Institute of Atomic Energy performs activity to substantiate the establishment and development of national nuclear industry. Within this area, following works are carried out:
- Feasibility studies to prove NPP construction in Kazakhstan also under the remit of the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO);
- Development of legal and technical documents to ensure the formation of a comprehensive and self-sufficient regulatory and legal framework for the nuclear industry and nuclear power in Kazakhstan;
- Feasibility studies for further BN-350 spent nuclear fuel (SNF) management;
- Development of design for facilities and experimental devices under studies in support of safety of nuclear power plants.
EXPERIMENTAL STUDIES FOR NUCLEAR INDUSTRY SAFETY
Severe accident research facilities were used to carry the following activity:
- experiments to produce reactor core melt (corium) and its moving towards the tested objects;
- experiments to study interaction between the melt and water;
- experiments to study interaction between the melt and “dry” concrete trap;
- experiments to study interaction between the melt and reactor bottom material under supply of the water for cooling down outer surface of bottom mockup and simulation of residual heat in the melt;
- experiments to study interaction between the melt, water and concrete when the water coolant is supplied to the melt surface inside of the concrete trap and simulation of residual heat inside of the melt by means of electrical heaters.
Under the project, qualitative and quantitative database has been obtained, which is essential to predict the progress of severe accident and to formulate the measures limiting and isolating its consequences.
Currently, IGR reactor and EAGLE test bench is involved into experiments to substantiate the conception of controlled moving of fuel melt in fast reactors to prevent recriticality at severe accidents with core melting.
Out-of-pile studies for fast reactor safety include:
- experiments on producing melt with UO2 and corium simulator (Al2O3);
- experiments on moving UO2 corium through the discharge duct;
- experiments with aluminum oxide on moving corium simulator through the discharge duct filled with sodium;
- experiments on melt interaction with sodium coolant, including:
– melt fragmentation;
– melt cooling down in sodium pool.
In-pile experiments are implemented at IGR reactor using special experimental devices and methods.
The objective of INVECOR (IN-VEssel COrium Retention (INVECOR) project is improvement of substantiation of safe melt retention in the vessel of light water reactor in condition of severe accident through experimental modeling of thermal, physical and chemical processes during holding the pool of prototype corium melt at the bottom of water cooling reactor vessel.
The experiments implemented using LAVA-B facility, which is equipped by induction furnace to produce 60 kg of prototype corium melt and melt receiver (MR) containing water cooling model of a bottom of reactor pressure vessel, device for simulation of residual heat release with power up to 90 kW and a set of temperature, pressure and strain sensors.
To ensure temperature conditions for physical and chemical “corium/steel” interaction, heat insulation bag is used between cooling water and outer surface of vessel’s model.
The main outcomes of the Project are new experimental data on final structure of corium pool with natural convection when simulating residual heat release in a fuel and ablation of vessel’s bottom model at different melt composition and heat loads on a wall in 2-D configuration with real flection of reactor vessel bottom.
The overall goal of the CORMIT (Corium and Refractory Materials Interaction Test) Project is to prepare for and carry out experimental researches on corium interaction with refractory materials, which could be used as potential protective material for under-reactor melt trap. These researches enable to choose appropriate refractory coating material for under-reactor traps normally intended for enhancing safety of Japanese NPPs in case of severe accident with reactor core melting
Fukushima Debris Project
Fukushima Debris Project devoted to produce solidified melt of fuel and structural materials of reactor core and further study of their properties. Relevant data on the structure and physic-mechanical properties of ready corium prototype will enable to develop the procedure of debris removal from damaged Fukushima reactors and choose appropriate tools for this.
Currently, preparation of experimental program is carried out to test fuel of advanced ASTRID reactor (France) under emergencies, caused by the transient overpower in the fuel accompanied by the cessation of coolant flow through the core.
EXPERIMENTAL STUDIES OF NUCLEAR/THERMONUCLEAR STRUCTURAL MATERIALS
Within scientific, technical and innovative activity of the IAE, experimental material testing studies of nuclear/thermonuclear structural materials are carried out. The main objectives of these studies are the selection of existing, the creation and testing of new, with increased technological requirements, structural materials, the study of the properties of fuel and structural materials in unique operating conditions in nuclear and thermonuclear facilities
To implement these objectives in the IAE there is unique experimental equipment and material research laboratories (material testing complex) located at the Complex of Research Reactors Baikal-1.
The main tasks cover the following lines:
- study of the structure and composition of the power reactor core materials after electro-thermal or reactor melting;
- studying the condition of core melt interaction products of power reactors with coolant, reactor pressure vessel material and concrete;
- studying the physicochemical interaction of core components at their heating up to the melting point;
- hydrogen penetration of structural fusion materials;
- identifying the main parameters of hydrogen isotope interaction with advanced fusion reactor materials;
- studying the influence of gaseous medium and irradiation on the diffusive, sorption and physical and mechanical properties of structural materials;
- analysis of gas medium samples generated during experimental work;
- study of the changes of structural-phase composition, corrosion resistance and physical and mechanical properties of structural materials after radiation, thermal and mechanical-thermal impact;
- certification of IAE’s reactor complexes;
- determination of HEU and LEU fuel properties of enterprise research reactors;
- conduction of thermal gravimetric analysis and differential scanning calorimetry of constructional material in the different environment under the temperatures up to 1600 °С;
- study of spectral – luminescent characteristics of nuclear-excited plasma, formed with nuclear reaction products;
- creation of new materials of hydrogen energy.
The IAE has well-outfitted material testing complex able to solve above research tasks.