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#0662Mechanical Characteristic of Cr-Ni Steels

Development of the Long Term prediction Methods for Stress Rupture Strength (SRS) of Chloronium-nickel Austenitic steel on the basis of Shortcut Tests.

#0721In-Pile Tests of Very High Burn-Up Fuel

In-Pile Tests of New Generation Fuels for VVERs of Different Purpose

#0730Molten Salt Fluoride Fuel

Study of Physico-chemical and Electrochemical Properties of the ADTT Fluoride Nuclear Fuel and Foundation of its Reprocessing

#0858Computer Complex for Research Reactors

Multifunctional Computer Complex for Research Reactors

#1195Dynamic Simulator for Power Complex

A Multipurpose Dynamic Electric System Simulator for the RIAR Power Complex

#1347Lithium Detector of Solar Neutrinos

Development of Lithium Detector of Solar Neutrinos on a Basis of Prototype with 300 kgs of Lithium

#1756Zirconium Alloys High-Frequency Treatment

Investigation of Reactor Irradiation Effect on Radiation Stability and Mechanical Properties of Articles from Zirconium Alloys after High-Speed High-Frequency Treatment

#2024Monograph on Boron in Nuclear Engineering

Preparation of the Monograph "Boron in Nuclear Engineering"

#0534.2Uranium Dioxide Properties

Investigation Properties of Uranium Dioxide Advanced and Mixed Uranium-Plutonium fuels (MOX) for Prediction of Performance and Operating Characteristics to Provide Certification Fuel and Verification of Computational Models for Parameters Definition of Ox

#2066Radiation Resistant Zirconium Alloys

Development of Scientific Fundamentals for Creation of Irradiation Resistant Structural Zr Alloys for Nuclear Reactor Cores

#2092Processing of Weapon Plutonium into MOX-fuel

Processing Compact Weapon Plutonium into MOX-fuel by Anhydrous (Thermochemical and Thermoelectrochemical) Methods for Thermal and Fast Reactors

#2320Tomography for Fuel Elements

Development of the Tomographic Methods of Testing of the Fuel and its Components Distribution in Fuel Elements before and after Irradiation

#2522Neutron Therapy with Cf Source

Development of Treatment Methods, Technical Means and Small Sized Cf-252 Sources for Neutron Brachytherapy

#2925Nuclear Data for Minor Actinides Transmutation

Measurement of Transmutation Properties of Minor Actinides Irradiated in Intermediate Reactor Neutron Spectrum

#3365Experimental Mixed-Oxide Fuel

Fabrication of Experimental CANDU MOX Fuel Pins for Irradiation Testing in the MIR Reactor

#3587Monograph on Fuel Elements

Preparation of the Monograph “Fuel Elements with Vibropacked Oxide Fuel”

#3608Minor Actinide Transmutation in Inert Matrices Fuels

MATINE 2- Study of Minor Actinide Transmutation in INErt Matrices Fuels: Modeling, Fabrication and Measurements of Out-of-Pile Properties

#3692Zr-based Alloys under irradiation

Fundamental Investigations of Processes, Responsible for Changes in Crystallographic Texture and Microstructure of Model Samples from Zr-Based Alloys under Neutron Irradiation

#3919Fission Products Release from Heated Fuel

Investigation of Fission Products Release from Uranium Dioxide at Heat up Under Oxidizing and Reducing Media. VERONIKA Project.

#3979Measurement of Surface Segregations in Structural Steels

Development of an Auger Spectroscopy Method for Measurement of Phosphorus Intergranular Segregations in PWR Reactor Pressure Vessel Steels

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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.

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