Our work is related to analytical and numerical modeling of the materials behavior, structures and processes to ensure their safe operation. We continuously improve our models to increase their accuracy and initiate new searches projects to solve non-standard engineering tasks.
Brittle fracture probabilistic assessment of WWER-1000 RP
The investigations are aimed at developing a methodology for probabilistic assessment of the brittle fracture of the WWER-1000 RPVs. Main attention is focused on the definition of the stochastic input data: fracture toughness, CTB, size and shape of the defects.
The swelling of core baffle simulation
The main factor of limitation the operation of NPP's units after designed term is the phenomenon of radiation swelling of pressure vessel internals (PVI) under the influence of the flow of high-energy neutrons. Models of radiation processes of swelling and creep are very sensitive to the temperature distribution in the metal. The value of swelling changes by 2.5 times if the change of temperature equal to 30 degrees
Reactor dynamic model
Reactor dynamic model development is necessary for analyzing seismic stability, vibration and dynamic response of the reactor at the maximum design accident as it allows us to analyze the "whole system", to evaluate the strength of the support elements and the possibility of collision of individual parts of the model. IPP-Center LLC created a model of WWER-1000 reactor for analyzing the interaction between individual elements, taking into account their rigidity and contact non linearity.
The complex scientific-research tasks on the structural strength of the weld joint node of the collector with the nozzle in the steam generator PGV-1000M (welded joint #111)
Damage in the zone of welded joint #111 (WJ #111) significantly reduces the lifetime of the steam generators of the nuclear power plant. Taking into account numerous studies in this direction, the problem of the initiation and propagation of cracks in the WJ #111 remains relevant to the present
Method of estimation of structural reliability of heat exchangers tubes of steam generators
Steam Generators (SG) of the Ukrainian WWER-1000 type nuclear power plants (NPP) have been in operation for almost 30 years. And now, the main practical problem related to the SG is the leakage of heat exchanger tubes (HET).