Simulation of complex dynamic processes in nuclear power plants. Development of basic program codes. Adaptation of the codes to multi-processor computers.
Partnership for basic research and education in nuclear reactor safety and novel application of transport theory.
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.
Experimental and calculation investigation to validate the concept of the reactor technology with ultimate neutronics and thermal-hydraulic characteristics.
Functional Training-Simulating Complex of the Research Reactor PIK (FTSC PIK).
The Benchmarks for Mathematical Simulation of Chernobyl-4 Accident
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
Monte-Carlo Code Development for Spatial Neutron Transient Calculations of NPP Core
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
Development of a Non-Destructive Method and Equipment for Determination of Welding Residual Stress on the Basis of Coherent Photonics and Computer Modeling
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
Fuel Assembly Tests under Severe Accident Conditions
Source Term Assessment at Ex-vessel Stage of Severe Accident
Material Science Work Package for Lifetime Extension of VVER-1000 Reactor Pressure Vessels (RPV) from High Nickel Materials
Investigation of Corium Melt Interaction with NPP Reactor Vessel Steel
Scale Experimental Investigation of the Thermal and Structural Integrity of the VVER Pressure Vessel Lower Head in Severe Accident
Study of Fuel Assemblies under Severe Accident Top Quenching Conditions in the PARAMETER-SF Test Series
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.