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ISSN 1063-4258 (Print) ISSN 1573-8205 (Online)
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关键字 Benchmark Experiment Beryllium Emergency Protection Fuel Assembly Fuel Element Heat Transfer Heavy Water Irradiation Dose Rate Moscow Engineer Physics Institute National Research Center Kurchatov Institute Neutron Spectrum Normal Operating Condition Plutonium Radionuclide Radon Spend Fuel Spend Nuclear Fuel Steam Generator Tritium Uranium Uranium Dioxide
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关键字 Benchmark Experiment Beryllium Emergency Protection Fuel Assembly Fuel Element Heat Transfer Heavy Water Irradiation Dose Rate Moscow Engineer Physics Institute National Research Center Kurchatov Institute Neutron Spectrum Normal Operating Condition Plutonium Radionuclide Radon Spend Fuel Spend Nuclear Fuel Steam Generator Tritium Uranium Uranium Dioxide
首页 > 检索 > 作者的详细信息

作者的详细信息

Usov, E. V.

期 栏目 标题 文件
卷 120, 编号 2 (2016) Article Two-Dimensional Thermohydraulic Module of the Integrated Code Sokrat-BN: Mathematical Model and Computational Results
卷 122, 编号 3 (2017) Article Modeling of Oxide Layer Formation and Corrosion Products Coagulation and Transport in Lead Coolant Using the OXID Module of the HYDRA-IBRAE/LM Code
卷 122, 编号 5 (2017) Article HYDRA-IBRAE/LM/V1 Thermohydraulic Code Verification Based on BN-600 Experiments
卷 124, 编号 3 (2018) Article Fuel Pin Melting in a Fast Reactor and Melt Solidification: Simulation Using the SAFR/V1 Module of the EVKLID/V2 Integral Code
卷 124, 编号 4 (2018) Article SAFR/V1 (EVKLID/V2 Integral Code Module) Aided Simulation of Melt Movement Along the Surface of a Fuel Element in a Fast Reactor During a Serious Accident
卷 124, 编号 5 (2018) Article Experiment-Based Verification of the SAFR/V1 Module of the EVKLID/V2 Integral Code for Thermal Breakdown of Fuel Pins in a Fast Reactor
卷 124, 编号 6 (2018) Article Verification of Analytical Test Based Thermohydraulic Systems Codes for One- and Two-Phase Liquid-Metal Flows
卷 127, 编号 1 (2019) Article 3D EVKLID/V2 Code Aided Simulation of Severe Accidents
 

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