CFD MODELING OF THE MODERATOR TANK OF A PHWR NUCLEAR POWER PLANT

RAMAJO DAMIAN; SANTIAGO CORZO; NICOLAS SCHILIUK; ALEJANDRO LAZZARTE

San Carlos de Bariloche

Congreso; Encuentro Nacional de Investigadores de Elementos Finitos; 2014

Asociación Argentina de Mecánica Computacional

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, the control and safety rod guide tubes and the lances corresponding to the second shutdown system. Two heat sources were taken into account: the conduction/convection coming from the coolant channels? fluid and the heating caused by thermal neutron moderation. For the former, suitable boundary conditions at the CC tube walls considering 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 / implemented. After that, the effective conduction coefficient was estimated by means of fitting the overall expected transfer power. For the latter, a suitable function assuming a neutron flux distribution (in the axial and radial directions) as a heat flux or power density was implemented.Simulations allowed a thorough understanding of the complex flow patterns and the heat transfer while acquiring useful information about the temperature distribution in the moderator. These results could prove of prime importance when defining more accurate boundary conditions for modeling the in-channel flow with our in-house 1/3D multidimensional model of the reactor pressure vessel (RPV). It can also be the starting point for unsteady simulations in the moderator tank like as reactor Safety Control Rod Axe Man (SCRAM), the heat removal during a primary pump shut down or the boron distribution during a fast shut down injection.