This paper presents an analysis of the existing prerequisites and opportunities for the start-up of the Republic of Kazakhstan’s own full-fledged Nuclear Energy Program, a conceptual vision of the development of the Kazakhstani nuclear industry. Recommendations have been formulated for the timely and successful implementation of the program for the introduction of nuclear power plants into the country’s energy generation structure, the development of the necessary related nuclear infrastructure facilities, national legislation, and other issues directly related to nuclear power development.

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...

A large volume of scientific and research activities is currently carried out in substantiation of new projects on nuclear reactor safety, in which high emphasis is placed on “beyond-design basis” accidents, in spite of the fact that applied technical solutions reduce the probability of such accidents occurrence. To study parameters of severe accidents and develop measures to mitigate consequences an experimental base is necessary, which enables emergency situation modelling maximally close to real conditions. The National Nuclear Center (NNC) – the leading research organization in the atomic sphere in the Republic of Kazakhstan – possesses such an experimental base, providing scientific and technical support for peaceful uses of nuclear power, involving studies oriented towards the improvement of nuclear reactor safety [1]. The article presents the concept and general results of researches carried out on NNC’s experimental base in substantiation of nuclear reactors safety, realized in collaboration with JAEA, which is a key foreign partner in this area, as well as current plans on new experimental program implementation.

In support of global efforts to strengthen the nuclear non-proliferation regime, the IVG.1M research water-cooled thermal reactor at the National Nuclear Center of the Republic of Kazakhstan was successfully converted to low-enriched uranium (LEU, 19.75% ²³⁵U) fuel in 2023. The reactor’s operability with innovative bimetallic, fiber-type, dual-blade LEU fuel rods was experimentally verified during power start-up experiments. The test program included investigations of power distribution in the core, evaluation of temperature, power, and hydrodynamic reactivity effects, and the measurement of fission product release to the coolant. The results were in good agreement with safety calculations, confirming that the enrichment reduction did not degrade reactor performance characteristics. It was shown that the power reactivity effect increased by more than 1.5 times at a power level of 9 MW. The measured temperature reactivity coefficient (≈0.021 β_eff/°C) and the level of fission product release remained within acceptable and expected limits.