Международная конференция молодых специалистов, ученых и аспирантов по физике ядерных реакторов (ВОЛГА)
Постоянный URI для этого раздела
Обзор
Просмотр Международная конференция молодых специалистов, ученых и аспирантов по физике ядерных реакторов (ВОЛГА) по Issue Date
Теперь показываю 1 - 7 из 7
Количество результатов на страницу
Sort Options
- ПубликацияОткрытый доступInternational Conference for Young Scientists, Specialists, and Postgraduates on Nuclear Reactor Physics 2016 (ICNRP-2016)(IOP Publishing Ltd, 2017)We are pleased to introduce the Proceedings of the International research conference «International Conference for young scientists, specialists and post-graduates on Nuclear Reactor Physics 2016 (ICNRP-2016)» (5-9 September 2016, Health resort «Volga», Moscow, Russia) organized by the National Research Nuclear University MEPhI, with ROSATOM partnership. Representatives of research organizations and universities from twelve countries (Russia, Germany, Norway, Finland, Kazakhstan, Belarus, Italy, Slovakia etc.), delivered their presentations on various topics. The major topics are features of fast reactors, calculation for the needs of operation and design of nuclear reactors, computational reactor tests, codes and databases. Over a hundred people from 37 organizations attended the conference. More than 93 papers were presented. The received papers were reviewed according to the standards of the Journal of Physics: Conference Series and developed by the organizers' scientific criteria. This volume of the journal includes 65 papers devoted to various branches of nuclear reactor physics and technology. During the conference, various sports activities were held, as well as a workshop on the problems of nuclear education in Russia. Most of the participants, according to the results of the survey were satisfied and expressed a desire to take part in the next conference in 2018. The organizing committee is very grateful to the: • Participants of the conference for their valuable contribution with the delivered presentations and interesting papers, • Conference program committee chairman Strikhanov M.N., rector of National Research Nuclear University MEPhI, • Program committee co-chairs: Caruso G., professor, Sapienza University of Rome, Hascik J., professor, Technical University of Bratislava, Janardhanan N.K., assistant professor, Jawaharlala Nehru University, Pershukov V.A., deputy director general, Rosatom, Tikhomirov G.V., dean of Physical-technical faculty, NRNU MEPhI, Tikhomirov V.V., professor, Belarusian State University, • Competitiveness Program of National Research Nuclear University MEPhI for the financial support, • Personnel of Journal of Physics for the support and communication. Editor: Anton Smirnov Co-editors: Rynat Bahdanovich, Ekaterina Proshkina
- ПубликацияОткрытый доступPeer review statement(2017)All papers published in this volume of Journal of Physics: Conference Series have been peer reviewed through processes administered by the proceedings Editors. Reviews were conducted by expert referees to the professional and scientific standards expected of a proceedings journal published by IOP Publishing.
- ПубликацияОткрытый доступNeutronic calculation of fast reactors by the EUCLID/V1 integrated code(IOP Publishing Ltd, 2017) Koltashev, D. A.; Stakhanova, A. A.This article considers neutronic calculation of a fast-neutron lead-cooled reactor BREST-OD-300 by the EUCLID/V1 integrated code. The main goal of development and application of integrated codes is a nuclear power plant safety justification. EUCLID/V1 is integrated code designed for coupled neutronics, thermomechanical and thermohydraulic fast reactor calculations under normal and abnormal operating conditions. EUCLID/V1 code is being developed in the Nuclear Safety Institute of the Russian Academy of Sciences. The integrated code has a modular structure and consists of three main modules: thermohydraulic module HYDRA-IBRAE/LM/V1, thermomechanical module BERKUT and neutronic module DN3D. In addition, the integrated code includes databases with fuel, coolant and structural materials properties. Neutronic module DN3D provides full-scale simulation of neutronic processes in fast reactors. Heat sources distribution, control rods movement, reactivity level changes and other processes can be simulated. Neutron transport equation in multigroup diffusion approximation is solved. This paper contains some calculations implemented as a part of EUCLID/V1 code validation. A fast-neutron lead-cooled reactor BREST-OD-300 transient simulation (fuel assembly floating, decompression of passive feedback system channel) and cross-validation with MCU-FR code results are presented in this paper. The calculations demonstrate EUCLID/V1 code application for BREST-OD-300 simulating and safety justification.
- ПубликацияОткрытый доступInherent Safety Characteristics of Advanced Fast Reactors(IOP Publishing Ltd, 2017) Bochkarev, A. S.; Korsun, A. S.; Kharitonov, V. S.; Alekseev, P. N.; Бочкарев, Алексей Сергеевич; Харитонов, Владимир Степанович; Корсун, Александр СергеевичThe study presents SFR transient performance for ULOF events initiated by pump trip and pump seizure with simultaneous failure of all shutdown systems in both cases. The most severe cases leading to the pin cladding rupture and possible sodium boiling are demonstrated. The impact of various features on SFR inherent safety performance for ULOF events was analysed. The decrease in hydraulic resistance of primary loop and increase in primary pump coast down time were investigated. Performing analysis resulted in a set of recommendations to varying parameters for the purpose of enhancing the inherent safety performance of SFR. In order to prevent the safety barrier rupture for ULOF events the set of thermal hydraulic criteria defining the ULOF transient processes dynamics and requirements to these criteria were recommended based on achieved results: primary sodium flow dip under the natural circulation asymptotic level and natural circulation rise time.
- ПубликацияОткрытый доступThe concerted calculation of the BN-600 reactor for the deterministic and stochastic codes(IOP Publishing Ltd, 2017) Bogdanova, E. V.; Kuznetsov, A. N.; Богданова, Екатерина ВладимировнаThe solution of the problem of increasing the safety of nuclear power plants implies the existence of complete and reliable information about the processes occurring in the core of a working reactor. Nowadays the Monte-Carlo method is the most general-purpose method used to calculate the neutron-physical characteristic of the reactor. But it is characterized by large time of calculation. Therefore, it may be useful to carry out coupled calculations with stochastic and deterministic codes. This article presents the results of research for possibility of combining stochastic and deterministic algorithms in calculation the reactor BN-600. This is only one part of the work, which was carried out in the framework of the graduation project at the NRC "Kurchatov Institute" in cooperation with S. S. Gorodkov and M. A. Kalugin. It is considering the 2-D layer of the BN-600 reactor core from the international benchmark test, published in the report IAEA-TECDOC-1623. Calculations of the reactor were performed with MCU code and then with a standard operative diffusion algorithm with constants taken from the Monte - Carlo computation. Macro cross-section, diffusion coefficients, the effective multiplication factor and the distribution of neutron flux and power were obtained in 15 energy groups. The reasonable agreement between stochastic and deterministic calculations of the BN-600 is observed.
- ПубликацияОткрытый доступRadiogenic lead as coolant, reflector and moderator in advanced fast reactors(IOP Publishing Ltd, 2017) Kulikov, E. G.; Куликов, Евгений ГеннадьевичMain purpose of the study is assessing reasonability for recovery, production and application of radiogenic lead as a coolant, neutron moderator and neutron reflector in advanced fast reactors. When performing the study, thermal, physical and neutron-physical properties of natural and radiogenic lead were analyzed. The following results were obtained: 1. Radiogenic lead with high content of isotope 208Pb can be extracted from thorium or mixed thorium-uranium ores because 208Pb is a final product of 232Th natural decay chain. 2. The use of radiogenic lead with high 208Pb content in advanced fast reactors and accelerator-driven systems (ADS) makes it possible to improve significantly their neutron-physical and thermal-hydraulic parameters. 3. The use of radiogenic lead with high 208Pb content in advanced fast reactors as a coolant opens the possibilities for more intense fuel breeding and for application of well-known oxide fuel instead of the promising but not tested enough nitride fuel under the same safety parameters. 4. The use of radiogenic lead with high 208Pb content in ADS as a coolant can upgrade substantially the level of neutron flux in the ADS blanket, which enables effective transmutation of radioactive wastes with low cross-sections of radiative neutron capture.
- ПубликацияОткрытый доступМеждународная конференция молодых специалистов, ученых и аспирантов по физике ядерных реакторов «Волга-2024» (3-6 сентября 2024 г.)(НИЯУ МИФИ, 2024)Сборник тезисов включает доклады Международной конференции молодых специалистов, ученых и аспирантов по физике ядерных реакторов «Волга 2024», прошедшей 3–6 сентября 2024 г. на базе отдыха НИЯУ МИФИ «Волга» в Тверской области. Тезисы докладов специалистов атомной отрасли, аспирантов и молодых ученых российских и зарубежных университетов освещают актуальные аспекты разработки проектных кодов, эксплуатации и обеспечения безопасности ядерных реакторов. Особое внимание уделено перспективным технологиям и инновационным подходам к расчетному моделированию, комплексным расчетам и управлению ядерными знаниями. Представлены результаты исследований, направленных на повышение эффективности ядерных установок, замыкание ядерного топливного цикла и обеспечение кадрового потенциала отрасли. Сборник предназначен для специалистов, преподавателей, студентов и аспирантов, интересующихся перечисленными направлениями атомной отрасли. Внесены отдельные технические правки, однако, в основном, сохранена авторская редакция текстов. Ответственные редакторы: М.Ю. Захаров, М.А. Чубаров Тезисы получены до 30 июля 2024 года. Материалы издаются в авторской редакции.