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
- Institución
- Consejo Nacional de Investigaciones Científicas y Técnicas
- OAI Identificador
- oai:ri.conicet.gov.ar:11336/19561
Ver los metadatos del registro completo
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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 Técnicas2025-09-29T09:34:38Zoai: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:34982025-09-29 09:34:38.489CONICET 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 |
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1844613073241899008 |
score |
13.070432 |