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Чусов, Игорь Александрович

Организационные подразделения

Организационная единица

ИАТЭ НИЯУ МИФИ был образован в 1953 г. как вечернее отделение МИФИ. В 2009 г. ИАТЭ официально получил статус обособленного структурного подразделения НИЯУ «МИФИ», что дало новый мощный импульс для развития образовательной и научной деятельности на основе инновационной составляющей. В соответствии с лицензией Минобрнауки России ИАТЭ ведет образовательную деятельность в рамках очной, очно-заочной и заочной форм обучения. В настоящее время в ИАТЭ НИЯУ МИФИ осуществляется подготовка по очной форме обучения: бакалавриат- 16 направлений, специалитет – 4 направления, магистратура- 12 направлений; по очно-заочной: бакалавриат- 4 направления, специалитет – 1 направление; по заочной: бакалавриат- 3 направления, специалитет- 2 направления; аспирантура – 18 направлений. В структуре ИАТЭ 9 факультетов: физико-энергетический, естественных наук, кибернетики, социально-экономический, медицинский, вечерний, заочного обучения, подготовительный, повышения квалификации и профессиональной переподготовки специалистов. Образовательный процесс обеспечивают 18 общеобразовательных и 22 выпускающие кафедры.

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Чусов

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Игорь Александрович

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Correlations for calculating the transport and thermodynamic properties of lead-bismuth eutectic
(2020) Chusov, I. A.; Pronyaev, V. G.; Чусов, Игорь Александрович; Obysov, N. A.; Novikov, G. Y.; Pronyayev, V. G.
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Relations for calculating the transport and thermodynamic properties of lead-bismuth eutectics
(2020) Chusov, I. A.; Pronyaev, V. G.; Evgenievich, N. G.; Altksandrovich, O. N.; Чусов, Игорь Александрович
© 2020 Obninsk Institute for Nuclear Power Engineering, National Research Nuclear University 'MEPhI'. All rights reserved.The eutectic lead-bismuth alloy has relatively recently begun to be used as a coolant primarily in nuclear transport power plants. At present, this eutectic is regarded as a possible coolant for promising new generation reactor plants. The generalized systematic analysis of the results of experiments to determine the transport and thermodynamic properties of lead-bismuth eutectics was carried out by P.L. Kirillov with his staff and published back in 1998. However, they do not include numerous experimental results of studies of Russian specialists, which are characterized by methodical thoughtfulness, use of modern instrumentation and careful data processing. The emergence of new data has led to the need to clarify and correct the existing computational relationships. Recommended ratios for calculating the thermodynamic and transport properties of lead-bismuth eutectics (44.5% Pb + 55.5% Bi) are presented: density, coefficient of dynamic viscosity, specific heat capacity, thermal conductivity coefficient, surface tension coefficient, specific electrical resistance and local sound velocity as a function of temperature. The mentioned relations are based on the calculated analysis of data given in 38 experimental works carried out in our country and abroad and published for the period from 1923 to 2015. The authors had information about 1085 experimental points, but only 1058 points were suitable for direct estimates. The main difficulty in data processing was that the experiments considered in the work were carried out at different times using various measurement methods, ununiform statistical processing methods, different degree of bismuth purity, etc. The basis of data evaluation methods was a modified method of least squares, which allowed taking into account the errors of the experimental data accepted for consideration. Error values of the proposed relations and temperature ranges of their applicability are given in this paper. The article is based on the results of the work of the Thermodynamic Properties Data Center (TsDTS IATE NIYaU MIFI) of Rosatom State Corporation.
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Study of the Stratification Effect at the Reactor Installation in the Smolenskaya Area
(2020) Nguyen, T. B.; Shelegov, A. S.; Chusov, I. A.; Шелегов, Алексей Сергеевич; Чусов, Игорь Александрович
© 2020, Springer Nature Singapore Pte Ltd.This paper studies the Stratification Effect (SE) in the cooling basin for the Smolensk Nuclear Power Plant (NPP) and its dependence on the climatic condition. We investigated how climatic condition is the cause of the SE and analyzed all variations of the SE via program complex ANSYS-CFX. The obtained results were then compared to each other to assess the consequences of the possible impact of SE on the economy of the power units in its normal operation mode. The results indicate that SE happens when there is no cooling evaporation on the surface of the reservoir, moreover, air velocity greatly impact the SE, if the wind flows faster, the SE is less likely to appear, as the result, NPP has the better net efficiency.
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Ontologies and databases on thermophysical properties of nuclear reactor materials
(2019) Chusov, I. A.; Kirillov, P. L.; Pronyaev, V. G.; Erkimbaev, A. O.; Чусов, Игорь Александрович
© 2019 Obninsk Institute for Nuclear Power Engineering, National Research Nuclear University 'MEPhI'. All rights reserved.The study is dedicated to the information technologies for storage, systematization and distribution of thermophysical data for nuclear power engineering. The general trend existing in the areas involving wide use of scientific data is the shifting from conventional databases to the development of a consolidated infrastructure capable of overcoming sharply growing volumes of scientific data with continuously increasing complexity of the data structure due to the expansion of the range of materials. The above infrastructure ensures interoperability, including data exchange and dissemination. The general principle of data management for thermophysical properties of the nuclear reactor materials based on the subject-oriented ReactorThermoOntology (RTO) is suggested in the present paper. The ontology includes a unified glossary of all concepts, expanded through logical connections and axioms. The suggested RTO ontology combines the terms typical for reactor materials, their characteristics, as well as all types of information entities, determining textual, mathematical and computer structures. In the coded form, the ontology becomes the control add-in that can integrate heterogeneous data. Its most important feature is the possibility of its permanent expansion, which is necessary with introduction of new materials and terms related to them, e.g. nanostructures characteristics. Beside the ontology, description of the reactor materials, the possible scenarios for the use of the ontology during the phases of design, operation and integration of autonomous resources, primarily databases, are examined in the paper. The use of Big Data technology with diverse variations of logical structures of the data is suggested as the most efficient tool for data integration. Joint use of the technologies which were applied separately before, such as exchange standard in the form of the structured text documents, data control based on the ontology and platform for the work with big data, allows the conversion of multiple existing primary resources (databases, files, archives, etc.) to the standard JSON text format for the subsequent semantic integration.
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Ontologies and databases on thermophysical properties of nuclear reactor materials*
(2019) Chusov, I. A.; Kirillov, P. L.; Chusov, I. A.; Чусов, Игорь Александрович
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RESULTS OF VALIDATION AND CROSS-VERIFICATION OF THE ROK/B DESIGN CODE ON THE PROBLEM OF LOSS OF COOLING IN THE SPENT FUEL POOL РЕЗУЛЬТАТЫ ВАЛИДАЦИИ И ПЕРЕКРЕСТНОИ ПРОВЕРКИ РАСЧЕТНОГО КОДА РОК/Б НА ЗАДАЧЕ ПОТЕРИ ОХЛАЖДЕНИЯ В БАССЕИНЕ ВЫДЕРЖКИ
(2021) Mansurovich, S. R.; Evgenievich, K. V.; Evgenievich, S. O.; Matatovich, B. M.; Chusov, I. A.; Чусов, Игорь Александрович
© 2021 Obninsk Institute for Nuclear Power Engineering, National Research Nuclear University 'MEPhI'. All rights reserved.The procedures of validation and cross-verification of the newly developed computational code ROK/B are described. The main problem solved using the ROK/B code is the substantiation by calculation of the coolant density in the spent fuel pool (SFP) (untight reactor) and the temperature regime of the fuel assemblies during a protracted shutdown of the cooling systems (break in the supply of cooling water). In addition to the above, it is possible using the computational code ROK/B to carry out calculation of an accident with the discharge of the coolant from the SFP with simultaneous long-duration shutdown of the cooling systems. The ROK/B computational code allows carrying out calculations for various types of designs of the fuel assemblies and VVER reactors, in particular, VVER-1000, VVER-1200 and VVER-440 power units with single- and two-tiered fuel assemblies arrangement, with clad pipes in racks (for compacted assemblies storage) and pipes without cladding, with cased assemblies and caseless ones. During fuel reloading, a high level of the coolant is maintained, which makes it possible to do “wet” transportation of the assemblies from the reactor to the SFP. The mathematical model for heat and mass transfer calculation, including the boiling coolant model, implemented in the ROK/B code, includes: the motion equation, equations for calculating the enthalpy along the height of the fuel section of a fuel assembly with natural circulation of coolant within the channel containing the fuel assembly (lifting section) and in the inter-channel space (lowering section), the equation of mass balance between the channels of the racks with assemblies and in the inter-assembly space and the amount of evaporated (and outflowed) water, the heat balance equation for a fuel rod in a steam environment. The system of equations is supplemented by closing relations for calculating the thermal physics properties of water and steam, fuel and cladding materials, as well as the coefficients of heat transfer from the wall to the steam, hydraulic resistance and density of the steam-water mixture in the channels, and the heat released in the reaction of steam with zirconium. Validation of the computational code was carried out on the basis of the data of the ALADIN experiment performed by German specialists and the data of OKB Gidropress JSC. Cross-verification of the ROK/B computational code was carried out in comparison with the results of calculation using the KORSAR/GP and SOCRAT/B1 codes. Based on the results of the validation, it was concluded that the deviation of the ROK/B results from the experimental data is not more than 2 - 10% (10% for the option with a fuel rod power of 20 W). Based on the results of cross-verification, it was concluded that the discrepancy between the ROC/B results and the calculation results for the KORSAR/GP and SOCRAT codes is not more than 0.5% (for SOCRAT/V1) and less than 10% (for KORSAR/GP).
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Results of validation and cross-verification of the ROK/B design code on the problem of loss of cooling in the spent fuel pool
(2022) Sledkov, R. M.; Karnaukhov V. Y.; Stepanov, O. Y.; Bedretdinov, M. M.; Chusov, I. A.; Чусов, Игорь Александрович
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CALCULATED RATIOS FOR DETERMINING THE LITHIUM COOLANT THERMODYNAMIC AND TRANSPORT PROPERTIES РАСЧЕТНЫЕ СООТНОШЕНИЯ ДЛЯ ОПРЕДЕЛЕНИЯ ТЕРМОДИНАМИЧЕСКИХ И ТРАНСПОРТНЫХ СВОИСТВ ЛИТИЕВОГО ТЕПЛОНОСИТЕЛЯ
(2022) Chusov, I. A.; Babaeva, Yu. A.; Novikov, G.E.; Чусов, Игорь Александрович
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A Model of the Coolant Flow in Supercritical Nuclear Reactors Based on the Highest Approximations of the Chapman Enskog Method МОДЕЛЬ ТЕЧЕНИЯ ТЕПЛОНОСИТЕЛЯ В ЯДЕРНЫХ РЕАКТОРАХ СО СВЕРХКРИТИЧЕСКИМИ ПАРАМЕТРАМИ НА ОСНОВЕ ВЫСШИХ ПРИБЛИЖЕНИИ МЕТОДА ЧЕПМЕНА-ЭНСКОГА
(2023) Chusov, I. A.; Чусов, Игорь Александрович