Investigation of the small break conditions in the primary circuit of a VVER-1000 reactor
Modes with violation of the reactor facility cooling conditions in the VVER reactor primary circuit have been simulated using the TRAC-PD2 and Open FOAM thermohydraulic codes [1 – 3] based on energy and mass conservation equations for a three-dimensional unsteady flow of a two-phase mixture. Coupled simulation of the dynamics of neutronic and thermohydraulic processes [1 – 8] aims to improve the qualitative understanding and the quantitative notion of their effects on safety.
Studying these modes using the above thermohydraulic codes makes it possible to analyze the course of transients and certain emergency processes without using the industrial testing method, this providing the basis for solving the problems of ensuring the reliability, operational safety and efficiency of nuclear power plants.
A modern nuclear reactor is a complex system for studying and calculating which it is not enough to use simple theoretical models. Thermohydraulic calculations are an essential component of most engineering and technological development works in nuclear power. Since, in conditions of an NPP, it is not possible to use a technologically conventional way to verify and update the results and findings of an a priori analysis on the basis of industrial tests, investigations based on codes are used in some cases as the tools for studying and predicting the parameters of thermohydraulic processes in the reactor’s circulation circuit.
The main purpose of the study is to calculate and investigate, with the use of codes, modes with violation of the reactor facility cooling conditions in the primary circuit of a VVER reactor in order to determine the conformity of the calculated parameters to the acceptance criteria established by regulatory documentation.
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