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.

In 2022, the physical start-up stage of the IVG.1M research reactor was successfully completed, initiated by reducing the fuel enrichment of ²³⁵U. This phase included the loading of nuclear fuel into the reactor core and conducting experiments to determine the neutron-physical characteristics of the reactor. Prior to the physical start-up, preliminary calculations were performed using the computational code MCNP6 and a full-scale model of the IVG.1M reactor with low-enriched uranium fuel (LEU). During the start-up series, the reactivity worth curves of the reactor control and protection system’s operating and compensating elements were determined. Additionally, experiments were performed to measure the reactivity effects of technological channel draining and to obtain activation reaction rates in the central experimental channel using nickel and gold activation indicators. The results of determining the neutron-physical characteristics of the IVG.1M reactor have confirmed the operability of the reactor core with LEU fuel.

The paper presents the results of a numerical interdisciplinary analysis of the fuel element of a heat capacity type research pulsed reactor (IGR). The fuel element of the IGR reactor is a graphite block impregnated with a solution of uranyl dinitrate. During the operation of the reactor, graphite blocks can be heated to high temperatures in a short period of time, which, together with the uneven impregnation of the block with uranium, which is due to the technological process of its manufacture, leads to the appearance of internal structural stresses. The purpose of this work is to make numerical estimation of the magnitude of thermal stresses arising in a graphite block. To carry out such an assessment, two computational models of a graphite block were built. One model is designed to perform neutron-physical calculations using a verified model of the IGR reactor core and the MCNP code, the other is designed to perform thermal strength analysis in the ANSYS software package. Thermal strength analysis includes two stages of calculations – thermal and structural (strength). Both models developed in the way to have the most similar topology, since this directly affects the correct distribution of the energy release over the volume of the block when transferring the results of the neutron-physical calculation to the thermal model. The operation of a graphite block as part of the reactor core was simulated during a start-up lasting 4 s at a stationary power of 2 GW, followed by cooling down for 5 s. Numerical values and distribution diagrams of temperature and stresses arising in the volume of the block are obtained. The results of the analysis confirm the effect of the uneven impregnation of the graphite block of the IGR reactor with a solution of uranyl dinitrate on its thermal strength characteristics. В работе представлены результаты численного междисциплинарного анализа топливного элемента исследовательского импульсного реактора теплоемкостного типа (ИГР). Топливный элемент реактора ИГР представляет собой графитовый блок, пропитанный раствором уранилдинитрата. Во время работы реактора графитовые блоки могут разогреваться до высоких температур за короткое время, что, в совокупности с неравномерностью пропитки блока ураном, которая обусловлена технологическим процессом его изготовления, приводит к появлению внутренних структурных напряжений. Целью данной работы является расчетная оценка величины термических напряжений возникающих в графитовом блоке. Для проведения такой оценки построены две расчетные модели графитового блока. Одна модель предназначена для выполнения нейтронно-физических расчетов с использованием верифицированной модели активной зоны реактора ИГР и кода MCNP, другая – для проведения термопрочностного анализа в программном комплексе ANSYS. Термопрочностной анализ включает в себя два последовательных расчета – тепловой и структурный (прочностной). Обе модели были разработаны таким образом, чтобы иметь максимально схожую топологию, поскольку это напрямую влияет на правильное распределение энерговыделения по объему блока при передаче результатов нейтронно-физического расчета в тепловую модель. Была смоделирована работа графитового блока в составе активной зоны реактора во время пуска продолжительностью 4 с на стационарной мощности 2 ГВт с последующим расхолаживанием в течение 5 с. Получены численные значения и эпюры распределения температуры и напряжений, возникающих в объеме блока. Результаты проведенного анализа подтверждают влияние неравномерности пропитки графитового блока реактора ИГР раствором уранилдинитрата на его термопрочностные характеристики. Бұл жұмыста ИГР зерттеу реакторының отындық элементіне жасалған дисциплина аралық сандық есептеу нәтижелері көрсетілген. ИГР реакторының отындық элементі бойына уранилнитрат ертіндісі сіңірілген графит блогы болып табылады. Реактордың жұмыс жасау барысында графиттік блок қысқа уақыт аралығында үлкен температураға дейін қызуы мүмкін. Бұл эффект, отындық элементтердің графит блогының жасалу технологиясының әсерінен пайда болған...