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Journal articles on the topic 'BREST-OD-300'

Author: Grafiati

Published: 3 June 2025

Last updated: 31 July 2025

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1

Lemekhov, V. V., A. V. Petrenko, and A. V. Yashkin. "Power unit with RP BREST-OD-300." Journal of Physics: Conference Series 1475 (March 2020): 012013. http://dx.doi.org/10.1088/1742-6596/1475/1/012013.

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2

Dubenkov, N. E., A. V. Proukhin, A. A. Bazhanov, et al. "Radiation Safety Validation of a BREST-OD-300 Installation." Atomic Energy 130, no. 4 (2021): 238–43. http://dx.doi.org/10.1007/s10512-021-00802-y.

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3

Дубенков, Н. Е., В. П. Васюхно, and Г. А. Хачересов. "POTENTIAL IODINE COMPOUNDS IN THE BREST-OD-300 REACTOR LEAD COOLANT." ЯДЕРНАЯ И РАДИАЦИОННАЯ БЕЗОПАСНОСТЬ, no. 1(99) (March 22, 2021): 5–13. http://dx.doi.org/10.26277/secnrs.2021.99.1.001.

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Одним из наиболее важных элементов при обосновании радиационной безопасности реактора БРЕСТ-ОД-300 является йод, поступающий в теплоноситель из топлива. Его последующий массоперенос зависит от физико-химических параметров соединения, в котором он находится. Для определения возможных соединений йода в свинце необходимо проведение теоретических и экспериментальных исследований как для системы «свинец – йод», так и для многокомпонентной системы с учетом других элементов, которые могут накапливаться в свинцовом теплоносителе при работе реакторной установки на мощности. Представлены результаты эксп

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Shadrin, A. Yu, K. N. Dvoeglazov, V. A. Kascheyev, et al. "Hydrometallurgical Reprocessing of BREST-OD-300 Mixed Uranium-plutonium Nuclear Fuel." Procedia Chemistry 21 (2016): 148–55. http://dx.doi.org/10.1016/j.proche.2016.10.021.

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5

Grabezhnaya, V., A. Mikheyev, A. Alekhin, A. Kryukov, and A. Tikhomirov. "EXPERIMENTAL JUSTIFICATION OF DESIGN CHARACTERISTICS OF STEAM GENERATOR RP BREST-OD-300." PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS 2021, no. 2 (2021): 218–35. http://dx.doi.org/10.55176/2414-1038-2021-2-218-235.

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The project BREST-OD-300 reactor plant (RP) with a fast neutron reactor and a lead coolant in the primary circuit is being developed in NIKIET JSC. As a steam generator (SG), a helical-type steam generator with coiled tubes with subcritical pressure water in the second circuit is considered. To substantiate the design characteristics of the secondary coolant at the State Research Center of the Russian Federation - IPPE, thermohydraulic tests of various SG models were carried out at the SPRUT stand Initially, tests were carried out on a model of a coiled steam generator consisting of two three-

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6

Karazhelevskaya, Y. E., M. A. Levon, A. M. Terekhova, and A. S. Zlobin. "Irregularity of plutonium isotopic composition of the BREST-OD-300 initial load." Journal of Physics: Conference Series 1689 (November 2020): 012049. http://dx.doi.org/10.1088/1742-6596/1689/1/012049.

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7

Dragunov, Yu G., V. V. Lemekhov, V. S. Smirnov, and N. G. Chernetsov. "Technical solutions and development stages for the BREST-OD-300 reactor unit." Atomic Energy 113, no. 1 (2012): 70–77. http://dx.doi.org/10.1007/s10512-012-9597-3.

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8

Solonin, V. I., A. N. Terekhin, and E. A. Shiverskiy. "Metal Liner Reliability Assessment for BREST-OD-300 Reactor Vessel Accounting for Brittle Fracture and Leaks." Herald of the Bauman Moscow State Technical University. Series Mechanical Engineering, no. 5 (128) (October 2019): 119–34. http://dx.doi.org/10.18698/0236-3941-2019-5-119-134.

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Multi-layer metal-concrete vessel of BREST-OD-300 reactor comprising metal liner that covers internal cavities and ducts has an original design with no known analogues. Due to this, statistical reliability assessment methods, based on operating or testing experience, are not applicable in this case. We propose a method to assess the reliability of the reactor vessel liner taking into account a random nature of loads and mechanical properties affecting brittle rupture and leakage probability. The assessment is based on numerical simulation of postulated defects growth, allowance for the probabi

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9

Davydov, V. K., A. P. Zhirnov, K. M. Kalugina, et al. "Reactivity Effects of Steam Bubbles Injected into the BREST-OD-300 Reactor Core." Atomic Energy 130, no. 3 (2021): 136–42. http://dx.doi.org/10.1007/s10512-021-00784-x.

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10

Smirnov, V. P., A. I. Filin, A. G. Sila-Novitskiy, et al. "ICONE11-36407 THERMOHYDRAULIC RESEARCH FOR THE CORE OF THE BREST-OD-300 REACTOR." Proceedings of the International Conference on Nuclear Engineering (ICONE) 2003 (2003): 172. http://dx.doi.org/10.1299/jsmeicone.2003.172.

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11

Grabezhnaya, Grabezhnaya V. A., Alexandr Sergeevich Mikheev, Stein Yu Yu Stein, and Semchenkov A. A. Semchenkov. "Numerical and Experimental Investigation of the Model Steam Generator Reactor Facility BREST-OD-300." Izvestiya Wysshikh Uchebnykh Zawedeniy, Yadernaya Energetika 2013, no. 1 (2013): 101–9. http://dx.doi.org/10.26583/npe.2013.1.13.

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12

Lopatkin, A. V., V. G. Muratov, V. V. Orlov, et al. "ICONE11-36406 EXPERIMENTAL AND CALCULATED VALIDATION OF THE BREST-OD-300 REACTOR NEUTRONIC CHARACTERISTICS." Proceedings of the International Conference on Nuclear Engineering (ICONE) 2003 (2003): 171. http://dx.doi.org/10.1299/jsmeicone.2003.171.

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13

Ivanov, V. K., E. V. Spirin, S. S. Lovachev, A. N. Menyajlo, S. Yu Chekin, and V. M. Solomatin. "Radiological protection of the public during the normal operating of the Pilot-demonstration energy complex (PDEC) and in increased total power of reactor plants in the Industrial power complex (IPC) within the framework of the Proryv Project." "Radiation and Risk" Bulletin of the National Radiation and Epidemiological Registry 30, no. 4 (2021): 5–23. http://dx.doi.org/10.21870/0131-3878-2021-30-4-5-23.

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The paper presents results of research on radiological protection of the public during normal op-eration of Pilot-demonstration energy complex (PDEC) and in increased total power of reactor plants in the Industrial power complex (IPC) based on the current national radiation safety stand-ards (NRB-99/2009), UNSCEAR conclusions and ICRP recommendations. To evaluate radiologi-cal protection of the public the concepts of radiological detriment (RD) and the level of radiation protection (LRP) were used. The concepts were also used to examine the compliance of the BREST-OD-300, fabrication/refabrica

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14

Grabezhnaya, V., and A. Mikheyev. "EXPERIMENTAL STUDY OF THERMAL HYDRAULICS ON THE MODEL OF HELICAL COILED STEAM GENERATOR HEATED BY LIQUID LEAD WITH LONGITUDINAL AND TRANSVERSE FLOW." PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS 2019, no. 1 (2019): 132–51. http://dx.doi.org/10.55176/2414-1038-2019-1-132-151.

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The study of heat transfer in spiral coiled tubes is of great interest in view of the widespread use of such channels in engineering practice, in particular, in nuclear power engineering in the form of steam generators at research reactors and nuclear power plants. In the projected BREST-OD-300 reactor facility (RF), a configuration of helical coiled tubes is considered as a steam generator. Thermal hydraulic tests of the model steam generator RF BREST-OD-300 (version 2000) with helical coiled tubes with longitudinal coolant flow were carried out in SSC RF - IPPE at the SPRUT facility in 2011-

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15

Shadrin, A. Yu, K. N. Dvoeglazov, A. G. Maslennikov та ін. "РH process as a technology for reprocessing mixed uranium–plutonium fuel from BREST-OD-300 reactor". Radiochemistry 58, № 3 (2016): 271–79. http://dx.doi.org/10.1134/s1066362216030085.

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16

Orlov, A. I., Yu I. Orlov, and V. A. Gulevskiy. "State of development of the heavy coolant quality support and control system for NF BREST-OD-300." Journal of Physics: Conference Series 1475 (March 2020): 012018. http://dx.doi.org/10.1088/1742-6596/1475/1/012018.

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17

Tolokonskiy, A. O., and D. G. Kovalionok. "Visualization of the Assembly and Control Area of the BREST-OD-300 Reactor FA Using Virtual Reality Technologies." Scientific Visualization 18, no. 2 (2025): 23–35. https://doi.org/10.26583/sv.18.2.02.

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Currently, virtual reality technologies are used to train specialists in the field of nuclear energy, which allow the student to directly immerse himself in the environment of his activities, to conduct training as close as possible to real conditions, without causing harm to his health. In order to manufacture a fuel assembly (FA), it is necessary to pass a number of control settings at the production stage, confirming the quality and safety. The authors proposed the development of a virtual simulator with FA control settings for the BREST-OD-300 reactor plant, such as: FA washing and drying,

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Tolokonskiy, A. O., and D. G. Kovalionok. "Visualization of the Assembly and Control Area of the BREST-OD-300 Reactor FA Using Virtual Reality Technologies." Scientific Visualization 17, no. 2 (2025): 23–35. https://doi.org/10.26583/sv.17.2.02.

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Currently, virtual reality technologies are used to train specialists in the field of nuclear energy, which allow the student to directly immerse himself in the environment of his activities, to conduct training as close as possible to real conditions, without causing harm to his health. In order to manufacture a fuel assembly (FA), it is necessary to pass a number of control settings at the production stage, confirming the quality and safety. The authors proposed the development of a virtual simulator with FA control settings for the BREST-OD-300 reactor plant, such as: FA washing and drying,

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19

Abramov, V. Ya, S. N. Bozin, S. V. Evropin, et al. "ICONE11-36413 CORROSION AND MECHANICAL PROPERTIES OF STRUCTURAL MATERIALS TO BE USED IN THE BREST-OD-300 REACTOR." Proceedings of the International Conference on Nuclear Engineering (ICONE) 2003 (2003): 178. http://dx.doi.org/10.1299/jsmeicone.2003.178.

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20

Spiridonov, S. I., A. N. Perevolotskii, T. V. Perevolotskaya, R. M. Aleksakhin, and E. V. Spirin. "Analysis of the Human Biohazard of Long-Lived Fission Products and Actinides for BREST-OD-300 Spent Fuel." Atomic Energy 123, no. 2 (2017): 122–26. http://dx.doi.org/10.1007/s10512-017-0312-2.

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21

Gorin, N. V., E. V. Kuznetsov, V. P. Kuchinov, et al. "BARRIERS TO NUCLEAR PROLIFERATION IN THE EXPORT OF THE RUSSIAN FAST REACTORS WITH CLOSED NFC (USING EXAMPLE BREST-OD-300)." NNC RK Bulletin, no. 4 (February 1, 2022): 16–21. http://dx.doi.org/10.52676/1729-7885-2021-4-16-21.

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In the coming decades, it is possible to start exporting fast reactors with a closed nuclear fuel cycle to non-nuclear-weapon countries, which will require strengthening the nuclear non-proliferation regime and increasing the effectiveness of IAEA safeguards. This can be achieved both by creating technical barriers and by improving the system of accounting and control of nuclear materials and ensuring their reliable physical protection. Using the example of the BREST-OD-300 reactor under construction as part of a pilot demonstration power complex, the analysis of design and technological featu

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22

Kulakov, E. N., A. V. Popov, and P. A. Kruglikov. "Optimization of parameters of the steam generator feedwaterinlet temperature maintenance system of the reactor BREST-OD-300 turbine unit." Nuclear Propulsion Reactor Plants. Life Cycle Management Technologies., no. 3 (2021): 23–35. http://dx.doi.org/10.52069/2414-5726_2021_3_25_23.

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23

Ivanov, V. K., S. Yu Chekin, A. N. Menyajlo, et al. "Potential radiological risk for the population during implementation of the Rosatom Proryv project. Part 2. Radiation detriment assessment." "Radiation and Risk" Bulletin of the National Radiation and Epidemiological Registry 29, no. 4 (2020): 48–68. http://dx.doi.org/10.21870/0131-3878-2020-29-4-48-68.

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Nuclear power is effective and safe source of electricity. Meanwhile, uranium reserves in the earth's crust will run out in 100 years with the development of traditional nuclear reactors. The Rosatom “Proryv” project implementation will allow multiplying fuel sources for the new genera-tion nuclear power through the closing fuel cycle. Radiation safety of the new nuclear powers should be based on the state of the art Russian national and international regulations, as well as on predicted radiation doses, estimates of potential radiation risks and radiation detriment of the public. Developed me

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24

Rogalev, Nikolay, Andrey Rogalev, Vladimir Kindra, Ivan Komarov, and Olga Zlyvko. "Structural and Parametric Optimization of S–CO2 Nuclear Power Plants." Entropy 23, no. 8 (2021): 1079. http://dx.doi.org/10.3390/e23081079.

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The transition to the use of supercritical carbon dioxide as a working fluid for power generation units will significantly reduce the equipment′s overall dimensions while increasing fuel efficiency and environmental safety. Structural and parametric optimization of S–CO2 nuclear power plants was carried out to ensure the maximum efficiency of electricity production. Based on the results of mathematical modeling, it was found that the transition to a carbon dioxide working fluid for the nuclear power plant with the BREST–OD–300 reactor leads to an increase of efficiency from 39.8 to 43.1%. Nucl

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25

Frolova, Anna V., Ksenia Y. Belova, and Sergey E. Vinokurov. "Medium-Temperature Glass-Composite Phosphate Materials for the Immobilization of Chloride Radioactive Waste." Journal of Composites Science 7, no. 9 (2023): 363. http://dx.doi.org/10.3390/jcs7090363.

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Among the many radiochemical problems, the search for new materials and technologies for the immobilization of radioactive waste remains relevant, and the range continues to change and expand. The possibility of immobilizing the spent chloride electrolyte after the pyrochemical processing of the mixed uranium-plutonium spent nuclear fuel of the new fast reactor BREST-OD-300 on lead coolant into glass-composite phosphate materials synthesized at temperatures of 650–750 °C was studied. The structure of the obtained samples was studied using XRD and SEM/EDS methods. It has been shown that the spe

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26

Isayev, Rafael Sh, Pavel S. Dzhumaev, Irina A. Naumenko, and Maria V. Leontieva-Smirnova. "Corrosion resistance of chromium coating on the inner surface of EP823-Sh steel cladding." Nuclear Energy and Technology 10, no. 2 (2024): 81–88. http://dx.doi.org/10.3897/nucet.10.119642.

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The processes of corrosion damage of the inner surface of the cladding are determined by corrosive reagents aggressive with respect to the cladding and the type of fuel used. Reactor irradiation of cladding made of EP823-Sh steel with mixed nitride fuel planned for use in the BREST-OD-300 reactor revealed non-uniform corrosion of the inner surface of the cladding. In this paper, the use of the chromium coating is proposed to prevent the corrosion of the inner surface of the steel fuel cladding. The results of corrosion tests of chromium coating applied to the inner surface of cladding made of

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Isayev, Rafael Sh., Pavel S. Dzhumaev, Irina A. Naumenko, and Maria V. Leontieva-Smirnova. "Corrosion resistance of chromium coating on the inner surface of EP823-Sh steel cladding." Nuclear Energy and Technology 10, no. (2) (2024): 81–88. https://doi.org/10.3897/nucet.10.119642.

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The processes of corrosion damage of the inner surface of the cladding are determined by corrosive reagents aggressive with respect to the cladding and the type of fuel used. Reactor irradiation of cladding made of EP823-Sh steel with mixed nitride fuel planned for use in the BREST-OD-300 reactor revealed non-uniform corrosion of the inner surface of the cladding. In this paper, the use of the chromium coating is proposed to prevent the corrosion of the inner surface of the steel fuel cladding. The results of corrosion tests of chromium coating applied to the inner surface of cladding made of

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28

Kordyukov, A. G., V. N. Leonov, A. A. Pikalov, et al. "Tests of Models of BREST-OD-300 Reactor Fuel Elements in an Autonomous Lead-Cooled Channel of a BOR-60 Reactor." Atomic Energy 97, no. 2 (2004): 564–70. http://dx.doi.org/10.1023/b:aten.0000047683.77555.98.

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29

Spiridonov, S. I., A. N. Perevolotskii, R. M. Aleksakhin, E. V. Spirin, and G. N. Vlaskin. "Radioecological Validation of the Extraction Parameters of Fission Products and Actinides from Spent Nuclear Fuel from the BREST-OD-300 Reactor." Atomic Energy 121, no. 3 (2017): 214–19. http://dx.doi.org/10.1007/s10512-017-0186-3.

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30

Popov, Aleksey V., Egor N. Kulakov, Daria R. Danilova, et al. "A New Circuit Solution to Minimize the Consequences of the Failure of the Feedwater Cooldown Line of the BREST-OD-300." Izvestiya Wysshikh Uchebnykh Zawedeniy, Yadernaya Energetika 2025, no. 2 (2025): 153–66. https://doi.org/10.26583/npe.2025.2.13.

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31

Aleksandrov, N. V., Ye D. Blank, A. D. Kashtanov, et al. "On the experimental lead-cooled installation." Voprosy Materialovedeniya, no. 4(100) (March 20, 2020): 185–92. http://dx.doi.org/10.22349/1994-6716-2019-100-4-185-192.

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Extensive experience in operating nuclear power plants convincingly proves that fast liquid metal cooled reactors are among the most promising. The advantages of using liquid lead coolants in nuclear power industry are shown. In Russia, lately, much attention has been paid to the natural safety of fast reactors. At the stage of testing materials for components of reactor plants, a number of problems arose for basic systems. An experimental lead-cooled installation was developed for testing large structures, continuous monitoring and maintaining specified technical parameters. For reliable cool

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32

Krechetnikov, V. V., I. E. Titov, S. I. Spiridonov, E. I. Karpenko, V. M. Solomatin, and A. I. Feygin. "The use of the built digital twin for the assessment and prediction of radiation exposure to humans in the area of the PDEC location." "Radiation and Risk" Bulletin of the National Radiation and Epidemiological Registry 33, no. 4 (2024): 29–39. https://doi.org/10.21870/0131-3878-2024-33-4-29-39.

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The results of the use of the built digital twin for the assessment and prediction of radiation exposure to humans in the area of the Pilot Demonstrative Energy Complex (PDEC) location are presented. Information on radioactive releases from the BREST-OD-300 reactor during the reactor normal operation was collected, the impact of releases from developed different types of emergency scenarios was estimated as well. Based on these data, the radiation doses to the humans exposed to releases from the normally operated reactor were measured. The doses from designed and beyond-designed accidents were

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Orlov, Alexander I., and Boris A. Gabaraev. "Heavy liquid metal cooled fast reactors: peculiarities and development status of the major projects." Nuclear Energy and Technology 9, no. 1 (2023): 1–18. http://dx.doi.org/10.3897/nucet.9.90993.

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Fast reactors with heavy liquid metal coolant (lead or eutectic bismuth-lead alloy) are one of the most advanced technologies capable to address the accumulated world nuclear energy issues. This innovative power technology is being developed in Russia, the USA, China and the European Union. Russia is the leader since it has focused on this topic for a number of decades. First concrete started to be poured in June 2021 to form the foundation of the Russian BREST-OD-300 lead cooled reactor scheduled to be started up in 2026. Attention is also given to the development status of the Chinese CLEAR

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Orlov, Alexander I., and Boris A. Gabaraev. "Heavy liquid metal cooled fast reactors: peculiarities and development status of the major projects." Nuclear Energy and Technology 9, no. (1) (2023): 1–18. https://doi.org/10.3897/nucet.9.90993.

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Fast reactors with heavy liquid metal coolant (lead or eutectic bismuth-lead alloy) are one of the most advanced technologies capable to address the accumulated world nuclear energy issues. This innovative power technology is being developed in Russia, the USA, China and the European Union. Russia is the leader since it has focused on this topic for a number of decades. First concrete started to be poured in June 2021 to form the foundation of the Russian BREST-OD-300 lead cooled reactor scheduled to be started up in 2026. Attention is also given to the development status of the Chinese CLEAR

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Abramov, A. V., E. O. Kovalev, P. A. Kolesnikov, et al. "Investigation of Processes in Lead Coolant with Loss-of-Integrity of a Heat-Exchange Tube in a Brest-OD-300 Steam Generator." Atomic Energy 119, no. 3 (2015): 200–206. http://dx.doi.org/10.1007/s10512-015-0041-3.

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36

Balovnev, A., V. Davidov, A. Zhirnov, et al. "SYSTEM OF CODES FOR PHYSICAL DESIGN OF THE LEAD-COOLED FAST REACTOR." PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS 2020, no. 3 (2020): 30–38. http://dx.doi.org/10.55176/2414-1038-2020-3-30-38.

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One of the actual task at present is the substantiation of the project of the pilot demonstration reactor with a lead coolant BREST-OD-300. For the implementation of large-scale development of nuclear power, which meets modern requirements for new generation reactors, a competitive commercial power unit BR-1200 with an electric capacity of 1200 MW is being designed. To solve complex problems in the study for the optimal configurations of the core, it is required to develop a system of design codes that allows us to perform works on physical design and safety justification. System of codes incl

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Okunev, Viacheslav. "The concept of a fast reactor with liquid metal fuel in tungsten capsules." E3S Web of Conferences 411 (2023): 01013. http://dx.doi.org/10.1051/e3sconf/202341101013.

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The concept of a dual-purpose high-power nuclear reactor is proposed. One of the goals is the production of electricity, the other is the production of high-potential thermal energy. It is proposed to use liquid fuel based on waste uranium and plutonium extracted from the spent fuel of VVER reactors (purified from the 238Pu isotope). The fuel is in sealed tungsten capsules. Lead extracted from thorium ores is used to cool the reactor. The electrical power of the reactor is 3.3 GW. The layout of the reactor is identical to the BREST-OD-300 reactor under construction. The analysis of emergency m

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Solomatin, V. M., and E. V. Spirin. "Impact of radioactive discharges of Siberian Chemical Plant (SCP) and Pilot-demonstration energy complex (PDEC) on aquatic biota inhabiting in the SCP 30-km zone." "Radiation and Risk" Bulletin of the National Radiation and Epidemiological Registry 31, no. 3 (2022): 26–36. http://dx.doi.org/10.21870/0131-3878-2022-31-3-26-36.

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Radiation doses to aquatic biota exposed to Siberian Chemical Plant (SCP, Tomsk) radioactive discharges, and Pilot-demonstration energy complex (PDEC) forecasting radiation doses from designed radioactive discharges were estimated. Doses to the biota in the habitat in the existing radiation situation were assessed with the use of measurements of water and bottom sediments samples collected during environmental monitoring of airborne radioactivity in 2017. Designed radioactivity discharges were assumed from normally operated PDEC modules for fabrication and refabrication and fuel reprocessing a

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Melikhov, V. I., O. I. Melikhov, and B. Saleh. "Model of a stationary thermal detonation wave in the “liquid lead – water” system for safety analysis of NPP with the reactor BREST-OD-300 during heat exchanger tube break accident." IOP Conference Series: Earth and Environmental Science 1154, no. 1 (2023): 012006. http://dx.doi.org/10.1088/1755-1315/1154/1/012006.

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Abstract This paper presents a mathematical model of a stationary wave of thermal detonation in the “liquid lead - water” system, which can occur after the rupture of a steam generator tube of the reactor BREST-OD-300. The model is based on the mechanics of multiphase fluid flows. The system under study includes a continuous phase of liquid lead, in which drops of water surrounded by a steam film are dispersed. Heat exchange between high-temperature molten lead and water drops is carried out in the film boiling mode. A shock wave propagating in this multiphase system causes all phases in motio

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Okunev, Viacheslav Sergeevich. "Compacting the core of a commercial high-power lead-cooled fast reactor." E3S Web of Conferences 494 (2024): 03006. http://dx.doi.org/10.1051/e3sconf/202449403006.

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Two options for the layout of a fast reactor with lead coolant are being considered. The design of the fuel assemblies and core fuel rods is identical to the BREST-OD-300 reactor. One of the layouts is traditional with a flattened core. The other layout is the most compact. It has an elongated core and is characterized by an optimal ratio of core diameter to height in terms of minimal neutron leakage. The electric capacity of the first of them is about 2.9 GW, the second - 13.3 GW. The second layout is a high-temperature reactor. When operating at rated power, the maximum coolant temperature r

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Chukhlov, Aleksey G., and Ekaterina О. Zherebtsova. "Numerical Study of the Temperature Field in the BREST-OD-300 Reactor Plant’s Core with the Partially Blocked Flow Cross Section at the Coolant Inlet." Vestnik MEI, no. 1 (February 2018): 67–71. http://dx.doi.org/10.24160/1993-6982-2018-1-67-71.

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Askhadullin, R., A. Legkikh, V. Ulyanov, and I. Voronin. "CURRENT STATE AND ISSUES OF THE HEAVY LIQUID METAL COOLANT TECHNOLOGY DEVELOPMENT (Pb, Pb-Bi)." PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS 2021, no. 2 (2021): 105–15. http://dx.doi.org/10.55176/2414-1038-2021-2-105-115.

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The technology of heavy liquid metal coolant (HLMC) is an important part of the safety system for the operation of reactor facilities with HLMC at all stages of their life cycle: from the preparation of a coolant and loading into the reactor to the decommissioning of the reactor facility (RF). The technology of heavy liquid metal coolant is a set of measures that allow: - to prepare the coolant for filling in the primary circuit of the RF; - to maintain the conditions in the coolant to ensure the corrosion resistance of structural steels; - to perform the coolant cleaning from solid-phase slag

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Chernov, E. I., M. E. Chernov, V. I. Rachkov, A. I. Orlov, and Yu M. Sysoev. "The Current State of Development and Prospects for the Creation of Sensors of Thermodynamic Activity of Oxygen in Relation to Reactor Units with a Heavy Liquid Metal Coolant." Известия Российской академии наук. Энергетика, no. 1 (January 1, 2023): 18–34. http://dx.doi.org/10.31857/s0002331023010041.

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Successful development of the world nuclear energy is not possible without a transition to a new technological platform. The most promising nuclear technology for this today seems to be fast reactors cooled by liquid metal coolants. In 2021, the construction of the world’s first nuclear power plant with a fast reactor BREST-OD-300 began in the city of Seversk, Russia. This reactor, having a number of innovative solutions, uses liquid lead as the primary coolant. Lead belongs to the class of heavy liquid metal coolants and is similar in technological properties to the Pb-Bi eutectic alloy. Sinc

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Leonov, V. N., A. A. Pikalov, A. G. Sila-Novitsky, et al. "ICONE11-36409 PRE-AND IN-PILE TESTS OF FUEL ELEMENT MOCKUPS FOR THE BREST-OD-300 REACTOR IN INDEPENDENT LEAD-COOLED CHANNEL OF THE BOR-60 REACTOR." Proceedings of the International Conference on Nuclear Engineering (ICONE) 2003 (2003): 174. http://dx.doi.org/10.1299/jsmeicone.2003.174.

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Chudinova, V., and S. Nikonov. "THE THERMOHYDRAULIC MODEL OF A REACTOR WITH A LIQUID METAL COOLANT." PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS 2019, no. 3 (2019): 120–27. http://dx.doi.org/10.55176/2414-1038-2019-3-120-127.

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This paper presents the results of calculations on the heat-hydraulic model of a reactor with lead coolant, which is based on the design scheme for the ATHLET code, obtained on the basis of public information on the BREST-OD-300. In the model used in this work, firstly, the descending sections of each loop are interconnected by transverse hydraulic connections and, secondly, the in-reactor space from the lower pressure collector of the reactor to the upper distributing collector of the reactor is also divided into a system of parallel cross-linked channels. The model includes a reactor, four c

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Beznosov, A., T. Bokova, P. Bokov, A. Marov, A. L'vov, and N. Volkov. "SUBSTANTIATION OF TECHNICAL SOLUTIONS OF THE REACTOR CIRCUIT OF THE BRS-GPG UNITS OF SMALL AND MEDIUM POWER WITH A HEAVY LIQUID METAL COOLANT." PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS 2020, no. 1 (2020): 132–39. http://dx.doi.org/10.55176/2414-1038-2020-1-132-139.

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The analysis and new scientific and technical solutions corresponding to the evolutionary development of low and medium power reactor plants with heavy liquid metal coolants are presented. The analysis was carried out on the basis of experience in the creation and operation of reactor installation with lead-bismuth coolant and research, primarily experimental, performed at the Nizhny Novgorod State Technical University (NSTU) in rationale of the small and medium-sized power plants developed at NNSTU reactor installation with horizontal steam generators (BRS-GPG). The paper substantiates the ch

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Ivanov, V. K., A. N. Menyajlo, A. M. Korelo, et al. "Comparative analysis of radiological risks for personnel during the normal operation of the BREST-OD-300 reactor and risks from other adverse environmental factors of a non-radiation nature." "Radiation and Risk" Bulletin of the National Radiation and Epidemiological Registry 34, no. 1 (2025): 5–13. https://doi.org/10.21870/0131-3878-2025-34-1-5-13.

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Currently, within the framework of the “Proryv” Project, a pilot demonstration energy complex based on the BREST-OD-300 reactor unit is being developed at the industrial site of JSC “Siberian Chemical Combine” (JSC “SChC”). This complex will allow testing nuclear energy technologies of the new generation with a closed nuclear fuel cycle. The key issue in this project is ensuring the safety of personnel. This includes protection from both radiation and non-radiation carcinogenic risk factors. This article presents a comparative analysis of carcinogenic risks for JSC “SChC” employees from exposu

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Nesterov, Yu V., A. S. Lisyanskii, E. I. Makarova, L. Ya Bal’va, and P. Yu Prikhod’ko. "The thermal process diagram and equipment of the secondary coolant circuit of a nuclear power station unit based on the BREST-OD-300 reactor installation for subcritical steam conditions." Thermal Engineering 58, no. 6 (2011): 478–82. http://dx.doi.org/10.1134/s0040601511060103.

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Balovnev, Aleksey V., Vladimir K. Davydov, Andrey P. Zhirnov, Andrey V. Moiseev, and Evgenii O. Soldatov. "Simulating the fuel cycle of a lead-cooled fast reactor." Nuclear Energy and Technology 8, no. 1 (2022): 71–76. http://dx.doi.org/10.3897/nucet.8.83062.

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The development of nuclear power with fast reactors is associated with the implementation of a closed nuclear fuel cycle (CNFC). In this regard, one actual task is to simulate the stages of the fuel cycle with study of the neutron-physical characteristics of the core. The design of a reactor for operation in the closed nuclear fuel cycle mode is impossible without the using of verified and certified software packages for calculating fast reactors, capable of simulating all stages of the operation of the reactor facility and the fuel cycle. For the calculations, the FACT-BR software package was

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Ivanov, V. K., A. N. Menyajlo, S. Yu Chekin, E. V. Spirin, and V. M. Solomatin. "Pilot-demonstration energy complex (PDEC): the level of radiological protection of the population due to the modern “dose-effect” model of the ICRP." "Radiation and Risk" Bulletin of the National Radiation and Epidemiological Registry 32, no. 1 (2023): 5–20. http://dx.doi.org/10.21870/0131-3878-2023-32-1-5-20.

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After receiving the notification of the license granting and its registration in the license register it has become topical the preparation of the final version of the project documentation to guarantee the safety of the public residing in the proximity of the operating Pilot-demonstration energy complex (PDEC), which includes a reactor unit BREST-OD-300, reprocessing module and a fuel fabrication/refabrication module. Currently International Commission on Radiological Protection (ICRP, Publication 103) recommends for estimating radiological protection of the public to use carcinogenic risks e

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