Персона:
Волков, Юрий Николаевич

Организационные подразделения

Организационная единица

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.

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Волков

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Юрий Николаевич

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Теперь показываю 1 - 10 из 14

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Enhancing the efficiency of the MOX fuel cycle for VVER-1200 using burnable absorbers
(2023) Ozoani, J.; Volkov, Y.; Волков, Юрий Николаевич
Margin adoption in a nuclear power plant (NPP) design is a frequent approach to strengthen the design’s robustness and provide an efficient way to handle uncertainties. However, the current trend of increasing fuel enrichment, including the use of MOX fuel to achieve a higher burnup, leads to non-uniformity in the energy release (power peaking factor) at the level of the fuel rod lattice, thereby causing a great effect on the reactor margins. One of the ways to reduce the power peaking factor is the use of burnable absorbers (BAs) which helps to minimize the power peaking factor. This work aims at enhancing the efficiency of the MOX fuel cycle for VVER-1200 reactor by replacing the Gadolinium burnable absorber to Erbia burnable absorber.
Только метаданные
Axial Distributions of Fission Reaction Rates in the Annular Core of the ASTRA Critical Facility with Poison Profiling Elements in the Internal Reflector
(2021) Nevinitsa, V. A.; Boyarinov, V. F.; Fomichenko, P. A.; Zimin, A. A.; Volkov, Y. N.; Kruglikov, A. E.; Волков, Юрий Николаевич; Кругликов, Антон Евгеньевич
Computational modeling of experiments on measurement of axial distributions of fission reaction rates in configurations of the ASTRA critical facility simulating a HTGR with annular core is performed. Absorbing elements were installed in the internal reflector of critical assemblies to flatten the neutron flux radial distribution. Measurements of the fission reaction rates were performed both in a configuration with a flattened neutron field and in a configuration with control rods (CR) inserted in an external graphite reflector.
Только метаданные
Development of Reduced 3D Models for Neutronic Calculations of the ASTRA Critical Facility
(2022) Nevinitsa, V. A.; Fomichenko, P. A.; Boyarinov, V. F.; Zimin, A. A.; Kruglikov, A. E.; Volkov, Y. N.; Кругликов, Антон Евгеньевич; Волков, Юрий Николаевич
Только метаданные
Simulation-based studies on graphite absorption properties for ASTRA critical experiments
(2022) Nevinitsa, V. A.; Fomichenko, P. A.; Jitarev, V. E.; Shchurovskaya, M. V.; Kruglikov, A. E.; Volkov, Y. N.; Geraskin, N. I.; Щуровская, Мария Владимировна; Кругликов, Антон Евгеньевич; Волков, Юрий Николаевич; Гераскин, Николай Иванович
© 2022 Elsevier LtdThis work is focused on the improvement of the graphite impurity content characterization for the ASTRA critical facility. The equivalent boron content for the graphite reflectors of the ASTRA facility reported in the benchmark specifications was estimated using the results of the exponential experiments performed by graphite manufacturer. Because raw measured data are not available for these experiments, we tried to reassess the equivalent boron content for the ASTRA graphite reflectors using the results of analogous exponential experiments performed earlier on the empty RBMK critical facility at the National Research Center «Kurchatov Institute» and further simulation of the experiments utilizing the Monte Carlo code Serpent with ENDF/B-VII.0, ENDF/B-VII.1, and ENDF/B-VIII.0 cross-section libraries. The same grade graphite was used for manufacturing the graphite blocks for the RBMK critical facility and the ASTRA critical facility. The Monte Carlo simulations confirmed the accuracy of the exponential measurements performed for RBMK graphite characterization. It was also shown that the simulation must reproduce the graphite absorption cross-section estimated using the exponential experiments for the agreement between the measured and calculated distribution of thermal neutron flux. Therefore, the equivalent boron content should be assessed individually for each cross-section library used in the simulation. In case of the ASTRA facility the carbon capture cross section of 3.36 mb at 0.0253 eV (ENDF/B-VII.0) was used to deduce the equivalent boron content of 1 ppm wt from the measured results. Thermal capture cross-section of carbon was significantly increased in the ENDF/B-VII.1 and other recent cross-section libraries. Therefore, when calculating the ASTRA benchmark with these cross-section libraries the equivalent boron content reported in the benchmark specification should be reassessed.
Только метаданные
Boiling Water Reactor with a Combined Moderator for Utilization of Reactor Grade Plutonium
(2020) Zhdamirov, V. Y.; Naumov, V. I.; Volkov, Y. N.; Волков, Юрий Николаевич
© 2020, Pleiades Publishing, Ltd.Abstract: The article presents the concept of a channel-type reactor with a combined cooled moderator. The moderator consists of ceramic beryllium oxide and a liquid moderator—water with a possible variation in its composition due to the ratio between light water and heavy water components. Variants of fuel cycles based on the combination of “power” plutonium with depleted uranium, regenerated uranium, and thorium are analyzed. Quantitative estimates of the consumption of raw materials and the transformation of the isotopic composition of the irradiated fuel components are presented. The issues of safety and the possibility of controlling the reactivity coefficient by the density of the boiling coolant when changing the fuel cycle are discussed.
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Физические особенности режимов ускоренной утилизации оружейного плутония в реакторах типа ВВЭР
(НИЯУ МИФИ, 2015) Волков, Ю. Н.; Волков, Юрий Николаевич; Наумов, В. И.
Только метаданные
Comprehensive analysis of ASTRA benchmark with MCU-HTR and Serpent codes
(2024) Kruglikov, A. E.; Shchurovskaya, M. V.; Volkov, Yu. N.; Geraskin, N. I.; Кругликов, Антон Евгеньевич; Щуровская, Мария Владимировна; Волков, Юрий Николаевич; Гераскин, Николай Иванович
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Parallel calculation of the stationary diffusion equation
(2020) Zizin, M. N.; Soldatov, E. O.; Volkov, Yu. N.; Волков, Юрий Николаевич
© Published under licence by IOP Publishing Ltd.Considering the complex neutron-physical systems with the number of calculation points than 1 million, the calculation time can be an hour or more, and in case of non-stationary process with the small step of time, the waiting can be a month. The aim of the study is to develop effective techniques for the stationary computational modeling using parallelization technology to imply in non-stationary calculations. The parallelization scheme with height decomposition for stationary model was developed. The calculated speedup on the parallel part of the code are in good agreement with the linear dependence. The theoretical possible acceleration was derived using Amdal's law. Comparing the obtained and theoretical curves demonstrates that the developed parallelization scheme has less acceleration due to losses. The losses and their relative contribution were identified.
Только метаданные
Verification of the stationary module of the ShIPR software system for modelling experiments of the ASTRA HTGR type critical facility
(2020) Zizin, M. N.; Boyarinov, V. E.; Nevinitsa, V. A.; Fomichenko, P. A.; Volkov, Y. N.; Kruglikov, A. E.; Волков, Юрий Николаевич; Кругликов, Антон Евгеньевич
Coupled neutronic and thermal hydraulic calculation codes are verified for calculating the design of modern and prospective types of nuclear reactors. This verification is done by comparing experimental and calculated results for stationary and transient conditions. This paper presents ShIPR (Shell of Intelligent Package for Reactor) Integrated Development Environment with automatic generation of head programs based on the chain of computational modules. The aim of this study is to find the reason of a discrepancy in the modelling of sub-critical states that was found in previous work. The comparison of ShIPR stationary module with Monte-Carlo code (MCU) and experimental results on ASTRA HTGR critical facility was presented in the paper. To compare the detector readings and MCU calculation with the ShIPR module, the interpolation cross-section procedure was performed. This procedure allows simulating 235 U fission reaction rates (detector readings) in the complicate, annular core, using the micro-cross sections prepared by the cell-lattice code. We found that the calculation accuracy of stationary ShIPR module is on an acceptable level but the macro constants for control rods need to be prepared independently, given with surroundings.
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Computational analysis of experiments to determine the kinetic parameters at the ASTRA HTGR type critical facility
(2020) Zizin, M. N.; Boyarinov, V. F.; Nevinitsa, V. A.; Fomichenko, P. A.; Kruglikov, A. E.; Volkov, Y. N.; Кругликов, Антон Евгеньевич; Волков, Юрий Николаевич
© Published under licence by IOP Publishing Ltd.The zero power ASTRA critical facility is intended for experimental investigations of neutronic features of modular High Temperature Gas Cooled Reactors. The computational analysis of non-stationary experiments carried out at the ASTRA critical stand to determine the kinetic parameters is carried out using SHIPR diffusion program. To measure the kinetic parameters, the impulse Simmons-King method was used. The computational analysis were carried out for 4 different configurations of ASTRA critical stand. Since in case of nonstationary calculation by implementation of the spatial kinetics approximation, it is necessary to set the neutron group velocities the full-scale calculations of neutron spectra were carry oud using the MCU program in order to take into account the spectrum that is established in the full-scale reactor model. On the basis of a comparison of the experimental and calculated results, a conclusion was made about the applicability of the applied effective calculation models for neutronic calculations of HTGR-type reactors.