAbstract:Abstract The EUCLID/V1 integrated computer code has been developed for the safety analysis and justification of reactor facilities with fast reactors under normal operating conditions and design basis accidents. The EUCLID/V1 code uses the BERKUT module to describe physical processes occurring in Fuel Rods with oxide or nitride Fuel in fast reactors. The quasi two dimensional approach is employed to simulate the thermomechanical behavior of Fuel Rods. Verification and validation of the BERKUT module have been carried out using analytical and numerical tests and the experimental data observed in the in-pile Fuel performance studies. The results of validation and verification obtained taking into account the uncertainties of the Fuel Rod fabrication data, material properties and irradiation conditions indicate that the BERKUT module describes adequately the Fuel Rod behavior in fast reactors.
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a step in the verification of the hydra ibrae lm v1 thermohydraulic code for calculating sodium coolant flow in Fuel Rod assemblies
Atomic Energy, 2015Co-Authors: E. V. Usov, N A Mosunova, A. A. Butov, I. G. Kudashov, V. F. Strizhov, N. A. Pribaturin, G. A. Dugarov, E. N. IvanovAbstract:The basic relations used in the HYDRA-IBRAE/LM/V1 code for calculating the flow and heat-exchange of sodium coolant in Fuel-Rod assemblies in regimes with and without boiling and in the presence of a crisis of heat exchange are presented. The experimental data suitable for verifying the thermohydraulic code for calculating friction in one- and two-phase regimes as well as heat exchange in regimes with and without boiling are picked on the basis of a review of the literature. The computational error of different parameters is obtained on the basis of the verification results. It is shown that the thermohydraulic code HYDRA-IBRAE/LV/V1 can be used to calculate correctly the primary processes in design basis and beyond-design basis accidents in sodium-cooled reactor facilities.
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A Step in the Verification of the Hydra-Ibrae/LM/V1 Thermohydraulic Code for Calculating Sodium Coolant Flow in Fuel-Rod Assemblies
Atomic Energy, 2015Co-Authors: E. V. Usov, N A Mosunova, A. A. Butov, I. G. Kudashov, V. F. Strizhov, N. A. Pribaturin, G. A. Dugarov, E. N. IvanovAbstract:The basic relations used in the HYDRA-IBRAE/LM/V1 code for calculating the flow and heat-exchange of sodium coolant in Fuel-Rod assemblies in regimes with and without boiling and in the presence of a crisis of heat exchange are presented. The experimental data suitable for verifying the thermohydraulic code for calculating friction in one- and two-phase regimes as well as heat exchange in regimes with and without boiling are picked on the basis of a review of the literature. The computational error of different parameters is obtained on the basis of the verification results. It is shown that the thermohydraulic code HYDRA-IBRAE/LV/V1 can be used to calculate correctly the primary processes in design basis and beyond-design basis accidents in sodium-cooled reactor facilities.
V. F. Strizhov - One of the best experts on this subject based on the ideXlab platform.
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The EUCLID/V2 Code Physical Models for Calculating Fuel Rod and Core Failures in a Liquid Metal Cooled Reactor
Thermal Engineering, 2019Co-Authors: A. A. Butov, N A Mosunova, V. S. Zhdanov, I. A. Klimonov, I. G. Kudashov, A. E. Kutlimetov, V. F. Strizhov, A. A. Sorokin, S. A. Frolov, E. V. UsovAbstract:— The article describes the basic models laid down in the second version of the EUCLID/V2 integrated code developed for carrying out end-to-end analysis of severe accidents in liquid metal cooled reactors. Brief information about the basic analogs of the code is given. Unlike the first version of the code, its second version includes additional tools for analyzing design-basis and beyond-design-basis accidents involving Fuel pin, Fuel assembly, and reactor core failures. To this end, the code is supplemented with additional modules using which it is possible to calculate Fuel Rod tightness failure as a consequence of its melting, escape of fission pRoducts into the coolant, their transport over the circuit, and release into the nuclear power plant rooms. The code also incorporates modules for calculating the core failure processes. Special attention is paid to the physical models for calculating the core materials' melting processes, motion of the pRoduced melt, its interaction with the coolant and with other materials, and propagation of fission materials. For calculating the core failure processes, a multicomponent 3D model has been implemented. The methods used for calculating heat transfer and friction between the components are based on well-proven analytical and empirical relations for determining the heat transfer and friction coefficients. The coefficients presented in the article also depend on the obtained multicomponent flow motion regime and the type of components (metal and ceramics). The algorithms governing joint operation of the thermomechanical, thermal-hydraulic, neutronics, and the Fuel Rod thermal failure module are described. Emphasis is placed on data exchange methods in the course of an accident in the reactor. The approaches used for calculating the transport of fission pRoducts in the coolant and in the NPP rooms are presented.
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a step in the verification of the hydra ibrae lm v1 thermohydraulic code for calculating sodium coolant flow in Fuel Rod assemblies
Atomic Energy, 2015Co-Authors: E. V. Usov, N A Mosunova, A. A. Butov, I. G. Kudashov, V. F. Strizhov, N. A. Pribaturin, G. A. Dugarov, E. N. IvanovAbstract:The basic relations used in the HYDRA-IBRAE/LM/V1 code for calculating the flow and heat-exchange of sodium coolant in Fuel-Rod assemblies in regimes with and without boiling and in the presence of a crisis of heat exchange are presented. The experimental data suitable for verifying the thermohydraulic code for calculating friction in one- and two-phase regimes as well as heat exchange in regimes with and without boiling are picked on the basis of a review of the literature. The computational error of different parameters is obtained on the basis of the verification results. It is shown that the thermohydraulic code HYDRA-IBRAE/LV/V1 can be used to calculate correctly the primary processes in design basis and beyond-design basis accidents in sodium-cooled reactor facilities.
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A Step in the Verification of the Hydra-Ibrae/LM/V1 Thermohydraulic Code for Calculating Sodium Coolant Flow in Fuel-Rod Assemblies
Atomic Energy, 2015Co-Authors: E. V. Usov, N A Mosunova, A. A. Butov, I. G. Kudashov, V. F. Strizhov, N. A. Pribaturin, G. A. Dugarov, E. N. IvanovAbstract:The basic relations used in the HYDRA-IBRAE/LM/V1 code for calculating the flow and heat-exchange of sodium coolant in Fuel-Rod assemblies in regimes with and without boiling and in the presence of a crisis of heat exchange are presented. The experimental data suitable for verifying the thermohydraulic code for calculating friction in one- and two-phase regimes as well as heat exchange in regimes with and without boiling are picked on the basis of a review of the literature. The computational error of different parameters is obtained on the basis of the verification results. It is shown that the thermohydraulic code HYDRA-IBRAE/LV/V1 can be used to calculate correctly the primary processes in design basis and beyond-design basis accidents in sodium-cooled reactor facilities.
I. G. Kudashov - One of the best experts on this subject based on the ideXlab platform.
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The EUCLID/V2 Code Physical Models for Calculating Fuel Rod and Core Failures in a Liquid Metal Cooled Reactor
Thermal Engineering, 2019Co-Authors: A. A. Butov, N A Mosunova, V. S. Zhdanov, I. A. Klimonov, I. G. Kudashov, A. E. Kutlimetov, V. F. Strizhov, A. A. Sorokin, S. A. Frolov, E. V. UsovAbstract:— The article describes the basic models laid down in the second version of the EUCLID/V2 integrated code developed for carrying out end-to-end analysis of severe accidents in liquid metal cooled reactors. Brief information about the basic analogs of the code is given. Unlike the first version of the code, its second version includes additional tools for analyzing design-basis and beyond-design-basis accidents involving Fuel pin, Fuel assembly, and reactor core failures. To this end, the code is supplemented with additional modules using which it is possible to calculate Fuel Rod tightness failure as a consequence of its melting, escape of fission pRoducts into the coolant, their transport over the circuit, and release into the nuclear power plant rooms. The code also incorporates modules for calculating the core failure processes. Special attention is paid to the physical models for calculating the core materials' melting processes, motion of the pRoduced melt, its interaction with the coolant and with other materials, and propagation of fission materials. For calculating the core failure processes, a multicomponent 3D model has been implemented. The methods used for calculating heat transfer and friction between the components are based on well-proven analytical and empirical relations for determining the heat transfer and friction coefficients. The coefficients presented in the article also depend on the obtained multicomponent flow motion regime and the type of components (metal and ceramics). The algorithms governing joint operation of the thermomechanical, thermal-hydraulic, neutronics, and the Fuel Rod thermal failure module are described. Emphasis is placed on data exchange methods in the course of an accident in the reactor. The approaches used for calculating the transport of fission pRoducts in the coolant and in the NPP rooms are presented.
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a step in the verification of the hydra ibrae lm v1 thermohydraulic code for calculating sodium coolant flow in Fuel Rod assemblies
Atomic Energy, 2015Co-Authors: E. V. Usov, N A Mosunova, A. A. Butov, I. G. Kudashov, V. F. Strizhov, N. A. Pribaturin, G. A. Dugarov, E. N. IvanovAbstract:The basic relations used in the HYDRA-IBRAE/LM/V1 code for calculating the flow and heat-exchange of sodium coolant in Fuel-Rod assemblies in regimes with and without boiling and in the presence of a crisis of heat exchange are presented. The experimental data suitable for verifying the thermohydraulic code for calculating friction in one- and two-phase regimes as well as heat exchange in regimes with and without boiling are picked on the basis of a review of the literature. The computational error of different parameters is obtained on the basis of the verification results. It is shown that the thermohydraulic code HYDRA-IBRAE/LV/V1 can be used to calculate correctly the primary processes in design basis and beyond-design basis accidents in sodium-cooled reactor facilities.
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A Step in the Verification of the Hydra-Ibrae/LM/V1 Thermohydraulic Code for Calculating Sodium Coolant Flow in Fuel-Rod Assemblies
Atomic Energy, 2015Co-Authors: E. V. Usov, N A Mosunova, A. A. Butov, I. G. Kudashov, V. F. Strizhov, N. A. Pribaturin, G. A. Dugarov, E. N. IvanovAbstract:The basic relations used in the HYDRA-IBRAE/LM/V1 code for calculating the flow and heat-exchange of sodium coolant in Fuel-Rod assemblies in regimes with and without boiling and in the presence of a crisis of heat exchange are presented. The experimental data suitable for verifying the thermohydraulic code for calculating friction in one- and two-phase regimes as well as heat exchange in regimes with and without boiling are picked on the basis of a review of the literature. The computational error of different parameters is obtained on the basis of the verification results. It is shown that the thermohydraulic code HYDRA-IBRAE/LV/V1 can be used to calculate correctly the primary processes in design basis and beyond-design basis accidents in sodium-cooled reactor facilities.
