Relevant bibliographies by topics / BREST-OD-300

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Academic literature on the topic 'BREST-OD-300'

Author: Grafiati
Published: 3 June 2025
Last updated: 31 July 2025

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

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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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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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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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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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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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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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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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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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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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Conference papers on the topic "BREST-OD-300"

1

Fedorovich, D., and Irina Lvovna Paramonova. "CONVECTIVE HEAT TRANSFER IN THE BREST-OD-300 NUCLEAR REACTOR FUEL ROD." In Proceedings of CONV-22: Int. Symp. on Convective Heat and Mass Transfer June 5 – 10, 2022, Turkey. Begellhouse, 2022. http://dx.doi.org/10.1615/ichmt.2022.conv22.230.

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Fomichev, Dmitry V., and Vladimir I. Solonin. "Structure of the turbulent flow in the BREST-OD-300 wire-spaced fuel assemblies." In 2015 5th International Youth Conference on Energy (IYCE). IEEE, 2015. http://dx.doi.org/10.1109/iyce.2015.7180754.

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Fedorovich, Daria, Elena Ignatieva, and Irina Paramonova. "Perspective Fuel Rod with Alloyed Lead in Contact Layer for Reactor BREST-OD-300." In International Scientific and Practical Conference "Young Engineers of the Fuel and Energy Complex: Developing the Energy Agenda of the Future" (EAF 2021). Atlantis Press, 2022. http://dx.doi.org/10.2991/aer.k.220308.005.

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Gotovsky, M. A., V. F. Ermolov, V. E. Mikhailov, Yu G. Sukhorukov, and N. N. Trifonov. "On the Prospects of Using Deaeratorless Thermal Schemes With Direct Contact Low-Pressure Reheaters at New Nuclear Power Plants." In 2014 22nd International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2014. http://dx.doi.org/10.1115/icone22-30487.

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Now in Russia a new NPP “Brest – OD - 300” is developing which is to become a head in a series of fast reactors cooled by liquid lead or lead-bismuth alloy. Russian abbreviation OD is translated as Pilot-Demonstrational. The ideas are developed to use in this power plant deaeratorless thermal circuit. Such the schemes are widely and successfully used for conventional power plants in Russia. Deaeratorless thermal schemes are based on the use of direct contact low-pressure reheaters. These reheaters have deaeration ability. Such schemes improve power plants efficiency by 1.5 %. Some other deaera
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Journal articles
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