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Hot channel factors evaluation for thermalhydraulic analysis of MTR reactors

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Warianty tytułu
Języki publikacji
EN
Abstrakty
EN
This paper addresses the hot channel factors for the thermalhydraulic analysis of Material Testing Reactors (MTR). The criteria for choosing the hot channel factors and their combination methods are presented and discussed. A method for the evaluation of the hot channel factors for the thermalhydraulic analysis is proposed. In the proposed method, the values of the hot channel factor are calculated in terms of their fraction of variation and the degree of dependency of the thermalhydraulic safety parameter of interest on these factors. The equations for the calculation of the thermalhydraulic safety parameters in the hot channel are presented along with discussions. The application of the proposed method has been illustrated by an example.
Czasopismo
Rocznik
Strony
61--67
Opis fizyczny
Bibliogr. 11 poz., rys.
Twórcy
autor
  • Nuclear Engineering Department, Alexandria University, 21544, Alexandria, Egypt
  • ETRR-2, Atomic Energy Authority, 13759, Abou Zabal, Egypt, Tel.: +20 2 4691755, Fax: +20 2 4691754
autor
  • Nuclear Engineering Department, Alexandria University, 21544, Alexandria, Egypt
autor
  • ETRR-2, Atomic Energy Authority, 13759, Abou Zabal, Egypt, Tel.: +20 2 4691755, Fax: +20 2 4691754
Bibliografia
  • 1. Atomic Energy Authority and INVAP SE (1995) Standard Fuel Element Assembly. ETRR-2 Document 6767-0710-3AAIT-963-1A, Egypt
  • 2. Atomic Energy Authority and INVAP SE (1999) Final Safety Analysis Report. ETRR-2 Document 0767-5325-3IBLI-001-1A, Egypt
  • 3. Bokhari IH, Israr M, Pervez S (1999) Thermalhydraulic and safety analysis for Pakistan Research Reactor-1. In: Proc of the 22nd Int Meeting on Reduced Enrichment for Research and Test Reactors 1999, 3−8 October 1999, Budapest, Hungary,http://www.td.anl.gov/programs/RERTR/RERTR.html
  • 4. Fabrega J (1971) Calcul Termique des Reacteures de Recherche Refeoidis par Eau. Report CEA R-4114, France
  • 5. Gimenez M, Schlamp M, Vertullo A (2002) Uncertainties assessment for safety margins evaluation in MTR reactors core thermalhydraulic design. In: Proc of the 24th Int Meeting on Reduced Enrichment for Research and Test Reactors 2002, 3−8 November 2002, Bariloche, Argentina, http://www.td.anl.gov/programs/RERTR/RERTR.html
  • 6. INVAP SE (1995) TERMIC 1 H Mode 3.1: A Program for the Calculus and Thermalhydraulic Design of Reactor Cores, MTR_PC package, Argentina
  • 7. Matos JE, Mo SC, Woodruff WL (1992) Analysis for conversion of the Georgia Tech Research Reactor from HEU to LEU. Report ANL/RERTR/TM-19. ANL, USA
  • 8. Mirshak S, Durant WS, Towell RH (1959) Heat flux at burnout. AEC R&D Report DP-355, E I du Pont de Nemours & Co., Aiken, South Carolina, USA
  • 9. Mishima K, Kanada K, Shibata T (1990) Thermalhydraulic analysis for core conversion to the use of low enriched uranium fuels in the KUR. Report KURRI-TR-258. Research Reactors Institute, Kyoto University, Japan
  • 10. Ricque R, Siboul R (1970) Ebullition Locale de L’Eau en convection Force’e. Report CEA-R-3894, Centre d’Etudes Nucleaires de Grenoble, France
  • 11. Woodruff WL (1987) Evaluation and selection of hot channel (peaking) factors for research reactors applications. In: Proc of the 10th Int Meeting on Reduced Enrichment for Research and Test Reactors 1987, September 28 − October 1, 1987, Buenos Aires, Argentina, pp 443−452
Typ dokumentu
Bibliografia
Identyfikator YADDA
bwmeta1.element.baztech-article-BUJ6-0005-0050
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