The article is devoted to an issue of estimating the impurity gas amount in nuclear fuel in the aspect of the distracting contribution from released gases to the total pressure inside ampoule of the device in the simulating a severe accident with core melting. The paper presents a method based on measuring the pressure and temperature of gas in a closed values of the fuel elements during the fuel melting. The correctness of the developed methodology is confirmed by the results of experiments on the melting of fuel in a pulsed graphite reactor IGR with the implementation of a controlled neutron pulse.
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.
Развитие реакторов на быстрых нейтронах и атомной энергетики является актуальной задачей. Для разработки проектов в области ядерной энергетики необходимо проводить различные теплогидравлические расчеты. Использование результатов расчетов позволит своевременно проводить корректировку в проектировании, что резко повышает ответственность за надежность оборудования реактора.Данная работа посвящена исследованию процессов гидродинамики и теплообмена реактора на быстрых нейтронах электрической мощностью 600 МВт с объемным энерговыделением до 0,494 ГВт/м 3 .В данной работе приводятся результаты расчетов по гидродинамике и теплообмену в сегменте выделенной области ТВС быстрого натриевого реактора. В процессе работы создана 3D-модель выбранной области ТВС. Компьютерное моделирование проводилось в программном комлексе ANSYS FLUENT. Расчеты проводились с использованием турбулентной модели k-ε движения теплоносителя.Показаны неравномерности распределений температур по высоте активной зоны в различных областях ТВС, распределение скорости теплоносителя, а также показатели давления. Анализ полученных результатов показывает, что температуры конструктивных элементов не превышают допустимых температур, перепад давления значительно ниже, чем в реакторах другого типа.Ключевые слова: быстрые реакторы, теплообмен, теплоноситель, температура, тепловыделяющая сборка, перепад давления.The development of reactors on the fast neutrons and nuclear power engineering is generally responsible for its formation. One of them is that the responsibility for the reliability of the reactor equipment, its calculation, creation and operation sharply increases. For the development of projects in the field of nuclear energy, it is necessary to carry out various thermalhydraulic calculations. Using the results of calculations will allow for timely adjustment in the engineering.This work is devoted to the study of the processes of hydrodynamics and heat exchange of reactor on fast neutrons with an electrical power of 600 MW with a volumetric energy release up to 0.494 GW / m3.In the process of work, 3D model of the selected fuel assembly area was created in the program Gambit. Computer modeling was carried out in the ANSYS FLUENT software package as a result of which thermal state of fuel assembly for established mode of heat transfer was defined. The calculations were carried out using the k-ε coolant motion turbulent model. This article presents the results of calculations on hydrodynamics and heat transfer in a segment of the selected fuel assembly of a fast sodium reactor. The non-uniformity of temperature distributions along the height of the active zone in various areas of fuel assemblies, the distribution of the heat carrier velocity, as well as pressure indicators are shown. The analysis of the results obtained shows that the temperatures of the structural elements do not exceed the permissible temperatures; the pressure drop is significantly lower than in reactors of another type.
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