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Titre du document / Document title

Scaling analysis for the OSU AP600 test facility (APEX)

Auteur(s) / Author(s)

REYES J. N. (1) ; HOCHREITER L. (2) ;

Affiliation(s) du ou des auteurs / Author(s) Affiliation(s)

(1) Department of Nuclear Engineering, Oregon State University, 116 Radiation Center, Corvallis, OR 97331-5902, ETATS-UNIS
(2) Westinghouse Electric Corporation, PO Box 3, Pittsburgh, PA 15230, ETATS-UNIS

Résumé / Abstract

In this paper, the authors summarize the key aspects of a state-of-the-art scaling analysis (Reyes et al., 1995. Westinghouse Electric Corporation, WCAP-14270) performed to establish the facility design and test conditions for the Advanced Plant Experiment (APEX) at Oregon State University (OSU). This scaling analysis represents the first, and most comprehensive, application of the Hierarchical Two-Tiered Scaling (H2TS) Methodology (Zuber, 1991. US Nuclear Regulatory Commission. Washington DC, NUREG/CR-5809) in the design of an integral system test facility. The APEX test facility, designed and constructed on the basis of this scaling analysis, is the most accurate geometric representation of a Westinghouse AP600 nuclear steam supply system. The OSU APEX test facility has served to develop an essential component of the integral system database used to assess the AP600 thermal hydraulic safety analysis computer codes.

Revue / Journal Title

Nuclear engineering and design    ISSN  0029-5493   CODEN NEDEAU 

Source / Source

1998, vol. 186, no 1-2 (303 p.)  (31 ref.), pp. 53-109

Langue / Language

Anglais

Editeur / Publisher

Elsevier, Amsterdam, PAYS-BAS  (1966) (Revue)

Mots-clés anglais / English Keywords

Nuclear power plant

;

Pressurized water reactor

;

Nuclear safety

;

LOCA

;

Passive system

;

Heat transfer

;

Methodology

;

Algorithm

;

Thermohydraulics

;

Flow regime

;

Theoretical study

;

Modeling

;

Scaling analysis

;

Mots-clés français / French Keywords

Centrale nucléaire

;

Réacteur eau pressurisée

;

Sûreté nucléaire

;

Accident LOCA

;

Système passif

;

Transfert chaleur

;

Méthodologie

;

Algorithme

;

Thermohydraulique

;

Régime écoulement

;

Etude théorique

;

Modélisation

;

Analyse échelle

;

Mots-clés espagnols / Spanish Keywords

Central nuclear

;

Reactor agua a presión

;

Seguridad nuclear

;

Accidente LOCA

;

Sistema pasivo

;

Transferencia térmica

;

Metodología

;

Algoritmo

;

Termohidraúlica

;

Régimen flujo

;

Estudio teórico

;

Modelización

;

Localisation / Location

INIST-CNRS, Cote INIST : 12262, 35400007311153.0040

Nº notice refdoc (ud4) : 1643886



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