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Estimation of thermal loads on the VVER vessel under conditions of inversion of the stratified molten pool in a severe accident
Loktionov V.D., Mukhtarov E.S.
Verification of the EUCLID/V2 Code Based on Experiments Involving Destruction of a Liquid Metal Cooled Reactor’s Core Components
Butov A.A., Zhdanov V.S., Klimonov I.A., Kudashov I.G., Kutlimetov A.E., Lobanov P.D., Mosunova N.A., Sorokin A.A., Strizhov V.F., Usov E.V., Chukhno V.I.
Topical Problems Concerned with the Thermophysical Characteristics of New-Generation Light Water Reactors: Comprehensive Study Results
Sorokin A.P., Kuzina Y.A., Trufanov A.A., Loshchinin V.M., Levchenko Y.D., Morozov A.V.
Numerical Investigation of a Class of Accidents in the Generation IV Brest Reactor Involving the Formation of a Solid Phase in the Lead Coolant
Chistov A.S., Savikhin O.G., Ovchinnikov V.F., Kiryushina E.V.
The EUCLID/V1 Integrated Code for Safety Assessment of Liquid Metal Cooled Fast Reactors. Part 2: Validation and Verification
Alipchenkov V.M., Boldyrev A.V., Veprev D.P., Zeigarnik Y.A., Kolobaeva P.V., Moiseenko E.V., Mosunova N.A., Seleznev E.F., Strizhov V.F., Usov E.V., Osipov S.L., Gorbunov V.S., Afremov D.A., Semchenkov A.A.
Experimental studies of heat exchange for sodium boiling in the fuel assembly model: Safety substantiation of a promising fast reactor
Khafizov R.R., Poplavskii V.M., Rachkov V.I., Sorokin A.P., Trufanov A.A., Ashurko Y.M., Volkov A.V., Ivanov E.F., Privezentsev V.V.
Modeling of corrosion product migration in the secondary circuit of nuclear power plants with WWER-1200
Kritskii V.G., Berezina I.G., Gavrilov A.V., Motkova E.A., Zelenina E.V., Prokhorov N.A., Gorbatenko S.P., Tsitser A.A.
Achieving More Efficient Removal of α-Emitting Radionuclides from the Primary Coolant in Propulsion Reactors
Orlov S.N., Zmitrodan A.A., Mysik S.G.
The Impact of Design Deviations on Dynamic Characteristics of a Flexible Multispan Rotor on Complete Electromagnetic Suspension
Ovchinnikov V.F., Nikolaev M.Y., Litvinov V.N., Kodochigov N.G., Drumov I.V.
Comparison of the technical and economic parameters of different variants of the nuclear fuel cycle reactors of the nuclear power plants
Tolstoukhov D.A., Panov S.A., Adamov E.O., Rachkov V.I.
Fundamentals, current state of the development of, and prospects for further improvement of the new-generation thermal-hydraulic computational HYDRA-IBRAE/LM code for simulation of fast reactor systems
Alipchenkov V.M., Anfimov A.M., Afremov D.A., Gorbunov V.S., Zeigarnik Y.A., Kudryavtsev A.V., Osipov S.L., Mosunova N.A., Strizhov V.F., Usov E.V.
Verification of the TIGRSP Computer Code as Applied to a 19-Rod Fuel Assembly Using CFD Computations
Stepanov O.E., Galkin I.Y., Melekh S.S., Kurnosov M.M., Pronin A.A.
The EUCLID/V1 Integrated Code for Safety Assessment of Liquid Metal Cooled Fast Reactors. Part 1: Basic Models
Mosunova N.A.
Thermodynamic Simulation of Equilibrium Composition of Reaction Products at Dehydration of a Technological Channel in a Uranium-Graphite Reactor
Pavliuk A.O., Zagumennov V.S., Kotlyarevskiy S.G., Bespala E.V.
Analysis and selection of high pressure heaters design for a new generation of NPP with BN-1200 reactor plant
Yurchenko A.Y., Sukhorukov Y.G., Trifonov N.N., Grigor’eva E.B., Esin S.B., Svyatkin F.A., Nikolaenkova E.K., Prikhod’ko P.Y., Nazarov V.V.
The EUCLID/V2 Code Physical Models for Calculating Fuel Rod and Core Failures in a Liquid Metal Cooled Reactor
Butov A.A., Zhdanov V.S., Klimonov I.A., Kudashov I.G., Kutlimetov A.E., Mosunova N.A., Strizhov V.F., Sorokin A.A., Frolov S.A., Usov E.V., Chukhno V.I.
New Technological Platform for the National Nuclear Energy Strategy Development
Adamov E.O., Rachkov V.I.
Procedure of calculation of the spatial distribution of temperatures and heat fluxes in the steam generator of a nuclear power installation with an RBEC fast-neutron reactor
Frolov A.A., Sedov A.A.
Cross-Verification of 1D and 3D Models for a VVER-1000 Reactor’s Pressure Chamber Simulated by the KORSAR/CFD Computation Code in the Modes with Asymmetric Loop Operation
Yudov Y.V., Petkevich I.G., Artemov V.G.
Experimental investigation of a new method for advanced fast reactor shutdown cooling
Pakholkov V.V., Kandaurov A.A., Potseluev A.I., Rogozhkin S.A., Sergeev D.A., Troitskaya Y.I., Shepelev S.F.
Experimental studies of local coolant hydrodynamics using a scaled model of cassette-type fuel assembly of a KLT-40S reactor
Dmitriev S.M., Barinov A.A., Varentsov A.V., Doronkov D.V., Solntsev D.N., Khrobostov A.E.
Three-Dimensional Simulation of a VVER-1000 Reactor’s Pressure Chamber in the Modes with Asymmetrical Loop Operation Using a KORSAR/CFD Computation Code
Yudov Y.V., Petkevich I.G., Artemov V.G., Kasterin D.S., Rumyantsev S.N.
On the Question of “Decisive Advantages” of Thermionic Conversion for Space Power Systems
Koroteev A.S.
Fast reactor: an experimental study of thermohydraulic processes in different operating regimes
Opanasenko A.N., Sorokin A.P., Zaryugin D.G., Trufanov A.A.
Modeling the transport of nitrogen in an NPP-2006 reactor circuit
Stepanov O.E., Galkin I.Y., Sledkov R.M., Melekh S.S., Strebnev N.A.
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