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Fuketa, Toyoshi; Yamano, Norihiro; Inoue, Akira
Proceedings of the CSNI specialists meeting on fuel-coolant interactions1994
Proceedings of the CSNI specialists meeting on fuel-coolant interactions1994
AbstractAbstract
[en] During a reactivity initiated accident (RIA) a large and prompt amount of energy is deposited within fuel rods. Consequent fuel melting and rod failure can lead to a fine dispersal of fuel melt in coolant, resulting in a violent thermal interaction between the fuel melt and coolant (i.e. fuel/coolant interaction--FCI). The generation of destructive forces during FCIs under RIA conditions have been demonstrated by the in-pile experiment program in Japan Atomic Energy Research Institute. The FCIs in an RIA have been also studied in Tokyo Institute of Technology by out-of-pile experiments and analytical modeling. The outlines and primary results of both the programs are described and discussed in this paper
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Nuclear Regulatory Commission, Washington, DC (United States). Div. of Systems Research; Nuclear Energy Agency, 75 - Paris (France); California Univ., Santa Barbara, CA (United States). Center for Risk Studies and Safety; 370 p; Mar 1994; p. 282-295; Specialist meeting on fuel-coolant interactions; Santa Barbara, CA (United States); 5-7 Jan 1993; Also available from OSTI as TI94009269; NTIS; GPO
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Podowski, M.
Proceedings of the topical meeting on advances in reactor physics and core thermal hydraulics1982
Proceedings of the topical meeting on advances in reactor physics and core thermal hydraulics1982
AbstractAbstract
[en] The analysis of stability in bounded domains of initial perturbations is presented for nonlinear point kinetics models. It is shown that the method proposed can also be used to establish relationships between the magnitude of initial trajectories and that of the reactor responses. In particular, an analytical expression is derived for the peak power of a reactor subjected to a sudden reactivity increase. The results of theoretical evaluation are compared with those of computer calculations, showing a good agreement. Some extensions to nodalized space dependent models are discussed
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American Nuclear Society. Northeastern New York Section; p. 671-683; Aug 1982; p. 671-683; NRC meeting on advances in reactor physics and core thermal hydraulics; Kiamesha Lake, NY (USA); 22 - 24 Sep 1982; Available from NTIS, PC A99/MF A01 - GPO $13.00 as DE82906282
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AbstractAbstract
[en] Short communication; 3 refs.; and, 2 figures
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Canadian Nuclear Association, Toronto, ON (Canada); Canadian Nuclear Society, Toronto, ON (Canada); 311 p; 1991; p. 4.3-9-4.3-11; 31. Canadian Nuclear Association annual conference; Saskatoon, SK (Canada); 9-12 Jun 1991; 12. Canadian Nuclear Society annual conference; Saskatoon, SK (Canada); 9-12 Jun 1991
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Miscellaneous
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AbstractAbstract
[en] A method for at-power reactivity measurements is presented. The method is based on a trapezoidal perturbation of the reactor state, which can be generated easily by means of ordinary control rods. Both theoretical and experimental work are discussed, and a simple algorithm for the correction of space-time effects is suggested
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Nuclear Science and Engineering; ISSN 0029-5639;
; v. 71(3); p. 309-318

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AbstractAbstract
[en] KWU's long time experiences with fuel elements are demonstrating that errors made in the beginning, could be extensively eliminated by improved manufacturing methods. In regard to power ramps, upper boundary values are presupposed in the modern PWR's of the KWU by the reactor efficiency limiting system. Start-up ramps have never caused damages on fuel elements after a fuel changing, at medium efficiency increase rates of < 5%/h, and this load change speed is therefore recommended as a guide value for a safe restart operation. Ramp malfunctions however occurred in one cycle of an older PWR as a result of an extraordinary regulation process, and in several cycles of an older BWR, however with decreasing frequency. The carefully performed check-ups and analyses showed that prepressurization of the fuel rods with helium decreases the susceptibility against ramp malfunctions; with other words, the limitation under which no damages occurred is higher now. KWU performs several research reactors - partially in cooperation with other institutions - during which a large parametric field will be examined. Some of the studies of the still incomplete check-ups, on fuel rods of the first experiments will be described in the lecture. The model development shall indicate the fuel rod chemistry as well as the fuel rod mechanic, and requires a correct description of the initial state of the fuel rod, which is depending on the efficiency history, burning off and the design. The interference between pellet and cladding tube is seen as a simple comprehensible parameter, and the results will be classified in accordance with this parameter. According to these results, a satisfactory fuel element behaviour for all KWU light water reactors is expected, under all operational conditions. (orig.)
[de]
Wie die langjaehrige KWU-Brennelement-Erfahrung zeigt, konnten fruehere Anfangsfehler durch verbesserte Fabrikationsmethoden weitgehend eliminiert werden. Bezueglich Leistungsrampen sind in den modernen Druckwasserreaktoren der KWU durch das Reaktorleistungsbegrenzungssystem obere Grenzwerte vorgegeben. Anfahrrampen nach Brennelementwechsel haben bei mittleren Leistungssteigerungsraten vpm < 5%/h nie zu Brennstabschaeden gefuehrt und diese Lastaenderungsgeschwindigkeit wird daher als Richtwert fuer einen sicheren Wiederanfahrbetrieb empfohlen. Rampenfehler sind in einem Zyklus eines aelteren Druckwasserreaktors als Folge eines aussergewoehnlichen Regelvorganges und in mehreren Zyklen eines aelteren Siedewasserreaktors jedoch mit abnehmender Haeufigkeit aufgetreten. Die sorgfaeltigen Nachuntersuchungen und Analysen geben den Hinweis, dass Vorinnendruck der Brennstaebe mit Helium die Anfaelligkeit gegen Rampenfehler vermindert, d.h. die Grenze unterhalb der keine Defekte zu beobachten sind, hoeher schiebt. KWU fuehrt, zum Teil in Zusammenarbeit mit anderen Institutionen, mehrere Versuchsprogramme in Leistungs- und Forschungsreaktoren durch, bei denen ein weites Parameterfeld untersucht wird. Einige Beobachtungen der noch unvollstaendigen Nachuntersuchungen an Brennstaeben der ersten Experimente werden mitgeteilt. Die Modellentwicklung muss sowohl die Brennstabchemie als auch die Brennstabmechanik umfassen und bedarf einer korrekten Beschreibung des Ausgangszustands des Brennstabs, der von Leistungsgeschichte, Abbrand und Auslegung abhaengt. Als einfach ueberschaubarer Parameter wird die Interferenz zwischen Pellet und Huellrohr angesehen, und die Ergebnisse werden anhand dieses Parameters geordnet. Die Ergebnisse lassen fuer alle KWU-Leichtwasserreaktoren ein gutes Brennelementverhalten unter allen Betriebsbedingungen erwarten. (orig.)Original Title
Das Rampenverhalten von Brennelementen
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VGB Technische Vereinigung der Grosskraftwerksbetreiber e.V., Essen (Germany, F.R.); p. 123-142; 1977; p. 123-142; VGB congress on power plants and VGB annual general meeting; Copenhagen, Denmark; 30 Aug - 2 Sep 1977; AED-CONF--77-296-012
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AbstractAbstract
No abstract available
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23 Jan 1974; 61 p; Published in summary form only.
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[en] In this paper the mechanical and magnetic layout of the first three insertion devices for DORIS III, an upgraded reconstruction of DORIS II, is described and results of the magnetic characterization are given as well
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12. international conference on magnet technology; Leningrad (USSR); 23-28 Jun 1991; CONF-910662--
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No abstract available
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Annual meeting of the American Nuclear Society; Philadelphia, PA; 23 Jun 1974; Published in summary form only.
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Trans. Amer. Nucl. Soc; v. 18 p. 224-225
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[en] The probability table method, originally developed for unresolved resonance calculations, is adapted to the calculation of bubble worths in a reactor core. Unlike previous approaches to this problem, this method is applicable to a broad class of problems, avoids a high variance Monte Carlo calculation, and preserves the main features of a random bubble distribution. This approach is illustrated by the solution to several problems of varying degrees of complexity
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LMFBR
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Nuclear Science and Engineering; v. 66(1); p. 67-74
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Warnemuende, R.
Technische Univ. Muenchen, Garching (Germany, F.R.). Lab. fuer Reaktorregelung und Anlagensicherung1976
Technische Univ. Muenchen, Garching (Germany, F.R.). Lab. fuer Reaktorregelung und Anlagensicherung1976
AbstractAbstract
[en] The prompt critical Point Kinetical Equations are solved for linear and quadratic temperature feed back for 'Step Input' reactivities in the frame of the Hansen-Fuchs-Model. The influence of the precursor neutrons on energy in the power maximum can be calculated in the first approximation. Comparisons of AIREK-Code calculations and KEWB experiments confirm in the validity of neglecting the precursor neutrons for input reactivities rsub(o) >= 1.2[S]. The analytical considerations are mostly conservative and agree with the KEWB experiments within 30%. Finally, the magnitude and influence of the parameters are described, which govern the 'Step Input' excursions. (orig.)
[de]
Es werden die 'punktkinetischen Gleichungen' im Rahmen des Hansen-Fuchs-Modells bei linearer und quadratischer Temperaturrueckkoppelung fuer sprungfoermige Reaktivitaetszufuhr geloest. Die verzoegerten Neutronen koennen bei der analytischen Berechnung der Energie im Leistungsmaximum in erster Naeherung beruecksichtigt werden. Vergleiche mit dem AIREK-Code anhand von Parametern aus dem KEWB-Programm bestaetigen die berechtigte Vernachlaessigung der verzoegerten Neutronen fuer Eingabereaktivitaeten rsub(o) >= 1,2[S]. Die analytischen Nachrechnungen sind meist konservativ und stimmen im allgemeinen bis auf 30% mitden KEWB-Experimenten ueberein. (orig.)Original Title
Exkursionen in Spaltstoffloesungen durch sprungfoermige Reaktivitaetszufuhr
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Aug 1976; vp; 21 figs.; 13 tabs.; 7 refs.
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