Design of a nuclear reactor core melt catcher on the basis of zirconia concrete.
Fundamental Research of Materials Structure and Properties Changes Resulted from Irradiation by Means of Complex of Modern Physical Methods
Methodical Development for VVER Reactor Pressure Vessel State and Safe Operation Lifetime Evaluation by Fracture Mechanics Criteria
Methodical Development for WWER-1000 Reactor Pressure Vessel Safe Operation Lifetime Evaluation Allowing for Anticorrosive Cladding
Phase Diagrams for Multicomponent Systems Containing Corium and Products of its Interaction with NPP Materials (CORPHAD)
#1950.2Phase Diagrams for Corium
Phase Diagrams for Multicomponent Systems Containing Corium and Products of its Interaction with NPP Materials
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
MATINE - Study of Minor Actinide Transmutation in Nitrides: Modelling and Measurements of Out-of-pile Properties
Development of Analytical Methods of Hydrogen and Helium Embrittlement of Nuclear Reactor Materials and Containers for Storage and Transport of Radioactive Materials
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
Neutron-Diffraction Study of Micro- and Macrostresses in Structural Ageing Alloys for Nuclear Power Engineering after Thermal and Radiation Exposure and Predicting Resistance to Radiation-Induced Swelling
Material Science Work Package for Lifetime Extension of VVER-1000 Reactor Pressure Vessels (RPV) from High Nickel Materials
Complete Data Library on Nucleon-Induced Fission Product Yields for Applications in Wide Energy Region
Development and Creation of Reactor Materials Containing Isotope Modified Boron Enriched with Boron-11 Isotope.
Investigation of Destruction Mechanisms in Reactor Steels and Alloys under Cycling Deformation
Development of Structural Materials Testing Complex Located at the Former Semipalatinsk Test-Site. Project SMTC-STS.
Recycling Technology of Irradiated Beryllium
Study of Structural Materials of the Nuclear Reactor WWR-K in View of its Recommissioning
Application of Structural Materials Data from the BN-350 Fast Reactor to Life Extension of Light Water Reactors
Behavior of Nuclear Reactor Fuel Assembly Materials during Their Long-Term Dry Storage
The International Science and Technology Center (ISTC) is an intergovernmental organization connecting scientists from Kazakhstan, Armenia, Tajikistan, Kyrgyzstan, and Georgia with their peers and research organizations in the EU, Japan, Republic of Korea, Norway and the United States.
ISTC facilitates international science projects and assists the global scientific and business community to source and engage with CIS and Georgian institutes that develop or possess an excellence of scientific know-how.