A. D. Efanov

发表

Hydraulic losses in the section of mutual influence of local resistances

Yu. D. Levchenko, V. S. Fedotovskii, A. D. Efanov, 1997 .

Development of a Containment Code KUPOL-M for a Substantiation of the VVER’s Accident Localization System Functions

A. D. Efanov, A. A. Loukianov, N. N. Shan'gin, 2002 .

The General Post-CHF Heat Transfer Prediction Method for Tubes on the Base of the Look-up Table

V. N. Vinogradov, V. V. Sergeev, P. L. Kirillov, 1998 .

A modified table for calculating critical heat fluxes in assemblies of triangularly packed fuel rods

V. P. Bobkov, A. D. Efanov, I. P. Smogalev, 2011 .

Calculated assessment of hydrogen release for a VVER-1000 reactor plant taking into account the reaction of concrete with spreading corium

A. A. Kazantsev, T. V. Popova, A. D. Efanov, 2014 .

Hydrodynamic problems of heavy liquid metal coolants technology in loop-type and mono-block-type reactor installations

P. N. Martynov, Yu. D. Levchenko, V. V. Ulyanov, 2007 .

Verifying the thermohydraulic codes using the standard problem of bottom reflooding

V. V. Sergeev, V. N. Vinogradov, A. D. Efanov, 2003 .

Use of liquid metals in nuclear and thermonuclear engineering, and in other innovative technologies

V. I. Rachkov, S. G. Kalyakin, A. D. Efanov, 2014 .

Investigation of Operation of VVER Steam Generator in Condensation Mode at the Large-Scale Test Rig

S. G. Kalyakin, O. V. Remizov, A. V. Morozov, 2008 .

Verification of the KUPOL-M code using the results of experiments

V. V. Bezlepkin, S. E. Semashko, A. D. Efanov, 2004 .

Analytical and experimental validation of the hydrogen mitigation system for the containment shell of a nuclear power station with a VVER-640 reactor

Yu. N. Shebeko, V. V. Bezlepkin, S. E. Semashko, 1995 .

Verifying the post-burnout heat-transfer model of the thermohydraulic code KORSAR

V. V. Sergeev, V. N. Vinogradov, A. D. Efanov, 2002 .

Substantiation and verification of the model for burnout in rod bundles of the KORSAR thermohydraulic code

V. P. Bobkov, V. V. Sergeev, V. N. Vinogradov, 2003 .

ICONE11-36407 THERMOHYDRAULIC RESEARCH FOR THE CORE OF THE BREST-OD-300 REACTOR

A. P. Sorokin, A. I. Filin, A. V. Zhukov, 2003 .

Heavy Liquid Metal Coolant – Lead–Bismuth and Lead – Technology

P. N. Martynov, A. E. Rusanov, A. V. Zrodnikov, 2004 .

Thermophysical studies for validating new-generation nuclear-reactor designs

S. G. Kalyakin, A. D. Efanov, A. Sorokin, 2012 .

Table for Critical Heat Fluxes in Square-Packed Fuel-Element Assemblies

A. D. Efanov, I. P. Smogalev, V. Bobkov, 2005 .

Investigation of thermodynamic properties of sodium-lead system

A. D. Efanov, N. I. Loginov, V. A. Morozov, 2008 .

Heat transfer during boiling of a liquid metal during emergency cooldown of a fast-neutron reactor

A. D. Efanov, A. Sorokin, G. Bogoslovskaya, 1999 .

Heat transfer under natural convection of liquid metal during its boiling in a system of channels

A. D. Efanov, V. Ivanov, A. Sorokin, 2007 .

An investigation of the heat transfer and stability of liquid-metal coolant boiling in a natural circulation circuit

A. D. Efanov, A. Sorokin, G. Bogoslovskaya, 2003 .