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AbstractAbstract
[en] The selection of facility model and code options have been recognized as influencing strongly the results of calculations of accident analysis with large complex computer codes. This paper discusses the influences of models and options during RELAP4/MOD6 calculations of large and small breaks. Plant analysis and standard problem exercises are taken into consideration. (author). 13 refs, 4 figs
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International Atomic Energy Agency, Vienna (Austria); 194 p; May 1988; p. 125-131; Technical committee/workshop on computer aided safety analysis; Warsaw (Poland); 25-29 May 1987
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[en] The IAEA Programme on Computer Aided Safety Analysis is presented together with the statistics of computer utilization under the programme. Difficulties encountered during programme implementation are discussed and a reformulation is proposed as a possible solution. (author). 1 ref., 5 figs, 2 tabs
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International Atomic Energy Agency, Vienna (Austria); 194 p; May 1988; p. 11-16; Technical committee/workshop on computer aided safety analysis; Warsaw (Poland); 25-29 May 1987
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[en] The document reproduces 20 selected papers from the 38 papers presented at the Technical Committee/Workshop on Computer Aided Safety Analysis organized by the IAEA in co-operation with the Institute of Atomic Energy in Otwock-Swierk, Poland on 25-29 May 1987. A separate abstract was prepared for each of these 20 technical papers. Refs, figs and tabs
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Source
May 1988; 194 p; Technical committee/workshop on computer aided safety analysis; Warsaw (Poland); 25-29 May 1987
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[en] This paper presents results of post-test calculations relative to the first PMK-NVH Standard Problem Exercise, performed with RELAP4/MOD6 code utilizing a quite simple nodalization scheme. Influence of different modelling options on final results are discussed in some detail. (author). 5 refs, 19 figs
Primary Subject
Source
International Atomic Energy Agency, Vienna (Austria); 194 p; May 1988; p. 111-125; Technical committee/workshop on computer aided safety analysis; Warsaw (Poland); 25-29 May 1987
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AbstractAbstract
[en] A loss of flow transient test was conducted at the PMK-NVH integral test facility. This paper describes the test conditions and presents some of the experimental results. Comparison between experiments and analytical results obtained with RELAP4/MOD6 are also discussed. (author). 4 refs, figs and tabs
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Source
International Atomic Energy Agency, Vienna (Austria); 194 p; May 1988; p. 160-177; Technical committee/workshop on computer aided safety analysis; Warsaw (Poland); 25-29 May 1987
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[en] Post-test calculations of SB LOCA experiments are presented. The 7,4% SB LOCA on PMK-NVH integral at the experimental facility were analysed by computer codes RELAP4 and SLAP2, used in CSSR for nuclear safety analyses of PWRs. The calculations were performed on the EC 1045 computer at CRIP Budapest during an IAEA fellowship. (author). 4 refs, 10 figs, 1 tab
Primary Subject
Source
International Atomic Energy Agency, Vienna (Austria); 194 p; May 1988; p. 62-71; Technical committee/workshop on computer aided safety analysis; Warsaw (Poland); 25-29 May 1987
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AbstractAbstract
[en] The computerized simulations of the accident and abnormal conditions at nuclear power plants are very important tools for its safety assessment. This presentation is a survey of computer codes used in the USSR for safety analysis of WWER type reactors. It also presents a review of the experimental facilities for safety experiments in the field. The plans in safety analysis programs for today and the future are presented
Original Title
Sostoyanie raschetnykh programm i ehksperimental'nykh issledovanij po obosnovannyu bezopasnosti AEhS s WWER
Primary Subject
Source
International Atomic Energy Agency, Vienna (Austria); 194 p; May 1988; p. 181-189; Technical committee/workshop on computer aided safety analysis; Warsaw (Poland); 25-29 May 1987
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[en] The SSYST code is a modular computer code developed at Karlsruhe Nuclear Research Centre (KFK) for modelling light water reactor fuel rods. This paper describes the experiences in modelling WWER type fuel rods with SSYST at the Institute of Nuclear Research Rossendorf. Problems of coupling SSYST code with STOFFEL-1, a fuel performance modelling code, are also discussed. (author). 4 refs, 4 figs, 1 tab
Primary Subject
Source
International Atomic Energy Agency, Vienna (Austria); 194 p; May 1988; p. 141-148; Technical committee/workshop on computer aided safety analysis; Warsaw (Poland); 25-29 May 1987
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[en] In the last few years about 3% steam generator tubes were plugged at NPP Krsko. Due to tube plugging RCS flow was reduced, and also other plant parameters at 100% power were changed. For NPP Krsko one question is very important from the safety point of view: how many tubes may be plugged before the power reduction is needed. To answer this question it is required to perform or evaluate the events of Chapter 15 of NPP Krsko Final Safety Analysis Report. In this article turbine trip is analysed after 5% and 10% SG tubes were plugged. The analysis is performed using ALMOD3W2 code. (author). 4 refs, 3 figs
Primary Subject
Source
International Atomic Energy Agency, Vienna (Austria); 194 p; May 1988; p. 155-160; Technical committee/workshop on computer aided safety analysis; Warsaw (Poland); 25-29 May 1987
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AbstractAbstract
[en] This report summarizes the criteria adopted to perform the post-test calculations for the PMK-NVH Standard Problem Test, sponsored by the IAEA. The nodalization corresponding to the TRAC-PF1 code runs is shown, the correlations adopted are specified, a comparison of the experimental and simulated conditions is given and, finally, the graphs of the time evolution of the thermohydraulic parameters corresponding to the specified table of initiating events are shown, as well as the comparison of predicted and experimental major events. A discussion of results is provided to consider some differences between the calculated and the experimental results. The results obtained with the TRAC-PF1 code show good agreement versus the experimental ones. (author). 1 ref., 11 figs, 3 tabs
Primary Subject
Source
International Atomic Energy Agency, Vienna (Austria); 194 p; May 1988; p. 72-78; Technical committee/workshop on computer aided safety analysis; Warsaw (Poland); 25-29 May 1987
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