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Kruglikov, Yu.S.; Poplavskij, V.M.
Results of the scientific and technical activities of the Nuclear Reactors and Thermal Physics Institute for 2014. Scientific and technical collection2015
Results of the scientific and technical activities of the Nuclear Reactors and Thermal Physics Institute for 2014. Scientific and technical collection2015
AbstractAbstract
[en] Basic proposals on completing calculational procedures of sodium cooled reactors hydrogen safety justification, particularly steam generator (SG) box, are presented. It is pointed out that hydrogen appearance in SG box is possible only in accidents with multiple failures. The data presented shows that hydrogen safety and explosion-proofness of fast reactor facility SG box are provided without any special complex and expensive means
[ru]
Приведены основные предложения по дополнению расчетных методик обоснования водородной безопасности охлаждаемых натрием реакторов, в частности бокса парогенератора (ПГ). Отмечается, что появление водорода в боксе ПГ возможно лишь при авариях с многочисленными отказами. Представленные данные показывают, что водородная безопасность и взрывозащищенность бокса ПГ реакторной установки с реактором на быстрых нейтронах обеспечивается без применения каких-либо специальных сложных и дорогостоящих средствOriginal Title
Metodika obosnovaniya vodorodnoj bezopasnosti pomeshchenij RU BN na primere boksa parogeneratora
Primary Subject
Source
Trufanov, A.A.; Sorokin, A.P.; Vereshchagina, T.N. (eds.); Predpriyatie Goskorporatsii Rosatom AO Gosudarstvennyj Nauchnyj Tsentr Rossijskoj Federatsii - Fiziko-Ehnergeticheskij Inst. imeni A.I. Lejpunskogo, Obninsk (Russian Federation); 318 p; ISBN 978-5-906512-69-7;
; 2015; p. 130-140; 22 refs., 5 figs., 1 tab.

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Book
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Poplavskij, V.M.
Fast nuclear reactors. Russian scientific and engineering forum. Theses of reports2003
Fast nuclear reactors. Russian scientific and engineering forum. Theses of reports2003
AbstractAbstract
No abstract available
Original Title
Tekhnologiya bystrykh reaktorov. Sostoyanie i perspektivy
Primary Subject
Source
Gosudarstvennyj Nauchnyj Tsentr Rossijskoj Federatsii Fiziko-Ehnergeticheskij Inst. im. A.I. Lejpunskogo, Obninsk (Russian Federation); 80 p; 2003; p. 8-9; Nuclear power engineering with fast reactors; Yadernye ehnergeticheskie tekhnologii s reaktorami na bystrykh nejtronakh; Obninsk (Russian Federation); 9-10 Dec 2003; Available from the Federal State Unitary Enterprise TSNIIATOMINFORM, Russian Federation, 127434, Moscow Dmitrovskoe sh., 2
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Miscellaneous
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Conference
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AbstractAbstract
[en] The results of the analysis of the sate and prospects of developing NPPs with the fast neutrons reactors are presented. The basic trends of their development are considered. Comparison of the technical-economical indices of the power units with different coolants is given. Special attention is paid to developing the new generation of gas-cooled reactors, capable of increasing the safety levels and competitive efficiency as well as to solving the problem on the long-lived wastes storage
[ru]
Представлены результаты анализа состояния и перспектив развития АЭС с реакторами на быстрых нейтронах. Рассмотрены основные тенденции их развития. Приведено сравнение технико-экономических показателей энергоблоков с различными теплоносителями. Особое внимание уделяется разработке нового поколения быстрых газоохлаждаемых реакторов, способных повысить уровни безопасности и экономической конкурентноспособности, а также решить задачу утилизации долгоживущих отходовOriginal Title
Sostoyanie i perspektivy razvitiya AEhS s reaktorami na bystrykh nejtronakh
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24 refs., 3 tabs.
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Journal Article
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AbstractAbstract
[en] The experience gained in the BN-350 plant is of fundamental importance with a view to sodium/water operation and separation in the steam generator by means of a simple wall. It was found that damage due to water ingress in the sodium loop can be prevented. Partitioned steam generators with a suitable protective system have certain advantages which enable localisation of the defect and also, if necessary, continued operation of the intact channels, independent of the extent of the original leak. (orig.)
[de]
Aus der Sicht der moeglichen Kombination von Natrium und Dampf in einem Kernkraftwerk und ihrer Trennung durch eine einfache Wand in praktischen Dampferzeugern haben die Erfahrungen mit der BN-350-Anlage grundsaetzliche Bedeutung. Ferner wurde festgestellt, dass auch bei einem Gelangen einer groesseren Menge Wasser in das Natrium die Fehlervorgaenge im Na-Kreislauf ohne dessen Zerstoerung begrenzt werden koennen. In diesem Zusammenhang weisen sektionierte Dampferzeuger mit einem geeigneten Schutzsystem gewisse Vorzuege auf, die eine Lokalisierung der Fehlerwirkung im Bereich der schadhaften Sektion erlauben und (notfalls) den Weiterbetrieb des Dampferzeugers mit den intakten Kanaelen, unabhaengig vom Ausmass der urspruenglichen Leckstelle erlauben. (orig.)Original Title
Fragen der Sicherheit von Natrium/Wasserdampferzeugern und ihre Loesung in der Sowjetunion
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Source
Translated from Russian: Voprosy bezopasnosti parogeneratorov natrij-voda i ich resenie v SSR; published in Atomnaja energija, v. 46(5), p. 311-315, May 1979.
Record Type
Journal Article
Journal
Archiv fuer Energiewirtschaft; ISSN 0003-9047;
; v. 33(11); p. 914-922

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AbstractAbstract
[en] Design specific features of steam generators being a part of a NPP with the BN-350 type reactor are considered. The results of investigation of conditions for the steam generator reliable and safe operation are presented. The problems of steam generator maintenance and repair are discussed. Operational experience of the BN-350 type steam generator at different loadings and pressures has shown that they are convenient and simple in operation. Their characteristics correspond to calculational ones, automatic control systems assure the designed level of parameters. Accidents with water leakage in evaporators testify to the fact that in order to assure reliable operation of steam generator a thoroughful material and welded joints control is required in the process of manufacturing and assembly including tube production. The operational experience of the BN-350 type steam generators shows that even at considerable water leaks in sodium the emergency protection system can prevent the plant failure. It is concluded that the BN-350 type steam generators may be considered as one of possible variants when selecting construction of steam generators for prospective NPPs with sodium cooled fast breeder reactors
Original Title
Issledovanie i opyt ehkspluatatsii parogeneratorov natrij-voda AEhS s reaktorom BN-350
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Source
For English translation see the journal Thermal Engineering (UK).
Record Type
Journal Article
Journal
Teploehnergetika; ISSN 0040-3636;
; (no.1); p. 7-13

Country of publication
ALKALI METALS, BOILERS, BREEDER REACTORS, CHEMICAL REACTIONS, DESALINATION REACTORS, ELEMENTS, EPITHERMAL REACTORS, FAST REACTORS, FBR TYPE REACTORS, HYDROGEN COMPOUNDS, INFORMATION, LIQUID METAL COOLED REACTORS, METALS, NUCLEAR FACILITIES, OXYGEN COMPOUNDS, POWER PLANTS, POWER REACTORS, REACTORS, SODIUM COOLED REACTORS, THERMAL POWER PLANTS, VAPOR GENERATORS
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AbstractAbstract
[en] Our national experience demonstrates reliable operation and high level of nuclear and radiational safety of fast reactors
Original Title
Stimuly razvitiya bystrykh reaktorov s natrievym teplonositelem
Primary Subject
Source
4. Annual Scientific and Technical Conference of the Nuclear Society; YaEh-93. Yadernaya ehnergiya i bezopasnost' cheloveka; Nizhnij Novgorod (Russian Federation); 28 Jun - 2 Jul 1993
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Journal Article
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AbstractAbstract
[en] The main results of malfunction experimental studies connected with sodium-water interaction are presented. Considered are the main principles of calculational methods elaborated in order to analyze accidents in the steam generator and in the sodium circuit of an NPP with a fast reactor. Stated are the USSR modern approach to safety problems of the sodium-water steam generators with one-wall separation of coolants and demands to their emergency protective systems. It is shown that steam generators with sectioned design and appropriate procective system have some advantages; these designs allow one to localize consequences of an accident in a demaged area and to provide (if necessary) running of the steam generator using intact sections regardless of the initial volume of water flow into sodium
Original Title
Voprosy bezopasnosti parogeneratorov natrij-voda i ikh reshenie v SSSR
Primary Subject
Source
For English translation see the journal Soviet Journal of Atomic Energy (USA).
Record Type
Journal Article
Journal
Atomnaya Ehnergiya; ISSN 0004-7163;
; v. 46(5); p. 311-316

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AbstractAbstract
[en] Steps of the development of fast reactors in the world are analyzed. Achievements in the field of the elaboration and operation of sodium cooled fast reactors are performed. Means for the advancement of concurrent ability of this type of NPP units are discussed. State of works on types of gas, vapor, heavy metal cooled fast reactors briefly reviewed. Main problems in the realization of these directions of the development of fast reactors are posed
[ru]
Анализируются этапы развития быстрых реакторов в мире. Представлены достижения в области разработки и эксплуатации быстрых реакторов с натриевым теплоносителем. Обсуждаются пути достижения конкурентоспособности этого типа блоков АЭС. Приведен краткий обзор состояния работ по типам быстрых реакторов с другими теплоносителями: газ, пар, тяжелые металлы. Высказаны соображения о главных проблемах, которые требуют своего решения для реализации этих направлений развития быстрых реакторовOriginal Title
Bystrye reaktory. Sostoyanie i perspektivy
Primary Subject
Source
Forum devoted to the centenary of A.I. Lejpunskij; Forum, posvyashchennyj 100-letiyu so dnya rozhdeniya A.I. Lejpunskogo; Obninsk-Moscow (Russian Federation); 8-12 Dec 2003; 11 refs., 3 tabs.
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Journal Article
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Conference
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BREEDER REACTORS, COOLING SYSTEMS, DOCUMENT TYPES, ENERGY SYSTEMS, ENGINEERING, ENRICHED URANIUM REACTORS, EPITHERMAL REACTORS, EVALUATION, FAST REACTORS, FBR TYPE REACTORS, LIQUID METAL COOLED REACTORS, LMFBR TYPE REACTORS, POWER REACTORS, PWR TYPE REACTORS, REACTOR COMPONENTS, REACTORS, SODIUM COOLED REACTORS, THERMAL REACTORS, WATER COOLED REACTORS, WATER MODERATED REACTORS, WWER TYPE REACTORS
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Ashurko, Yu.M.; Poplavskij, V.M.
10th International scientific-technical conference «Safety, efficiency and economics of nuclear energy». Plenary and sectional papers2016
10th International scientific-technical conference «Safety, efficiency and economics of nuclear energy». Plenary and sectional papers2016
AbstractAbstract
[en] The consideration is given to the problems and goals of the «Generation IV» International Forum, its structure, road map, basic characteristics, which is required to provide in the design of G-IV reactors. The main directions of works on reactor technologies of fast sodium reactor, fast lead reactor, fast gas reactor, supercritical water cooled reactor, molten salt reactor, supercritical water reactor are discussed
[ru]
Рассмотрены задачи и цели Международного форума «Поколение-IV», его структура, дорожная карта, основные характеристики, которые требуется обеспечить в проектах реакторов 4-го поколения. Обсуждаются главные направления работ по реакторным технологиям быстрого натриевого реактора, быстрого свинцового реактора, быстрого газового реактора, сверхкритического водяного реактора, жидкосолевого реактора, сверхвысокотемпературного газового реактораOriginal Title
Razvitie perspektivnykh reaktornykh tekhnologij 4-go pokoleniya v ramkakh mezhdunarodnogo foruma «Pokolenie-IV»
Primary Subject
Source
Rosehnergoatom (Ehlektroehnergeticheskij Divizion Rosatoma), Moscow (Russian Federation); 800 p; ISBN 978-5-88777-038-3;
; 2016; p. 537-543; 10. International scientific-technical conference ''Safety, efficiency and economics of nuclear energy''; Desyataya Mezhdunarodnaya nauchno-tekhnicheskaya konferentsiya «Bezopasnost', ehffektivnost' i ehkonomika atomnoj ehnergetiki»; Moscow (Russian Federation); 25-27 May 2016; 8 figs., 4 tabs.

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Book
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Conference
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Bogoslovskaya, G.P.; Zhukov, A.V.; Poplavskij, V.M.; Sorokin, A.P.; Tikhomirov, B.B.; Ushakov, P.A.
Gosudarstvennyj Komitet po Ispol'zovaniyu Atomnoj Ehnergii SSSR, Obninsk. Fiziko-Ehnergeticheskij Inst1982
Gosudarstvennyj Komitet po Ispol'zovaniyu Atomnoj Ehnergii SSSR, Obninsk. Fiziko-Ehnergeticheskij Inst1982
AbstractAbstract
[en] A statistical model of a fuel element cluster considering the laws of channel flow section distribution obtained in experiments at the BN-600 reactor fuel assembly mock-up is suggested. Technique for calculating the fuel assembly temperature field based on the sequential calculation of a set of fuel assembly versions taking into account the interchannel heat and mass transfer is described. Variants of channel flow section distribution formed by fuel elements are realized using the Monte-Carlo method according to the given function of channel flow section probability density. As a result of the calculations statistical distributions of temperature increase in channels without account for the interchannel heat and mass transfer and with it are obtained. The effect of interchannel transfer on the reduction of maximum nonuniformity in coolant temperature increase caused by dispersion of channel flow sections is shown. The conclusion is made that if heat and mass transfer is not taken into account the value of coolant temperature increase in separate channels coincides with limited value obtained by the dispersion method
Original Title
Metod rascheta temperaturnogo polya v kassete tvehlov bystrogo reaktora pri sluchajnom raspredelenii parametrov po metodu Monte-Karlo
Primary Subject
Source
1982; 14 p; 5 refs.; 4 figs.; 2 tabs.
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Report
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