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#0067Simulation of Reactor Dynamics

Simulation of complex dynamic processes in nuclear power plants. Development of basic program codes. Adaptation of the codes to multi-processor computers.

#0115Reactor Kinetics

Partnership for basic research and education in nuclear reactor safety and novel application of transport theory.

#0116Verification of Reactor Data Bases (R)

Development of methodical and calculation technology verification of nuclear data bases used in the calculation of neutron-physical characteristics and in analysis of nuclear safety of reactor facilities and nuclear conversion technological processes.

#0134Transmutation in Fast Reactors

Experimental and calculation investigation to validate the concept of the reactor technology with ultimate neutronics and thermal-hydraulic characteristics.

#0321Research and Training Reactor PIK

Functional Training-Simulating Complex of the Research Reactor PIK (FTSC PIK).

#0683Benchmarks for Chernobyl Simulation

The Benchmarks for Mathematical Simulation of Chernobyl-4 Accident

#0815.2Codes for Nuclear Safety Evaluation

Development of Computerized Technology for Critically Safety Uncertainty Evaluation based on the Analysis of Data for the International Bank for Critical Experiments

#0909-2Two-Cascade Power Blanket

Study of Neutron Multiplication in Media for Creating a Frequency Two-Cascade Energy Blanket

#1086Monte-Carlo Nuclear Reactor Code

Monte-Carlo Code Development for Spatial Neutron Transient Calculations of NPP Core

#1950Phase Diagrams for Corium

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

#2894Non-Destructive Determination of Welding Residual Stress

Development of a Non-Destructive Method and Equipment for Determination of Welding Residual Stress on the Basis of Coherent Photonics and Computer Modeling

#2916Nuclear Fuel Behavior During Chernobyl Accident

Development of the Models for Nuclear Fuel Behavior During Active Phase of Chernobyl Accident

#2936Reactor Core Melting

Modelling of Reactor Core Behaviour under Severe Accident Conditions. Melt Formation, Relocation and Evolution of Molten Pool

#3194Fuel Assembly Under Severe Accident Conditions

Fuel Assembly Tests under Severe Accident Conditions

#3345Ex-Vessel Source Term ANalysis (EVAN)

Source Term Assessment at Ex-vessel Stage of Severe Accident

#3420VVER-1000 Reactor Pressure Vessel

Material Science Work Package for Lifetime Extension of VVER-1000 Reactor Pressure Vessels (RPV) from High Nickel Materials

#3592Corium Melt Interaction with Reactor Vessel Steel

Investigation of Corium Melt Interaction with NPP Reactor Vessel Steel

#3635VVER Vessel in Severe Accident

Scale Experimental Investigation of the Thermal and Structural Integrity of the VVER Pressure Vessel Lower Head in Severe Accident

#3690Fuel Assemblies under Severe Accident

Study of Fuel Assemblies under Severe Accident Top Quenching Conditions in the PARAMETER-SF Test Series

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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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