Get started
Author

Vladimir V. Novikov

Observed names: V V Novikov, V. Novikov, V. V. Novikov, V.V Novikov, V.V. Novikov, Vladimir Novikov, Vladimir V. NOVIKOV, В. В. Новиков
ORCID: 0000-0001-7897-5957
Observed ORCID: 0000-0001-7897-5957
OpenAlex ID: A5102716180
Structure-Phase State, Corrosion and Irradiation Properties of Zr-Nb-Fe-Sn System Alloys
2008 · V. N. Shishov, M. M. Peregud, et al. · Journal of ASTM International · 56
Preliminary Results of the IAEA Review on Fuel Failures in Water Cooled Reactors
2010 · D. Dangouleme, V. Inozemtsev, et al. · Revue Générale Nucléaire · 23
Structure-Phase State, Corrosion and Irradiation Properties of Zr-Nb-Fe-Sn System Alloys
2009 · V. N. Shishov, M. M. Peregud, et al. · 17
Influence of irradiation on KISCC of Zr–1%Nb claddings
2000 · Yu.K Bibilashvili, A.V Medvedev, et al. · Journal of Nuclear Materials · 15
Structure-Phase Changes and Corrosion Behavior of E110 and E635 Claddings of Fuels in Water Cooled Reactors
2006 · Vladimir V. NOVIKOV, Vladimir A. MARKELOV, et al. · Journal of Nuclear Science and Technology · 14
PDF
Propagation of stress corrosion cracks in Zr-1% Nb claddings
1995 · Yu.K. Bibilashvily, Yu.N. Dolgov, et al. · Journal of Nuclear Materials · 13
Iodine induced SCC of Zr alloys at constant strain rate
1992 · S.B. Goryachev, A.R. Gritsuk, et al. · Journal of Nuclear Materials · 12
A Review of Hydride Reorientation in Zirconium Alloys for Water-Cooled Reactors
2020 · A. A. Plyasov, V. V. Novikov, et al. · Physics of Atomic Nuclei · 10
Preliminary Irradiation Effect on Corrosion Resistance of Zirconium Alloys
2018 · Vladimir Markelov, Vladimir Novikov, et al. · 10
Accident Tolerant Fuel with Chromium-Coated Fuel-Rod Cladding
2021 · L. A. Karpyuk, V. I. Kuznetsov, et al. · Atomic Energy · 7
Investigation of the Microstructure of Zirconium Alloys Irradiated by Zirconium Ions in an Accelerator
2014 · V. V. Novikov, V. N. Shishov, et al. · Atomic Energy · 7
Corrosion, Dimensional Stability and Microstructure of VVER-1000 E635 Alloy FA Components at Burnups up to 72 MWday/kgU
2014 · V. N. Shishov, V. A. Markelov, et al. · 6
Corrosion Model for Zirconium-Niobium Alloys in Pressurized Water Reactors
2014 · V. V. Likhanskii, I. A. Evdokimov, et al. · Atomic Energy · 6
Study of the radiation-induced thermal decomposition of cesium iodide and the corrosion cracking of water-moderated-water-cooled power-reactor fuel-element cans under gamma radiation
1986 · Yu. K. Bibilashvili, M. V. Vladimirova, et al. · Atomic Energy · 5
42KhNM Alloy and Silicon Carbide as Material for Accident-Tolerant Fuel-Rod Cladding
2021 · L. A. Karpyuk, V. V. Novikov, et al. · Atomic Energy · 4
PDF
Hydrogen Solubility in Zirconium Alloys
2020 · A. A. Plyasov, V. V. Novikov, et al. · Physics of Atomic Nuclei · 4
Alloying Effect of Niobium and Tin on the Zirconium Alloy Fuel Cladding Behavior at High Temperature Oxidation in Steam
2018 · Andrey Malgin, Vladimir Markelov, et al. · 4
Fracture Toughness of VVER and PWR Uranium-Dioxide Fuel Pellets with Different Grain Size
2015 · V. V. Novikov, R. B. Sivov, et al. · Atomic Energy · 4
Mechanism of iodine-induced cracking of zircalloy jackets
1991 · V. V. Novikov · Atomic Energy · 4
Fabrication of E110 Alloy Fuel Rod Claddings from Electrolytic Zirconium Base with Removing Fluorine Impurity for Providing Resistance to Breakaway Oxidation in High-Temperature Steam
2021 · Vladimir A. Markelov, Andrey G. Malgin, et al. · 3
Thermal expansion of cladding tubes and rods made of Zr-1% Nb alloy in the temperature range 293-873 K (20-600°C)based on X-ray and dilatometric measurements
2020 · M.G. Isaenkova, A.V. Tenishev, et al. · Physics and Chemistry of Materials Treatment · 3
Study of VVER fuel with dopant additives
2008 · V. V. Novikov, Yu. K. Bibilashvili, et al. · Atomic Energy · 3
Simulation of the Hydriding Failure of Fuel Elements in Water-Cooled Reactors
2003 · E. Yu. Afanas'eva, I. A. Evdokimov, et al. · Atomic Energy · 3
Delayed Hydride Cracking of Fuel Rod Claddings Made of Three Zirconium Alloys
2023 · Nikolay S. Saburov, Vladimir A. Markelov, et al. · 2
Identification of Relaxation Maxima of Internal Friction in Zr–Nb Alloy after Nitriding, Oxidation in Air, and In Situ Carburization
2022 · V. M. Anan’in, B. A. Kalin, et al. · Inorganic Materials Applied Research · 2
Works count: 58
Citation count: 259
H-index: 9
I10-index: 9
Year
Topic
Type