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Oleg I. Melikhov
Observed names:
O I Melikhov,
O. I. Melikhov,
O.I. Melikhov,
Oleg I. Melikhov,
Oleg Melikhov
Institution:
Moscow Power Engineering Institute
Observed institutions:
Russian Academy of Sciences,
Moscow Power Engineering Institute,
Institute for Problems in Mechanics,
All-Russian Scientific Research Institute for Operation of Nuclear Power Plants
ORCID:
0000-0002-1379-5995
Observed ORCID:
0000-0002-1379-5995
OpenAlex ID:
A5102916825
Steam generator tube rupture in lead-cooled fast reactors: Estimation of impact on neighboring tubes
2018
·
A.S. Iskhakov
, V.I. Melikhov
, et al.
·
Nuclear Engineering and Design
·
35
Three-phase VOF modeling of water jet – molten metal interaction
2024
·
S.E. Yakush
, N.S. Sivakov
, et al.
·
Nuclear Engineering and Design
·
23
Recommended set of interfacial drag correlations for the two-phase flow under thermal–hydraulic conditions of a horizontal steam generator
2021
·
T.T. Le
, V.I. Melikhov
, et al.
·
Nuclear Engineering and Design
·
20
Validation of the STEG code using PGV experiments on hydrodynamics of horizontal steam generator
2019
·
Tung Thanh Le
, V.I. Melikhov
, et al.
·
Nuclear Engineering and Design
·
18
Hugoniot analysis of energetic molten lead-water interaction
2019
·
A.S. Iskhakov
, V.I. Melikhov
, et al.
·
Annals of Nuclear Energy
·
18
Simulation of the Thermal Hydraulic Processes in the Horizontal Steam Generator with the Use of the Different Interfacial Friction Correlations
2011
·
Vladimir Melikhov
, Oleg Melikhov
, et al.
·
Science and Technology of Nuclear Installations
·
16
PDF
Investigation of the equalization capability of submerged perforated sheets under thermal–hydraulic conditions of a horizontal steam generator
2020
·
T.T. Le
, V.I. Melikhov
, et al.
·
Annals of Nuclear Energy
·
15
VAPEX code-aided analysis of large-scale experiments in corium/water interaction
2007
·
V. I. Melikhov
, O. I. Melikhov
, et al.
·
High Temperature
·
15
Numerical analysis of the equalization capability of submerged perforated sheets for wwer-1500 horizontal steam generator
2021
·
T.T. Le
, V.I. Melikhov
, et al.
·
Annals of Nuclear Energy
·
13
The influence of void fraction on the submerged perforated sheet hydraulic friction factor
2015
·
V. N. Blinkov
, I. V. Elkin
, et al.
·
Thermal Engineering
·
13
Numerical Modeling of Water Jet Plunging in Molten Heavy Metal Pool
2023
·
Sergey E. Yakush
, Nikita S. Sivakov
, et al.
·
Mathematics
·
12
PDF
Evaluation of energy and impulse generated by superheated steam bubble collapse in subcooled water
2020
·
V.I. Melikhov
, O.I. Melikhov
, et al.
·
Nuclear Engineering and Design
·
12
Pressure Waves due to Rapid Evaporation of Water Droplet in Liquid Lead Coolant
2018
·
S. E. Yakush
, A. S. Iskhakov
, et al.
·
Science and Technology of Nuclear Installations
·
12
PDF
Influence of nonuniformity of the submerged perforated sheet on steam demand leveling on the evaporation surface of a VVER steam generator
2015
·
V. N. Blinkov
, I. V. Elkin
, et al.
·
Thermal Engineering
·
10
Validation of Fuel‐Coolant Interaction Model for Severe Accident Simulations
2011
·
Vladimir Melikhov
, Oleg Melikhov
, et al.
·
Science and Technology of Nuclear Installations
·
10
PDF
Experimental Studies for the VVER-440/213 Bubble Condenser System for Kola NPP at the Integral Test Facility BC V-213
2011
·
Vladimir N. Blinkov
, Oleg I. Melikhov
, et al.
·
Science and Technology of Nuclear Installations
·
9
PDF
Investigation into Behavior of a Steam-Water Mixture Flow Through Holes in a Submerged Perforated Sheet at High Void Fractions
2018
·
V. I. Melikhov
, O. I. Melikhov
, et al.
·
Thermal Engineering
·
8
Current state of system thermohydraulic codes and trends in their development abroad
2014
·
V. G. Asmolov
, V. N. Blinkov
, et al.
·
High Temperature
·
8
Validation of the STEG code using experiments on Two-Phase flow across horizontal tube bundles
2022
·
Shixian Liu
, Fei Yin
, et al.
·
Nuclear Engineering and Design
·
7
Hugoniot analysis of experimental data on steam explosion in stratified melt-coolant configuration
2019
·
A.S. Iskhakov
, V.I. Melikhov
, et al.
·
Nuclear Engineering and Design
·
7
Analysis of uncertainties for results of experiment with boiling-off the coolant simulation using SOCRAT-B1/B2-code for spent fuel pool conditions
2023
·
M.M. Bedretdinov
, S.I. Pantyushin
, et al.
·
Nuclear Engineering and Design
·
6
Melt/Water Explosive Interaction: VAPEX-D Code Simulation
2002
·
V. I. Melikhov
, O. I. Melikhov
, et al.
·
High Temperature
·
6
Experimental and Calculated Studies of Condensation-Induced Water Hammer
2021
·
G. J. Volkov
, I. V. Elkin
, et al.
·
Thermal Engineering
·
5
Investigation on the Interphase Drag and Wall Friction in Vertically Oriented Upward and Downward Two-Phase Flows Under Accident Conditions in Light Water Reactors
2021
·
Vladimir N. Blinkov
, Oleg I. Melikhov
, et al.
·
Nuclear Engineering and Design
·
5
Hot melt splash upon impingement of a water jet
2020
·
S E Yakush
, N S Sivakov
, et al.
·
Journal of Physics Conference Series
·
5
PDF
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Works count:
79
Citation count:
347
H-index:
12
I10-index:
15
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