This paper presents the results of a calculation code approach providing a solution to the point kinetics problem for the IVG.1M research reactor of the National Nuclear Center of the Republic of Kazakhstan and allowing the simulation of dynamic processes going on during reactor start-ups, including changes in the thermal state of all its elements, reactor regulator displacement, accumulation of absorbers in the fuel, and the beryllium reflector. A mathematical description of the IVG.1M point kinetics model is presented, which provides a calculation of the reactor neutron parameters, taking into account the dependence of reactivity effects on the temperature, changes in the isotopic composition of materials, and thermal expansion of core structural elements. An array of data values was formed of reactivity added by separate elements of the core when changing their thermal state and other reactor parameters, as well as an array of data with the parameters of heat exchange of coolant-based reactor structural elements. These are used in the process of solving the point kinetics problem to directly replace formal parameters, eliminating the need to calculate the values of these parameters at each calculation step. Preliminary calculations to form an array of values of reactivity effects was applied to the reactor by separate structural elements when their temperature changes were performed using the IVG.1M precision reactor calculation model. The model was validated by the reactor parameters in the critical state. Preliminary calculations to form an array of data with the parameters of heat exchange of coolant-based reactor structural elements were performed in ANSYS Fluent software using the calculation model that describes the IVG.1M reactor fuel element in detail. Validation of the developed calculation code based on the results of two start-ups of the IVG.1M reactor was performed and its applicability for the analysis of transient and emergency modes of reactor operation and evaluation of its safe operation limits was confirmed.

Реактор ВВЭР (водо-водяной энергетический реактор) относится к типу PWR (реактор с водой под давлением), который широко применяется в мировой атомной энергетике благодаря своей надежности и высоким эксплуатационным характеристикам. В данном исследовании используется компьютерный код MCNP6 для разработки нейтронно-физической модели активной зоны реактора ВВЭР-1000 и проведения расчётов. Основной задачей исследования является воспроизведение ключевых характеристик реактора с высокой степенью точности. Проведены расчеты коэффициента размножения нейтронов и процесса выгорания топлива в активной зоне реактора. Особое внимание уделено расчетам выгорания с учетом реального профиля мощности во время первого топливного цикла. На основе расчетов определены изменения атомной концентрации 235U и 238U в различных зонах активной области. Также исследована динамика концентрации изотопов плутония, ксенона и самария для топливной сборки, что дает возможность оценить их влияние на нейтронно-физические характеристики реактора. Выполнено сравнение полученных результатов с расчетными и экспериментальными данными из литературных источников, что подтверждает их достоверность.

Experimental studies of the process of fuel destruction at the IGRY reactor make it possible to obtain data on the processes occurring in the fuel assemblies of various types of nuclear reactors during severe accidents. One of the stages of severe accidents is the melting of nuclear fuel. One of the stages of severe accidents is the melting of nuclear fuel. In this paper, we consider the complex application of physical models of the ANSYS FLUENT software module to achieve explicit dynamic melting of elements of a fuel assembly. Computational studies are carried out on a two-dimensional model of an experimental reactor device. The calculation models «volume of fluid» and «solidification melting» are used, which allow modeling the movements of the liquid fraction of the elements of the fuel assembly. To set the energy release, a user defined function (UDF) of the Fluent module is used.

For the safety assessment of existing reactor systems and the development of advanced nuclear technologies, it is essential to have experimental data on how nuclear fuel and structural materials respond under rapid transient conditions . Pulsed research reactors provide unique capabilities for reproducing short-duration power excursions with high neutron fluxes and rapid energy deposition. Many existing pulsed facilities were commissioned several decades ago and are approaching the limits of their experimental capabilities, highlighting the need for next-generation pulsed reactors. This paper reviews the principal concepts of pulsed research reactors and the physical mechanisms employed to generate intense neutron pulses. The characteristics of IBR-2M, self-quenching solution reactors, TRIGA, TREAT, the BIGR, and the IGR are analyzed to identify their advantages, limitations, and fields of application. The focus is on U-Gr reactors and the historical MIGR concept, which represented an early approach to increasing the performance of graphite pulsed reactors. The proposed large-scale uranium–graphite pulsed reactor is then assessed using preliminary neutronic and thermal results, with an emphasis on experimental channel dimensions, thermal neutron flux, and fluence in the self-quenching pulse and regulated transient modes.