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CFD Modeling of the Moderator Tank of a PHWR Nuclear Power Plant

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Etiquetas

  • centrales nucleares
  • arn argentina
  • termohidraulica
  • instalaciones de agua pesada
  • modelizacion
  • hidraulica termica
  • moderadores

Datos

Origen

Argentina

Idioma

Inglés

Organización

ARN (Argentina)

Autor

Ramajo, D.; Corzo, S.; Schiliuk, N.; Lazarte, A.I.; Nigro, N

Descripción

A steady state CFD simulation of the moderator tank of Atucha II Nuclear Power

Plant (a Pressurized Heavy Water Reactor PHWR) was performed. Three-dimensional (3D)

detailed modeling of the tank was achieved including inlet and outlet ring-shaped

distributors, the coolant channel (CC) tubes and the control and safety rod guide tubes. Two

heat sources were taken into account: the conduction/convection from the coolant channels’

fluid and the heat transfer by thermal neutron moderation. For the former, suitable

boundary conditions (wall temperature) at the CC tube walls were arisen from 2D

estimation of the conduction/convection heat through the coolant walls. The coolant

temperature profile along each CC (obtained from a previous 1/3D model) along with

correlations for the in-channel side convective coefficient were considered. The effective

conduction coefficient was estimated by fitting the overall expected transfer power. For the

latter, a homogeneoussource was implemented.

Simulations allowed a thorough understanding of the complex flow and the heat transfer

phenomena, while acquiring useful information about the temperature distribution in the

moderator. The most relevant conclusion is that the power transferred from the CCs to the

moderator does not show linear dependence on the fission power but on the coolant

temperature, which is very similar for all CCs. These results become of prime importance

when defining more accurate boundary conditions for modeling the in-channel flow with a

previously developed, in-house, 1/3D multidimensional model of the reactor pressure

vessel (RPV). The 3D model developed is the starting point to carry on unsteady

simulations in the moderator tank,such as reactor Safety Control Rod Axe Man (SCRAM),

heat removal during a primary pump shut down or boron distribution during a fast shut

down injection. XXI Congreso sobre Métodos Numéricos y sus Aplicaciones- ENEIF 2014. Bariloche, Argentina, 23 al 26 de septiembre de 2014