1/3D modeling of the core coolant circuit of a PHWR nuclear power plant

Autores
Corzo, Santiago Francisco; Ramajo, Damian Enrique; Nigro, Norberto Marcelo
Año de publicación
2015
Idioma
inglés
Tipo de recurso
artículo
Estado
versión publicada
Descripción
A multi-dimensional computational fluid dynamics (CFD) one-phase model to simulate the incore coolant circuit of a Pressurized Heavy Water Reactor (PHWR) of a nuclear power plant (NPP) was performed. Three-dimensional (3D) detailed modeling of the upper and lower plenums, the downcomer and the hot and cold leg nozzles was combined with finite volume one-dimensional (1D) code for modeling the behavior of all the 451 coolant channels. Suitable functions for introducing the distributed (friction losses) and concentrated (spacer grids, inlet restrictors and outlet throttles) pressure losses were used to consider the local pressure variation along the coolant channels. The special power distribution at each coolant channel was also taken into account. Results were compared with those previously obtained with a 0/3D model getting more realistic temperature patterns at the upper plenum. Although the present model is restricted to one-phase phenomena, the prediction of the local pressure and temperature along the channels allows for a preliminary identification of the location of incipient boiling by comparing with the local saturation temperature. The present model represents an improvement with respect to the previous 0/3D model. It corresponds to the necessary step before achieving a 1/3D two-phase model with which the pressure drop and subcooled boiling along the coolant channels as well as the overall reactor pressure vessel (RPV) void fraction distribution can be evaluated more accurately.
Fil: Corzo, Santiago Francisco. Consejo Nacional de Investigaciones Científicas y Técnicas. Centro Científico Tecnológico Conicet - Santa Fe. Centro de Investigaciones En Metodos Computacionales. Universidad Nacional del Litoral. Centro de Investigaciones En Metodos Computacionales; Argentina. Autoridad Regulatoria Nuclear; Argentina
Fil: Ramajo, Damian Enrique. Consejo Nacional de Investigaciones Científicas y Técnicas. Centro Científico Tecnológico Conicet - Santa Fe. Centro de Investigaciones En Metodos Computacionales. Universidad Nacional del Litoral. Centro de Investigaciones En Metodos Computacionales; Argentina
Fil: Nigro, Norberto Marcelo. Consejo Nacional de Investigaciones Científicas y Técnicas. Centro Científico Tecnológico Conicet - Santa Fe. Centro de Investigaciones En Metodos Computacionales. Universidad Nacional del Litoral. Centro de Investigaciones En Metodos Computacionales; Argentina
Materia
1/3d Modeling
Phwr
Flow And Thermal Distribution
Nivel de accesibilidad
acceso abierto
Condiciones de uso
https://creativecommons.org/licenses/by-nc-sa/2.5/ar/
Repositorio
CONICET Digital (CONICET)
Institución
Consejo Nacional de Investigaciones Científicas y Técnicas
OAI Identificador
oai:ri.conicet.gov.ar:11336/19561
Acceder
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network_name_str CONICET Digital (CONICET)
spelling 1/3D modeling of the core coolant circuit of a PHWR nuclear power plantCorzo, Santiago FranciscoRamajo, Damian EnriqueNigro, Norberto Marcelo1/3d ModelingPhwrFlow And Thermal Distributionhttps://purl.org/becyt/ford/2.3https://purl.org/becyt/ford/2A multi-dimensional computational fluid dynamics (CFD) one-phase model to simulate the incore coolant circuit of a Pressurized Heavy Water Reactor (PHWR) of a nuclear power plant (NPP) was performed. Three-dimensional (3D) detailed modeling of the upper and lower plenums, the downcomer and the hot and cold leg nozzles was combined with finite volume one-dimensional (1D) code for modeling the behavior of all the 451 coolant channels. Suitable functions for introducing the distributed (friction losses) and concentrated (spacer grids, inlet restrictors and outlet throttles) pressure losses were used to consider the local pressure variation along the coolant channels. The special power distribution at each coolant channel was also taken into account. Results were compared with those previously obtained with a 0/3D model getting more realistic temperature patterns at the upper plenum. Although the present model is restricted to one-phase phenomena, the prediction of the local pressure and temperature along the channels allows for a preliminary identification of the location of incipient boiling by comparing with the local saturation temperature. The present model represents an improvement with respect to the previous 0/3D model. It corresponds to the necessary step before achieving a 1/3D two-phase model with which the pressure drop and subcooled boiling along the coolant channels as well as the overall reactor pressure vessel (RPV) void fraction distribution can be evaluated more accurately.Fil: Corzo, Santiago Francisco. Consejo Nacional de Investigaciones Científicas y Técnicas. Centro Científico Tecnológico Conicet - Santa Fe. Centro de Investigaciones En Metodos Computacionales. Universidad Nacional del Litoral. Centro de Investigaciones En Metodos Computacionales; Argentina. Autoridad Regulatoria Nuclear; ArgentinaFil: Ramajo, Damian Enrique. Consejo Nacional de Investigaciones Científicas y Técnicas. Centro Científico Tecnológico Conicet - Santa Fe. Centro de Investigaciones En Metodos Computacionales. Universidad Nacional del Litoral. Centro de Investigaciones En Metodos Computacionales; ArgentinaFil: Nigro, Norberto Marcelo. Consejo Nacional de Investigaciones Científicas y Técnicas. Centro Científico Tecnológico Conicet - Santa Fe. Centro de Investigaciones En Metodos Computacionales. Universidad Nacional del Litoral. Centro de Investigaciones En Metodos Computacionales; ArgentinaElsevier2015-04info:eu-repo/semantics/articleinfo:eu-repo/semantics/publishedVersionhttp://purl.org/coar/resource_type/c_6501info:ar-repo/semantics/articuloapplication/pdfapplication/pdfhttp://hdl.handle.net/11336/19561Corzo, Santiago Francisco; Ramajo, Damian Enrique; Nigro, Norberto Marcelo; 1/3D modeling of the core coolant circuit of a PHWR nuclear power plant; Elsevier; Annals Of Nuclear Energy; 83; 4-2015; 386-3970306-4549CONICET DigitalCONICETenginfo:eu-repo/semantics/altIdentifier/doi/10.1016/j.anucene.2014.12.035info:eu-repo/semantics/openAccesshttps://creativecommons.org/licenses/by-nc-sa/2.5/ar/reponame:CONICET Digital (CONICET)instname:Consejo Nacional de Investigaciones Científicas y Técnicas2025-10-22T11:02:39Zoai:ri.conicet.gov.ar:11336/19561instacron:CONICETInstitucionalhttp://ri.conicet.gov.ar/Organismo científico-tecnológicoNo correspondehttp://ri.conicet.gov.ar/oai/requestdasensio@conicet.gov.ar; lcarlino@conicet.gov.arArgentinaNo correspondeNo correspondeNo correspondeopendoar:34982025-10-22 11:02:39.786CONICET Digital (CONICET) - Consejo Nacional de Investigaciones Científicas y Técnicasfalse
dc.title.none.fl_str_mv 1/3D modeling of the core coolant circuit of a PHWR nuclear power plant
title 1/3D modeling of the core coolant circuit of a PHWR nuclear power plant
spellingShingle 1/3D modeling of the core coolant circuit of a PHWR nuclear power plant
Corzo, Santiago Francisco
1/3d Modeling
Phwr
Flow And Thermal Distribution
title_short 1/3D modeling of the core coolant circuit of a PHWR nuclear power plant
title_full 1/3D modeling of the core coolant circuit of a PHWR nuclear power plant
title_fullStr 1/3D modeling of the core coolant circuit of a PHWR nuclear power plant
title_full_unstemmed 1/3D modeling of the core coolant circuit of a PHWR nuclear power plant
title_sort 1/3D modeling of the core coolant circuit of a PHWR nuclear power plant
dc.creator.none.fl_str_mv Corzo, Santiago Francisco
Ramajo, Damian Enrique
Nigro, Norberto Marcelo
author Corzo, Santiago Francisco
author_facet Corzo, Santiago Francisco
Ramajo, Damian Enrique
Nigro, Norberto Marcelo
author_role author
author2 Ramajo, Damian Enrique
Nigro, Norberto Marcelo
author2_role author
author
dc.subject.none.fl_str_mv 1/3d Modeling
Phwr
Flow And Thermal Distribution
topic 1/3d Modeling
Phwr
Flow And Thermal Distribution
purl_subject.fl_str_mv https://purl.org/becyt/ford/2.3
https://purl.org/becyt/ford/2
dc.description.none.fl_txt_mv A multi-dimensional computational fluid dynamics (CFD) one-phase model to simulate the incore coolant circuit of a Pressurized Heavy Water Reactor (PHWR) of a nuclear power plant (NPP) was performed. Three-dimensional (3D) detailed modeling of the upper and lower plenums, the downcomer and the hot and cold leg nozzles was combined with finite volume one-dimensional (1D) code for modeling the behavior of all the 451 coolant channels. Suitable functions for introducing the distributed (friction losses) and concentrated (spacer grids, inlet restrictors and outlet throttles) pressure losses were used to consider the local pressure variation along the coolant channels. The special power distribution at each coolant channel was also taken into account. Results were compared with those previously obtained with a 0/3D model getting more realistic temperature patterns at the upper plenum. Although the present model is restricted to one-phase phenomena, the prediction of the local pressure and temperature along the channels allows for a preliminary identification of the location of incipient boiling by comparing with the local saturation temperature. The present model represents an improvement with respect to the previous 0/3D model. It corresponds to the necessary step before achieving a 1/3D two-phase model with which the pressure drop and subcooled boiling along the coolant channels as well as the overall reactor pressure vessel (RPV) void fraction distribution can be evaluated more accurately.
Fil: Corzo, Santiago Francisco. Consejo Nacional de Investigaciones Científicas y Técnicas. Centro Científico Tecnológico Conicet - Santa Fe. Centro de Investigaciones En Metodos Computacionales. Universidad Nacional del Litoral. Centro de Investigaciones En Metodos Computacionales; Argentina. Autoridad Regulatoria Nuclear; Argentina
Fil: Ramajo, Damian Enrique. Consejo Nacional de Investigaciones Científicas y Técnicas. Centro Científico Tecnológico Conicet - Santa Fe. Centro de Investigaciones En Metodos Computacionales. Universidad Nacional del Litoral. Centro de Investigaciones En Metodos Computacionales; Argentina
Fil: Nigro, Norberto Marcelo. Consejo Nacional de Investigaciones Científicas y Técnicas. Centro Científico Tecnológico Conicet - Santa Fe. Centro de Investigaciones En Metodos Computacionales. Universidad Nacional del Litoral. Centro de Investigaciones En Metodos Computacionales; Argentina
description A multi-dimensional computational fluid dynamics (CFD) one-phase model to simulate the incore coolant circuit of a Pressurized Heavy Water Reactor (PHWR) of a nuclear power plant (NPP) was performed. Three-dimensional (3D) detailed modeling of the upper and lower plenums, the downcomer and the hot and cold leg nozzles was combined with finite volume one-dimensional (1D) code for modeling the behavior of all the 451 coolant channels. Suitable functions for introducing the distributed (friction losses) and concentrated (spacer grids, inlet restrictors and outlet throttles) pressure losses were used to consider the local pressure variation along the coolant channels. The special power distribution at each coolant channel was also taken into account. Results were compared with those previously obtained with a 0/3D model getting more realistic temperature patterns at the upper plenum. Although the present model is restricted to one-phase phenomena, the prediction of the local pressure and temperature along the channels allows for a preliminary identification of the location of incipient boiling by comparing with the local saturation temperature. The present model represents an improvement with respect to the previous 0/3D model. It corresponds to the necessary step before achieving a 1/3D two-phase model with which the pressure drop and subcooled boiling along the coolant channels as well as the overall reactor pressure vessel (RPV) void fraction distribution can be evaluated more accurately.
publishDate 2015
dc.date.none.fl_str_mv 2015-04
dc.type.none.fl_str_mv info:eu-repo/semantics/article
info:eu-repo/semantics/publishedVersion
http://purl.org/coar/resource_type/c_6501
info:ar-repo/semantics/articulo
format article
status_str publishedVersion
dc.identifier.none.fl_str_mv http://hdl.handle.net/11336/19561
Corzo, Santiago Francisco; Ramajo, Damian Enrique; Nigro, Norberto Marcelo; 1/3D modeling of the core coolant circuit of a PHWR nuclear power plant; Elsevier; Annals Of Nuclear Energy; 83; 4-2015; 386-397
0306-4549
CONICET Digital
CONICET
url http://hdl.handle.net/11336/19561
identifier_str_mv Corzo, Santiago Francisco; Ramajo, Damian Enrique; Nigro, Norberto Marcelo; 1/3D modeling of the core coolant circuit of a PHWR nuclear power plant; Elsevier; Annals Of Nuclear Energy; 83; 4-2015; 386-397
0306-4549
CONICET Digital
CONICET
dc.language.none.fl_str_mv eng
language eng
dc.relation.none.fl_str_mv info:eu-repo/semantics/altIdentifier/doi/10.1016/j.anucene.2014.12.035
dc.rights.none.fl_str_mv info:eu-repo/semantics/openAccess
https://creativecommons.org/licenses/by-nc-sa/2.5/ar/
eu_rights_str_mv openAccess
rights_invalid_str_mv https://creativecommons.org/licenses/by-nc-sa/2.5/ar/
dc.format.none.fl_str_mv application/pdf
application/pdf
dc.publisher.none.fl_str_mv Elsevier
publisher.none.fl_str_mv Elsevier
dc.source.none.fl_str_mv reponame:CONICET Digital (CONICET)
instname:Consejo Nacional de Investigaciones Científicas y Técnicas
reponame_str CONICET Digital (CONICET)
collection CONICET Digital (CONICET)
instname_str Consejo Nacional de Investigaciones Científicas y Técnicas
repository.name.fl_str_mv CONICET Digital (CONICET) - Consejo Nacional de Investigaciones Científicas y Técnicas
repository.mail.fl_str_mv dasensio@conicet.gov.ar; lcarlino@conicet.gov.ar
_version_ 1846781240230805504
score 12.982451

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