Development of the Long Term prediction Methods for Stress Rupture Strength (SRS) of Chloronium-nickel Austenitic steel on the basis of Shortcut Tests.
Methodical Development for WWER-1000 Reactor Pressure Vessel Safe Operation Lifetime Evaluation Allowing for Anticorrosive Cladding
Development of Technology for Immobilization of Irradiated Fuel not Subject to Reprocessing.
Improvement of Corrosion Resistance of Constructional Steels in Liquid Pb and Pb-Bi Alloys by Means of Their Surface Modification with the Help of Pulsed Electron Beams and Protective Coatings
Modelling of Brittle and Ductile Fracture and Prediction of Irradiation Damage Effect on Fracture Toughness Properties of Steels for Reactor Pressure Vessels on the Basis of Local Approach
Prediction of Fracture Properties of Irradiated Austenitic and Ferritic Materials for Reactor Components of Nuclear Power Plants on the Basis of Multi-Scale Approach