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dc.contributor.author
Corzo, Santiago Francisco
dc.contributor.author
Ramajo, Damian Enrique
dc.contributor.author
Nigro, Norberto Marcelo
dc.date.available
2017-07-04T21:47:13Z
dc.date.issued
2015-04
dc.identifier.citation
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
dc.identifier.issn
0306-4549
dc.identifier.uri
http://hdl.handle.net/11336/19561
dc.description.abstract
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.
dc.format
application/pdf
dc.language.iso
eng
dc.publisher
Elsevier
dc.rights
info:eu-repo/semantics/restrictedAccess
dc.rights.uri
https://creativecommons.org/licenses/by-nc-sa/2.5/ar/
dc.subject
1/3d Modeling
dc.subject
Phwr
dc.subject
Flow And Thermal Distribution
dc.subject.classification
Ingeniería Nuclear
dc.subject.classification
Ingeniería Mecánica
dc.subject.classification
INGENIERÍAS Y TECNOLOGÍAS
dc.title
1/3D modeling of the core coolant circuit of a PHWR nuclear power plant
dc.type
info:eu-repo/semantics/article
dc.type
info:ar-repo/semantics/artículo
dc.type
info:eu-repo/semantics/publishedVersion
dc.date.updated
2017-07-03T19:51:49Z
dc.journal.volume
83
dc.journal.pagination
386-397
dc.journal.pais
Países Bajos
dc.journal.ciudad
Amsterdam
dc.description.fil
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
dc.description.fil
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
dc.description.fil
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
dc.journal.title
Annals Of Nuclear Energy
dc.relation.alternativeid
info:eu-repo/semantics/altIdentifier/doi/http://dx.doi.org/10.1016/j.anucene.2014.12.035


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