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AbstractAbstract
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nd; 10 p
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Report
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Progress Report
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Hetrick, David L. (ed.); p. 383-404; 1972; Univ. of Arizona Press; Tucson, Ariz; Symposium on dynamics of nuclear systems; Tucson, Ariz; 23 Mar 1970
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Book
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Hinis, E. P.; Simopulous, S.E.
Proceedings; Yugoslav Nuclear Society ; Institute of Nuclear Sciences VINCA1997
Proceedings; Yugoslav Nuclear Society ; Institute of Nuclear Sciences VINCA1997
AbstractAbstract
[en] The rewetting of the overheated water cooled nuclear reactor fuel rods after a Loss-of-Coolant-Accident, which is defined as the re-establishment of a water film on the hot rod surfaces, is an important process for the safety of nuclear reactors. Several water experiments have been carried out to investigate the rewetting phenomena and to predict the liquid film behaviour either at atmospheric conditions or in a steam environment. The present paper describes the experimental work aimed at studying the wet front propagation along a stainless steel fuel rod in a top-flooding saturated water-steam environment in the pressure range 1-7 bar, under various initial wall temperatures up to 5500 C and at a liquid flowrate of 1 l min1. Following the experimental results a correlation has been derived to predict the rewetting rate as a function of pressure and initial wall temperature. This correlation agrees well with formerly proposed correlations at higher pressures, thus permitting to extend their validity down to the pressure of 1 bar. Furthermore, a numerical method is introduced to experimentally evaluate the wet front position along the rod.(author)
Primary Subject
Source
Antic, D. (ed.) (Institut za Nuklearne Nauke VINCA, Belgrade (Yugoslavia)); 746 p; ISBN 86-7306-012-5;
; 1997; p. 209-216; Institute of Nuclear Sciences VINCA; Belgrade (Yugoslavia); Yugoslav Nuclear Society Conference (YUNSC'96); Jugoslovensko nuklearno drustvo konferencija, zbornik radova; Belgrade (Yugoslavia); 6-9 Oct 1996; 9 refs., 4 figs.

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Book
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AbstractAbstract
No abstract available
Original Title
Crashbauweise zum Schutz gegen Flugzeugabsturz
Primary Subject
Source
Kerntechnische Gesellschaft im Deutschen Atomforum e.V., Bonn (Germany, F.R.); p. 391-394; 1978; p. 391-394; ZAED; Eggenstein-Leopoldshafen, Germany, F.R; Reactor congress; Hannover, Germany, F.R; 4 - 7 Apr 1978; AED-CONF--78-006-097; 3 figs.; 2 tabs. Short communication only.
Record Type
Book
Literature Type
Conference
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INIS VolumeINIS Volume
INIS IssueINIS Issue
Kucukboyaci, V. N.
Hacettepe Univ., Fen Bilimleri Enstitusu, Ankara (Turkey)1995
Hacettepe Univ., Fen Bilimleri Enstitusu, Ankara (Turkey)1995
AbstractAbstract
[en] The main object of this study is to perform thermal-hydraulic analyses of small (1 % of cold leg flow area), intermediate (30%). and large (100%) break loss of coolant accident (LOCA) of AP600 reactor system. Detailed examination ot AP600 thermal-hydraulic parameters during these transients is covered. Behaviour ot the system components under accident conditions is observed. Sufficiency and effectiveness of the passive safety systems are evaluated. TRAC-PF1 which is a best -estimate ode is used for for Ihe transient analyses. A detailed model is established to describe the AP600 features.Control of the system components is achieved by specifying proper setpoints, boundary and initial cconditions. It has been determined that the performance of the AP600 system in LOCA cases is satisfactory. It is observed that for small and intermediate breaks the system is succesfully depressurized without any core uncovery, and that the peak clad temperature do not exceed the steady-state values. For the design basis large break LOCA, the peak clad temperature during the accident is 1239 K which is below the design limit of 1477 K
Original Title
AP600 reaktorunun sogutucu kaybi kazasi analizi
Primary Subject
Source
1995; 80 p; ISBN 975-491-046-4;
; Available from Hacettepe Univ., Fen Bilimleri Enstitusu, Ankara (TR); Thesis (Ms)

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AbstractAbstract
[en] Full text: One of the major design basis accident is loss of coolant accident (LOCA). This accident is characterized by high sheath temperature and rapid depressurization of coolant. One of the ECCS acceptance criteria for this type of accident is for all LOCA events the release of radioactive material from the fuel in the reactor shall be limited such that the calculated radiological consequences at exclusion zone boundary are within acceptable limits for accidents as specified by AERB. The assessment of above criteria requires the estimation of fuel failure. Two phenomena which cause fuel failure during such condition are oxygen embrittlement and clad ballooning. Oxygen embrittlement is assessed by the amount of oxidation. Failure due to ballooning is analysed by calculating stress and strain on the clad and comparing with ultimate burst stress. Failure data generated by analysing these two phenomena are used in assessing damage on fuel. This paper discusses the criteria of fuel failure followed by various regulating agencies or countries along with Indian criteria. This paper also discusses the damage phenomena along with detailed method of assessment of fuel failure during LOCA
Primary Subject
Source
Gupta, Satish K. (comp.) (Reactor Safety Div., Bhabha Atomic Research Centre, Mumbai (India)); Board of Research in Nuclear Sciences, Department of Atomic Energy, Mumbai (India); 421 p; Nov 2002; p. 367-368; NRT-1: 1. nuclear reactor safety; Mumbai (India); 25-27 Nov 2002
Record Type
Book
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Conference
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Shah, H.C.; Benjamin, J.R.
International seminar on extreme load conditions and limit analysis procedures for structural reactor safeguards and containment structures1975
International seminar on extreme load conditions and limit analysis procedures for structural reactor safeguards and containment structures1975
AbstractAbstract
No abstract available
Primary Subject
Secondary Subject
Source
Jaeger, T.A. (comp.); International Association for Structural Mechanics in Reactor Technology; Bundesanstalt fuer Materialpruefung, Berlin (F.R. Germany); p. O2/2; 1975; International seminar on extreme load conditions and limit analysis procedures for structural reactor safeguards and containment structures; Berlin, F.R. Germany; 8 Sep 1975; AED-CONF--75-365-003; Short communication only. Available from BAM.
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Report
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AbstractAbstract
[en] Conclusions: Proposed V-1 ALS can cope with 2 x DN 500 DBA. Work on V-1 ALS able to cope with DN 200 DBA and to mitigate DN 500 BDBA will have been finished by April 1994. Further leaktightness improvements are going on on the V-1 HZ boundary during outages with the aim to reduce the existing leak rate by 50%
Primary Subject
Source
International Atomic Energy Agency, Division of Nuclear Safety, Vienna (Austria); 386 p; 2 Feb 1994; p. 107-147; Consultants' meeting on containment and confinement performance in NPPS with WWER 440/213 and 440/230 reactors; Vienna (Austria); 29 Nov - 3 Dec 1993; 24 figs
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Report
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A. K. MAJI; B. MARSHALL; D. V. RAO; B. LETELLIER; C. SHAFFER
Los Alamos National Lab., NM (United States). Funding organisation: US Department of Energy (United States)2001
Los Alamos National Lab., NM (United States). Funding organisation: US Department of Energy (United States)2001
AbstractAbstract
No abstract available
Primary Subject
Source
1 Mar 2001; 3400 Kilobytes; W--7405-ENG-36; Available from PURL: https://www.osti.gov/servlets/purl/777008-8Vs6jP/native/
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Report
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AbstractAbstract
[en] 1) Introduction: Summary of system technology and reactor protection equipment. 2) Definitions. 3) LOCA: a) blowdown and refilling phase; b) jet and reaction forces; c) flow and heat transfer behavior in the core; d) behavior of the heater rods; e) core melting. 4) Protection against and during LOCA: a) general measures; b) break of a primary coolant pipe; c) break of a small pipe; d) break of a secondary pipe. (orig.)
[de]
1) Einfuehrung: Beschreibung der Systemtechnik und der Reaktorschutzsysteme. 2) Definitionen. 3) LOCA: a) Blowdown- und Wiederauffuellphase; b) Strahl- und Reaktionskraefte; c) Stroemungs- und Waermetransportverhalten; d) im Reaktorkern; e) Verhalten der Heizstaebe; f) Kernschmelzen. 4) Schutz gegen und waehrend LOCAS: a) allgemeine Massnahmen; b) Bruch eines Rohrs im Primaerkuehlkreislauf; c) Bruch eines kleinen Rohrs; d) Bruch eines Rohrs im Sekundaerkuehlkreislauf. (orig.)Primary Subject
Secondary Subject
Source
1977; 47 p; IAEA interregional training course on construction and operation management of a nuclear power plant; Karlsruhe, Germany, F.R; 5 Sep - 25 Nov 1977; 13 figs.; 3 tabs.
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Miscellaneous
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Conference
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