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Milian Lorenzo, D.; Quintero Rosello, R.; Soler Iglesias, B.; Alonso Garcia, D.; Diaz Duennas, J.A.
International Atomic Energy Agency, Viena (Austria); Organizacion Panamericana de la Salud (OPS), La Habana (Cuba); International Radiation Protection Association (IRPA) Washington, DC, (United States)
Proceedings on Radiological and Nuclear Safety. 4ph Regional Congress. Regional Congress IRPA. Vol 41998
International Atomic Energy Agency, Viena (Austria); Organizacion Panamericana de la Salud (OPS), La Habana (Cuba); International Radiation Protection Association (IRPA) Washington, DC, (United States)
Proceedings on Radiological and Nuclear Safety. 4ph Regional Congress. Regional Congress IRPA. Vol 41998
AbstractAbstract
[en] A calculate methodology is presented to execute the analysis trips gives power place by reactivity inserts caused by explosion or uncontrollable extraction the nucleonic groups control the VVER-440 reactor
Original Title
Metodologia de calculo para el analisis de lo accidentes de reactividad de los reactores VVER-440
Primary Subject
Source
International Atomic Energy Agency, Viena (Austria); Organizacion Panamericana de la Salud (OPS), La Habana (Cuba); International Radiation Protection Association (IRPA) Washington, DC, (United States); 355 p; 1998; p. 13.17-13.19; 4. Regional Congress on Radiological and Nuclear Safety. Regional Congress IRPA; 4. Congreso Regional sobre Seguridad Radiologica y Nuclear. Congreso Regional IRPA; La Habana (Cuba); 19-23 Oct 1998; Available from CIEN, La Habana, Cuba
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Lopez Aldama, D.; Rodriguez Gual, R.
International Atomic Energy Agency, Viena (Austria); Organizacion Panamericana de la Salud (OPS), La Habana (Cuba); International Radiation Protection Association (IRPA) Washington, DC, (United States)
Proceedings on Radiological and Nuclear Safety. 4ph Regional Congress. Regional Congress IRPA. Vol 41998
International Atomic Energy Agency, Viena (Austria); Organizacion Panamericana de la Salud (OPS), La Habana (Cuba); International Radiation Protection Association (IRPA) Washington, DC, (United States)
Proceedings on Radiological and Nuclear Safety. 4ph Regional Congress. Regional Congress IRPA. Vol 41998
AbstractAbstract
[en] Presently work intends to validate the models and programs used in the Nuclear Technology Center for calculating the critical position of control rods by means of the analysis of the measurements performed at the critical facility IPEN/MB-01. The lattice calculations were carried out with the WIMS/D4 code and for the global calculations the diffusion code SNAP-3D was used
Original Title
Validacion de la metodologia de calculo de Hcrit de las barras de control del conjunto critico del IPEN/MB-01
Primary Subject
Source
International Atomic Energy Agency, Viena (Austria); Organizacion Panamericana de la Salud (OPS), La Habana (Cuba); International Radiation Protection Association (IRPA) Washington, DC, (United States); 355 p; 1998; p. 13.36-13.39; 4. Regional Congress on Radiological and Nuclear Safety. Regional Congress IRPA; 4. Congreso Regional sobre Seguridad Radiologica y Nuclear. Congreso Regional IRPA; La Habana (Cuba); 19-23 Oct 1998; Available from CIEN, La Habana, Cuba
Record Type
Miscellaneous
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Prieto Miranda, E.; Melo Crespo, J.C.
International Atomic Energy Agency, Viena (Austria); Organizacion Panamericana de la Salud (OPS), La Habana (Cuba); International Radiation Protection Association (IRPA) Washington, DC, (United States)
Proceedings on Radiological and Nuclear Safety. 4ph Regional Congress. Regional Congress IRPA. Vol 41998
International Atomic Energy Agency, Viena (Austria); Organizacion Panamericana de la Salud (OPS), La Habana (Cuba); International Radiation Protection Association (IRPA) Washington, DC, (United States)
Proceedings on Radiological and Nuclear Safety. 4ph Regional Congress. Regional Congress IRPA. Vol 41998
AbstractAbstract
[en] The present work has as objective to make a working paper where the approaches settle down Security in the design and the requirements for the sure explotation the nuclear facilities
Original Title
Criterios de seguridad y requerimientos para la explotacion de instalaciones de irradiacion
Primary Subject
Source
International Atomic Energy Agency, Viena (Austria); Organizacion Panamericana de la Salud (OPS), La Habana (Cuba); International Radiation Protection Association (IRPA) Washington, DC, (United States); 355 p; 1998; p. 13.52-13.54; 4. Regional Congress on Radiological and Nuclear Safety. Regional Congress IRPA; 4. Congreso Regional sobre Seguridad Radiologica y Nuclear. Congreso Regional IRPA; La Habana (Cuba); 19-23 Oct 1998; Available from CIEN, La Habana, Cuba
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Miscellaneous
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Pisica, A.; Horhoianu, G.
Nuclear Power - Current Status and Perspectives. INR 1971-2001 Symposium. Volume II2001
Nuclear Power - Current Status and Perspectives. INR 1971-2001 Symposium. Volume II2001
AbstractAbstract
[en] This work presents results of the research done in INR Pitesti on the new concept of SEU 43 fuel cluster designed for burnups as high as 25 Mw·day/kgU using slightly enriched uranium (up to 1.1% U235). The study of behaviour of fuel elements in accident conditions aims to find their capacity to preserve their physical integrity as well as how their failure can affect both the nuclear safety of nuclear power plant and the environment as well. By using ELOKA-MK6 and CAREB codes, the behaviour of two SEU43 fuel elements in LOCA conditions (RIH 20%, 35%, ROH 80% and 100%) was analyzed. An analysis of the main performance parameters of SEU43 fuel as compared with the results obtained for standard CANDU fuel is included. The characteristic LOCA thermohydraulic conditions were provided by FIREBIRD code and used as input data for CAREB and ELOCA. ELESIM code was used to establish the state of fuel prior to the accident. The temperature, internal pressure and total deformation of the can are plotted. The peak values of SEU43 parameters are lower than the values obtained for standard CANDU fuel. The can rupture due to excessive increase of internal pressure caused by fission gas release is one of the mechanisms characterizing the fuel integrity in LOCA type accidents. The peak total deformation of the can obtained with CAREB is -0.02% in case of 80%ROH and 0.4% in case of 20%RIH, values significantly lower then the rupture threshold. The peak pressure of fission gases in the can's interior is 6.93 MPa after 3.6 sec in case of 80%ROH, while in the 20%RIH it reaches 7.97 MPa after 3.8 sec from accident initiation. The peak temperature of the can in 20%RIH case was 1176.6 K and 1004.7 K for 80%ROH. The results obtained show the capability of SEU 43 cluster of functioning at power levels per cluster higher as compared with the standard fuel with consequences of an LOCA type accident not increased
Original Title
Comportarea elementului combustibil SEU 43 la puteri ridicate in conditii de echilibru
Primary Subject
Source
Cojan, Mihail (Institute for Nuclear Research - Pitesti, PO Box 78, RO-0300 Pitesti (Romania)); Institute for Nuclear Research - Pitesti, PO Box 78, RO-0300 Pitesti (Romania). Funding organisation: Ministerul Industriei si Resurselor, Ministerul Educatiei si Cercetarii, Bucharest (Romania); Societatea Nationala NUCLEARELECTRICA, Bucharest (Romania); Regia Autonoma pentru Activitati Nucleare, Bucharest (Romania); Institutul de Cercetari Nucleare, Pitesti (Romania); Fabrica de Combustibil Nuclear, Pitesti (Romania); 304 p; 2001; p. 111-112; Nuclear Power - Current Status and Perspectives. INR 1971-2001 Symposium; Conferinta ENERGETICA NUCLEARA - PREZENT SI VIITOR. ICN 1971-2001. Volum II; Pitesti (Romania); 13 Jul 2001; Available from author(s) or Institute for Nuclear Research - Pitesti, PO Box 78, RO-0300 Pitesti (RO). Fax 40-48-262449; Available from Institute for Nuclear Research - Pitesti, PO Box 78, RO-0300 Pitesti (RO). Fax 40-48-262449; 2 refs., 2 figs.
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Miscellaneous
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Kobylyanskij, G.P.; Makhin, V.M.; Shulimov, V.N.; Maershina, G.I.; Smirnov, A.V.; Lyadov, G.D.
Collection of reports of the 4. Inter-industry conference on reactor materials. Volume 1. Fuel and fuel elements for power reactors1996
Collection of reports of the 4. Inter-industry conference on reactor materials. Volume 1. Fuel and fuel elements for power reactors1996
AbstractAbstract
[en] The results of reactor tests of 18-element fuel assembly with the WWER-type fuel elements in the regime characteristic of the small leakage accident are presented. The state of fuel elements after overheating during reactor tests is determined and evaluation of their efficiency is carried out
[ru]
Приведены результаты реакторных испытаний 18-элементной ТВС с твэлами типа ВВЭР-1000 в режиме, характерном для аварии малая течь. Определено состояние твэлов после перегревов при реакторных испытаниях, проведена оценка их работоспособности и намечены пути совершенствования реакторного экспериментаOriginal Title
Sostoyanie ehksperimental'nykh TVS tipa VVEhR posle reaktornykh ispytanij v rezhimakh, modeliruyushchikh avarii s poterej teplonositelya
Primary Subject
Source
Ministerstvo Rossijskoj Federatsii po Atomnoj Ehnergii, Moscow (Russian Federation); Gosudarstvennyj Nauchnyj Tsentr Rossijskoj Federatsii Nauchno-Issledovatel'skij Inst. Atomnykh Reaktorov, Dimitrovgrad (Russian Federation); 431 p; ISBN 5-85165-169-5;
; 1996; p. 148-164; 4. Inter-industry conference on reactor materials; 4. Mezhotraslevaya konferentsiya po reaktornomu materialovedeniyu; Dimitrovgrad (Russian Federation); 15-19 May 1995; 2 refs., 9 figs.

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Book
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Conference
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Men'kin, L.I.; Noskov, S.V.; Tokarev, V.I.; Trubina, V.K.; Nikolaev, V.A.; Kupalov-Yaropolk, A.I.; Ivanov, A.V.
Collection of reports of the 4. Inter-industry conference on reactor materials. Volume 1. Fuel and fuel elements for power reactors1996
Collection of reports of the 4. Inter-industry conference on reactor materials. Volume 1. Fuel and fuel elements for power reactors1996
AbstractAbstract
[en] The results of measuring the release of gaseous fission products and fuel temperature by irradiation of the uranium-erbium and standard fuel are presented. The results of the above measurements from ventilated fuel elements with different fuel show that the dependences of the release on temperature, decay constant and irradiation time are similar. The release of the gaseous fission products by the fuel elements irradiation increases
[ru]
Представлены результаты измерений выхода (ГПД) и температуры топлива при облучении уран-эрбиевого и штатного топлива. Результаты измерений выхода ГПД из вентилируемых твэлов с различным топливом показывают, что зависимости выхода ГПД от температуры, постоянной распада, времени облучения одинаковы. Выход ГПД в рабочем диапазоне температур определяется диффузиейOriginal Title
Vnutrireaktornye ispytaniya maketov tvehlov RBMK s uran-ehrbievym toplivom v reaktore IVV-2M
Primary Subject
Source
Ministerstvo Rossijskoj Federatsii po Atomnoj Ehnergii, Moscow (Russian Federation); Gosudarstvennyj Nauchnyj Tsentr Rossijskoj Federatsii Nauchno-Issledovatel'skij Inst. Atomnykh Reaktorov, Dimitrovgrad (Russian Federation); 431 p; ISBN 5-85165-169-5;
; 1996; p. 221-230; 4. Inter-industry conference on reactor materials; 4. Mezhotraslevaya konferentsiya po reaktornomu materialovedeniyu; Dimitrovgrad (Russian Federation); 15-19 May 1995; 6 figs.

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Book
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Rogozyanov, A.Ya.; Kalinina, N.K.
Collection of reports of the 4. Inter-industry conference on reactor materials. Volume 1. Fuel and fuel elements for power reactors1996
Collection of reports of the 4. Inter-industry conference on reactor materials. Volume 1. Fuel and fuel elements for power reactors1996
AbstractAbstract
[en] he device for in-pile studies on the creep of fuel materials in the dispersion fuel elements at temperatures of 250-600 deg C and flux densities of thermal neutrons (2-3) x 1013 cm-2s-1 is described. It includes loading system, provision of assigned temperature, automated loading regime control and also deformation, loading and temperature measurements
[ru]
Описано устройство для внутриреакторных исследований ползучести топливных материалов дисперсионных твэлов при температурах 250-600 град. С и плотностях потока тепловых нейтронов (2-3) x 1013 см-2с-1. Оно включает в себя системы нагружения, обеспечения заданной температуры, автоматического регулирования режима нагружения, а также измерения деформации, нагрузки и температурыOriginal Title
Ustrojstvo dlya vnutrireaktornykh issledovanij polzuchesti toplivnykh materialov dispersionnykh tvehlov
Primary Subject
Source
Ministerstvo Rossijskoj Federatsii po Atomnoj Ehnergii, Moscow (Russian Federation); Gosudarstvennyj Nauchnyj Tsentr Rossijskoj Federatsii Nauchno-Issledovatel'skij Inst. Atomnykh Reaktorov, Dimitrovgrad (Russian Federation); 431 p; ISBN 5-85165-169-5;
; 1996; p. 400-404; 4. Inter-industry conference on reactor materials; 4. Mezhotraslevaya konferentsiya po reaktornomu materialovedeniyu; Dimitrovgrad (Russian Federation); 15-19 May 1995; 1 fig.

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AbstractAbstract
[en] The problem of using burnable poisons (gadolinium and erbium oxides) integrated with fuel pellets for suppression of the excess reactivity in the LWR reactor cores at fuel cycle begin when the fuel with maximum enrichment is loaded in the core is discussed. It is shown that application of the fuel elements with such pellets ensures sufficient burnup growth for fuel with increased enrichment, increase in the fuel cycle duration and decrease in neutron fluence on reactor vessel in the cases of optimized layouts of fresh and irradiated fuel assemblies in the reactor core. Basing on the analysis of studying into (U, Gd)O2 pellet heating and thermal conductivity under high burnups it is proved that the fuel with enrichment of 4.4 % of 235U may be used if the Gd2O3 content amounts to 2 %. Application of erbium absorber is recommended in uranium and plutonium fuel in inertial (nonfissible) matrix designed for burnups greater than 100 GeV · days/t
[ru]
Обсуждается проблема использования интегрированных с топливными таблетками выгорающих поглотителей (оксидов гадолиния и эрбия) для подавления избыточной реактивности активных зон реакторов LWR в начале топливного цикла при загрузке в зону топлива с большим обогащением. Показано, что использование твэлов с такими таблетками обеспечивает существенный рост выгорания топлива повышенного обогащения, увеличение длительности топливного цикла и снижение флюенса нейтронов на корпус реактора при оптимальных схемах размещения свежих и облученных ТВС в активной зоне. На основе анализа результатов исследования нагрева и теплопроводности таблеток (U, Gd)O2 при высоких выгораниях доказано, что при 2 %-ном содержании Gd2O3 можно использовать топливо с обогащением 4,4 % 235U. В плутониевом и урановом топливе с матрицей из инертного (неделящегося) материала, рассчитанном на выгорание более 100 ГВт · сут/т, рекомендуется использовать эрбиевый выгорающий поглотительOriginal Title
Primenenie integrirovannykh s toplivnymi tabletkami vygorayuchshikh poglotitelej
Primary Subject
Source
15 refs., 4 figs., 3 tabs.
Record Type
Journal Article
Journal
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ACTINIDE COMPOUNDS, ACTINIDES, BARYONS, CHALCOGENIDES, ELEMENTARY PARTICLES, ELEMENTS, ENERGY SOURCES, ENRICHED URANIUM, ENRICHED URANIUM REACTORS, ERBIUM COMPOUNDS, FERMIONS, FUELS, GADOLINIUM COMPOUNDS, GRAPHITE MODERATED REACTORS, HADRONS, INFORMATION, ISOTOPE ENRICHED MATERIALS, MATERIALS, METALS, NEUTRON ABSORBERS, NEUTRONS, NUCLEAR POISONS, NUCLEONS, OXIDES, OXYGEN COMPOUNDS, POWER REACTORS, RARE EARTH COMPOUNDS, REACTOR COMPONENTS, REACTOR MATERIALS, REACTORS, THERMAL REACTORS, URANIUM, URANIUM COMPOUNDS, URANIUM OXIDES, WATER COOLED REACTORS, WATER MODERATED REACTORS
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Secker, J.R.; Shatilla, Y.A.; Johansen, B.J.; Young, M.Y.; Sung, Y.
Westinghouse Electric Corporation, Monroeville, PA (United States)2001
Westinghouse Electric Corporation, Monroeville, PA (United States)2001
AbstractAbstract
[en] This paper describes the crud induced axial power shift phenomenon, commonly known as axial offset anomaly (AOA), and the Westinghouse Nuclear Fuel methodology developed to assess the risk potential during Loading Pattern (LP) development
Primary Subject
Source
17 Jun 2001; 3 p; 2001 Annual Meeting; Milwaukee, WI (United States); 17-21 Jun 2001; ISSN 0003-018X;
; CODEN TANSAO; Available from American Nuclear Society, P.O. Box 97781, Chicago, IL 60678 (US); Transactions of the American Nuclear Society, volume 84

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Miscellaneous
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Smith, K.S.; Grandi, G.
Studsvik of American, Idaho Falls, ID (United States)2001
Studsvik of American, Idaho Falls, ID (United States)2001
AbstractAbstract
[en] The SIMULATE-3K (S-3K) reactor analysis code has been applied to a variety of pressurized water reactor (PWR) and boiling water reactor (BWR) transients since 1993. Over the years, many changes have occurred in the S-3K channel hydraulics and ex-core component modeling. This paper summarizes those changes and outlines the status of existing vessel and steam line models. Examples are given for BWR transients that can be analyzed with S-3K
Primary Subject
Source
17 Jun 2001; 4 p; 2001 Annual Meeting; Milwaukee, WI (United States); 17-21 Jun 2001; ISSN 0003-018X;
; CODEN TANSAO; Available from American Nuclear Society, P.O. Box 97781, Chicago, IL 60678 (US); Transactions of the American Nuclear Society, volume 84

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