Персона: Слободчук, Виктор Иванович
Загружается...
Email Address
Birth Date
Научные группы
Организационные подразделения
Организационная единица
Институт ядерной физики и технологий
Цель ИЯФиТ и стратегия развития - создание и развитие научно-образовательного центра мирового уровня в области ядерной физики и технологий, радиационного материаловедения, физики элементарных частиц, астрофизики и космофизики.
Статус
Фамилия
Слободчук
Имя
Виктор Иванович
Имя
7 results
Результаты поиска
Теперь показываю 1 - 7 из 7
- ПубликацияТолько метаданныеPossibility of simulating natural circulation in fast neutron reactors using a light water test facility(2021) Slobodchuk, V. I.; Uralov, D. A.; Avramova, E. A.; Слободчук, Виктор Иванович
- ПубликацияТолько метаданныеPossibility of simulating natural circulation in fast neutron reactors using a light water test facility(2021) Slobodchuk, V. I.; Uralov, D. A.; Avramova, E. A.; Слободчук, Виктор Иванович© 2021 Obninsk Institute for Nuclear Power Engineering, National Research Nuclear University 'MEPhI'. All rights reserved.The paper evaluates the possibility of modeling the heat transfer phenomena in a liquid-metal coolant using a light water test facility. It considers the natural circulation of the coolant in the upper plenum of the fast-neutron reactor. The sodium-cooled BN- 1200 reactor was selected as the reactor installation to be modeled. The development of novel reactor designs must be based on the results of experimental studies. Some problems of modeling thermohydraulic processes in BN type reactors are studied by using sodium test facilities. Experimental studies of natural convection processes using light water test facilities can be considered as a good alternative to those using sodium test facilities. To validate the model, the similarity theory and the «black box» method were used and their principles and applicability were analyzed. Using the «black box» method makes it possible to avoid detailed modeling of such components as the reactor core and heat exchangers, replacing them by a simplified representation of these components to simulate the integral characteristics of the existing real life equipment. The paper considers the basic criteria which determine the similarity of the thermohydraulic processes under study. The governing criteria of similarity were estimated based on the fundamental differential equations of natural convection heat transfer. Based on these criteria, a set of dimensionless values was obtained which show the correlation between the model parameters and the characteristics of the reactor facility. Besides, generalized relationships were derived which can be used to estimate the scaling factors for calculating the key values of the reactor facility based on the model parameters. These relationships depend on the thermal-physics parameters of the working fluids, the geometrical scale value and the ratio of the thermal power of the model to that of the reactor facility, i.e., model-to-reactor thermal power ratio. The conditions under which it is possible to model sodium coolant by light water with adequate accuracy were analyzed. An example is given of the numerical values of the scaling factors for one of the reference light water test facilities. The paper uses the experience of a number of foreign researchers in this field, in particular, the accepted assumptions which do not result in serious loss in modeling accuracy. According to the available estimates, the assumptions used do not result in considerable losses in accuracy. Thus, the natural circulation of the sodium coolant in the upper plenum of the fast-neutron reactor can be simulated with adequate accuracy by using light water test facilities.
- ПубликацияОткрытый доступПредварительная оценка характеристик оборудования цикла Брайтона на сверхкритических параметрах СО2(2023) Шелегов, А. С. ; Соловьев, Д. С. ; Лескин, Сергей Терентьевич; Слободчук, Виктор ИвановичПредставлен предварительный расчет основных теплотехнических и геометрических параметров проточной части турбины на сверхкритическом диоксиде углерода. Турбины такого типа предполагается использовать в цикле Брайтона тепловой схемы энергоблока ЯЭУ с жидкометаллическим теплоносителем, термический КПД которого может достигать 50%. Расчет проточной части турбины проведен с использованием общепринятых в практике проектирования газовых и паровых турбин. В результате расчета получены основные характеристики ступеней новой турбины. Рассмотрены возможные схемы компоновки основного тепломеханического оборудования применительно к энергоблоку с реактором типа БН-1200М. На основании выполненных предварительных расчетов получен ряд определяющих параметров рабочего цикла применительно к использованию на АЭС с перспективными реакторными установками с жидкометаллическим теплоносителем. Выполнена первичная оценка характеристик основного теплообменного оборудования цикла.
- ПубликацияТолько метаданныеTemperature conditions in the RBMK spent fuel pool in the event of disturbances in its cooling mode(2021) Hakobyan, D. A.; Slobodchuk, V. I.; Слободчук, Виктор Иванович
- ПубликацияТолько метаданныеCOMPUTATIONAL ANALYSIS OF THE POWER CONVERSION LOOP OF A NUCLEAR POWER PLANT UNIT WITH THE CLOSED S-CO2 BRAYTON CYCLE РАСЧЕТНО-АНАЛИТИЧЕСКИИ АНАЛИЗ КОНТУРА ПРЕОБРАЗОВАНИЯ ЭНЕРГИИ ЭНЕРГОБЛОКА АЭС С ЗАМКНУТЫМ ЦИКЛОМ БРАИТОНА НА S-CО2(2022) Leskin, S. T.; Slobodchuk, V. I.; Shelegov, A. S.; Лескин, Сергей Терентьевич; Слободчук, Виктор Иванович; Шелегов, Алексей Сергеевич
- ПубликацияТолько метаданныеTEMPERATURE CONDITIONS in the RBMK SPENT FUEL STORAGE POOL in the EVENT of DISTURBANCES in ITS COOLING MODE ТЕМПЕРАТУРНЫИ РЕЖИМ В БАССЕИНЕ ВЫДЕРЖКИ РБМК ПРИ НАРУШЕНИИ УСЛОВИИ ЕГО ОХЛАЖДЕНИЯ(2020) Slobodchuk, V. I.; Hakobyan, D. A.; Слободчук, Виктор Иванович© 2020 Obninsk Institute for Nuclear Power Engineering, National Research Nuclear University 'MEPhI'. All rights reserved.The problems of reprocessing and long-term storage of spent nuclear fuel (SNF) at nuclear power plants with RBMK-type reactors have not been fully resolved so far. As a result, nuclear power plants are forced to search for new options of temporary storage of SNF. One of the possible temporary solutions to the problem is compacted SNF storage in the reactor spent fuel storage pool (SFSP). As the number of spent fuel assemblies (SFA) in the SFSP increases, a greater amount of heat is released. In addition, it is necessary to take into account the fact that provision should be made for some additional storage space in the pool where fuel assemblies can be placed in an emergency situation. The paper presents the results of a numerical simulation of the temperature regime in the spent fuel storage pool both for storage of compacted SFAs and for emergency unloading of fuel assemblies. Several types of disturbances in the normal SFSP cooling mode are considered, including partial loss of cooling water and uncovering of SFA. The simulation was performed using the ANSYS CFX code. The time it takes for the water to reach the boiling point is estimated, as well as the time over which the fuel cladding is heated to 650°C. The most critical conditions are observed in the compartment for emergency unloading of fuel assemblies. The results obtained make it possible to estimate the time that personnel have to restore the cooling mode of the spent fuel storage pool before the maximum water and SFA temperature is reached.
- ПубликацияОткрытый доступPRELIMINARY ASSESSMENT OF THE CHARACTERISTICS OF BRAYTON CYCLE EQUIPMENT AT SUPERCRITICAL CO2 PARAMETERS ПРЕДВАРИТЕЛЬНАЯ ОЦЕНКА ХАРАКТЕРИСТИК ОБОРУДОВАНИЯ ЦИКЛА БРАИТОНА НА СВЕРХКРИТИЧЕСКИХ ПАРАМЕТРАХ СО2(2023) Leskin, S. T.; Slobodchuk, V. I.; Shelegov, A. S.; Soloviev, D. S.; Лескин, Сергей Терентьевич; Слободчук, Виктор Иванович; Шелегов, Алексей Сергеевич