Analysis of the NUPEC PSBT Tests with FLICA-OVAP

This paper discusses the results of a computational activity devoted to the prediction of two-phase flows in subchannels and in rod bundles. The capabilities of the FLICA-OVAP code have been tested against an extensive experimental database made available by the Japanese Nuclear Power Energy Corpora...

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Main Authors: Matteo Bucci, Philippe Fillion
Format: Article
Language:English
Published: Wiley 2012-01-01
Series:Science and Technology of Nuclear Installations
Online Access:http://dx.doi.org/10.1155/2012/436142
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author Matteo Bucci
Philippe Fillion
author_facet Matteo Bucci
Philippe Fillion
author_sort Matteo Bucci
collection DOAJ
description This paper discusses the results of a computational activity devoted to the prediction of two-phase flows in subchannels and in rod bundles. The capabilities of the FLICA-OVAP code have been tested against an extensive experimental database made available by the Japanese Nuclear Power Energy Corporation (NUPEC) in the frame of the PWR subchannel and bundle tests (PSBT) international benchmark promoted by OECD and NRC. The experimental tests herein addressed involve void fraction distributions and boiling crisis phenomena in rod bundles with uniform and nonuniform heat flux conditions. Both steady-state and transient scenarios have been addressed, including power increase, flow reduction, temperature increase, and depressurization, representative of PWR thermal-hydraulics conditions. After a brief description of the main features of FLICA-OVAP, the relevant physical models available within the code are detailed. Results obtained in the different tests included in the PSBT void distribution and DNB benchmarks are therefore reported. The relevant role of selected physical models is discussed.
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spelling doaj-art-2d2d00f1855642e9be52b11c850c8d7a2025-02-03T05:46:28ZengWileyScience and Technology of Nuclear Installations1687-60751687-60832012-01-01201210.1155/2012/436142436142Analysis of the NUPEC PSBT Tests with FLICA-OVAPMatteo Bucci0Philippe Fillion1CEA, DEN, DM2S/STMF, 91191 Gif-sur-Yvette, FranceCEA, DEN, DM2S/STMF, 91191 Gif-sur-Yvette, FranceThis paper discusses the results of a computational activity devoted to the prediction of two-phase flows in subchannels and in rod bundles. The capabilities of the FLICA-OVAP code have been tested against an extensive experimental database made available by the Japanese Nuclear Power Energy Corporation (NUPEC) in the frame of the PWR subchannel and bundle tests (PSBT) international benchmark promoted by OECD and NRC. The experimental tests herein addressed involve void fraction distributions and boiling crisis phenomena in rod bundles with uniform and nonuniform heat flux conditions. Both steady-state and transient scenarios have been addressed, including power increase, flow reduction, temperature increase, and depressurization, representative of PWR thermal-hydraulics conditions. After a brief description of the main features of FLICA-OVAP, the relevant physical models available within the code are detailed. Results obtained in the different tests included in the PSBT void distribution and DNB benchmarks are therefore reported. The relevant role of selected physical models is discussed.http://dx.doi.org/10.1155/2012/436142
spellingShingle Matteo Bucci
Philippe Fillion
Analysis of the NUPEC PSBT Tests with FLICA-OVAP
Science and Technology of Nuclear Installations
title Analysis of the NUPEC PSBT Tests with FLICA-OVAP
title_full Analysis of the NUPEC PSBT Tests with FLICA-OVAP
title_fullStr Analysis of the NUPEC PSBT Tests with FLICA-OVAP
title_full_unstemmed Analysis of the NUPEC PSBT Tests with FLICA-OVAP
title_short Analysis of the NUPEC PSBT Tests with FLICA-OVAP
title_sort analysis of the nupec psbt tests with flica ovap
url http://dx.doi.org/10.1155/2012/436142
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