This thesis is focused on the actual scientific and technical problem of reliable determination of the safety parameters using NPP control systems based on the modern approaches to the simulation of the processes at the nuclear installations. Comprehensive analysis, systematization and detection of deficiencies in the determination of neutron-physical characteristics in NPP control systems have been carried out, advanced algorithms and software tools for determination of reactivity and reactor period have been developed and have been implemented in the following modern systems: equipment for neutron flux monitoring system (NFMS) AKNP-I (IF) VVER-1000, and VVER-440; automatic adjustment, control, operation and protection (AACOP) of VVR-M. Using MCNP program code, the models to describe signal shaping of AKNP ionization chambers (IC) and self-powered neutron detectors (SPND) for in-core reactor monitoring system (ICMS) were developed; based on these models the calculations were made and the following was determined: the contribution of the peripheral fuel elements of the fuel assemblies (FA) to the signal of ionization chambers, taking into account the effect of the concentration of boric acid and the coolant temperature on the IC signal, allowing for automatic correction of the determination of the power of WWER-1000; individual parameters of the burning out of the SPND, and also the formation of the signal of the SPND dependending on: the burn-out of the fuel, the concentration of boric acid and the coolant temperature. Based on the improved analytical model of dynamic processes in the core of the nuclear reactor, the calculation codes for the analysis of transient modes at VVER-1000 have been developed: accelerated uploaded protection actuation and termination of the last turbine steam driven operational pump. Experimental results of determination of reactivity parameters using analog and digital reactivitymeters for VVER-1000 are presented, as well as of the subcriticality of VVR-М on the basis of noise processing methods of a neutron detector signals. The model of the determination of most conservative assumptions in the analysis of nuclear safety of spent VVER fuel storage systems is considered. Validation of the code RELAP5 was carried out within the framework of standard problems of "Rivne NPP" and "Kozloduy NPP". The thesis contains the analysis of systems and models used for determination of safety indicators in the safety parameters representation system of VVER-1000, in the model of probabilistic safety analysis, etc. Scientific novelty of results, obtained by the researcher personally: 1) for the first time were performed systematization, analysis and generalization of the wide cycle results of computational and experimental research, conducted within the framework of scientific and methodological maintenance of NFMS and ICMS of VVER-1000; 2) for the first time software tools for calculating the reactor period and reactivity have been developed in SPDS of VVER-1000, NFMS of VVER-1000 and VVER-440, AACOP of VVR-M; 3) an analytical model for description of the neutron dynamics of a nuclear reactor is improved and scientifically substantiated, which differs from the existing models by taking into account feedback more correctly, which significantly increases the accuracy and precision of the results of calculations; 4) original method for correction of signals from NFMS ionization chambers was developed by taking into account the influence on the signal of the concentration of boric acid and temperature of the coolant, as well as other parameters, which allow to significantly improve the accuracy and precision of the determination of the VVER-1000 neutron power; 5) a fundamentally new calculation method was developed for the determination of the emitter burning out parameters individually for each SPND, which allows to improve the accuracy of the determination of the linear energy release of fuel elements, which is an important parameter for VVER safety. 6) a fundamentally new passive determination method of the moderator temperature coefficient of reactivity was developed. It is based on the utilization of noise analysis methods, which differs from existing by its operational efficiency, and also does not require any perturbation of the reactor coolant temperature; 7) for the first time the model of the determination of most conservative assumptions in the analysis of nuclear safety of spent VVER fuel storage systems was proposed and scientifically substantiated; Keywords: neutron flux monitoring system, reactivity and reactor period, self-powered neutron detectors, in-core monitoring system, in-core noise diagnostic system, neutronic parameters, nuclear reactor dynamic, safety parameters of nuclear installation, safety parameters display systems, subcriticality.
