ANALYSIS OF MODELS OF AN AUTOMATIC POWER CONTROL SYSTEM FOR A PRESSURIZED WATER REACTOR IN DYNAMIC MODE WITH A CHANGE IN THE STATIC CONTROL PROGRAM

Abstract

The process of decarbonization of the industry was intensified by the decision to include nuclear energy in the EU Green Taxonomy. The increase in the capacity of nuclear power plants and the attractiveness of this industry for a long time causes interest in the use of nuclear power plants for electricity dispatching. Thus, the development of an automatic control system for changing the power of nuclear power plants in dynamic modes becomes relevant. The energy release along the height of the reactor core under the action of xenon fluctuations is one of the problems in ensuring the safe operation and stability of the NPP power unit when the power level changes. Violation of the energy release safety criteria occurs when non-stationary transient xenon processes with positive feedback occur in the core. Within the framework of the automatic control theory, the modeling of the system and the control object is compiled on the basis of the use of physical and mathematical models. Transfer functions, structure and parameters of the control system are determined. As an alternative, it is proposed to use an approximation model of the process in the core. The model is presented on the basis of transformations in such a way that the results of the calculation on a certain interval coincide or are close to the results of the calculation when using the analytical model. Such a model can approximate a process that depends on two parameters: time and degree of load change. The results of changing the regulation of technological parameters by the automated power control system of nuclear power unit in dynamic mode, the model of which is based on an approximation or physical and mathematical model, were analyzed. The results of changing the parameters are obtained by comparing the results during the transition of control from one static program to another. Switching between static control programs was studied for such process parameters as coolant temperature at the entrance to the reactor core, average coolant temperature and steam pressure in the secondary circuit. The studies were carried out under the condition that the reactor power was reduced to 80 % during the operation of one control program and increased to 100% of the power when another static program was used in reverse.