Session 32: Thermohydraulic of Nuclear Reactors

Chairs:

Sérgio Viçosa Möller
PROMEC-UFRGS

Comissão Organizadora e Editorial do ENCIT 2000
Departamento de Endgenharia Mecânica
Universidade Federal do Rio Grande do Sul




s32p02

SIMULATION OF TWO-PHASE FLOW FOR COMPRESSIBLE AND NEARLY INCOMPRESSIBLE REGIMES

Maria de Lourdes Moreira - [email protected]
Paulo Augusto Berquó de Sampaio - [email protected]
Instituto de Engenharia Nuclear - CNEN
Cx.P. 68550 - 21945-970, Rio de Janeiro, RJ, Brasil

A new finite element formulation for compressible and nearly incompressible problems is applied to the analysis of two-phase flows.The method allows the analysis of nearly incompressible flows without resorting to the incompressible fluid model. Thus, the fluid equation of state is not discarded from the model. This feature is of foremost importance for the approach adopted in this work, where the thermodynamic behaviour of the mixture is constructed from the thermodynamic behaviour of the individual phases. Numerical examples of two-phase steam-water mixtures are presented. The models employed to characterise turbulence and the two-phase mixture are rather simple. Nevertheless, the examples presented show good qualitative behaviour and serve to indicate further developments towards the computational simulation of two-phase flows.

Keywords: Computational Fluid Dynamics, Two-Phase Flow, Finite Elements
 
 



s32p04

SIMULATION OF NONLINEAR DYNAMICS OF A PWR CORE BY AN IMPROVED LUMPED FORMULATION FOR FUEL HEAT TRANSFER

Jian Su - [email protected]
Nuclear Engineering Department, COPPE/UFRJ, CP 68509, Rio de Janeiro, 21945-970, Brazil
Renato M. Cotta - [email protected]
Mechanical Engineering Department, COPPE/UFRJ
CP 68503, Rio de Janeiro, 21945-970, Brazil

Abstract.  In this work, thermohydraulic behaviour of PWR, during reactivity insertion and partial loss-of-flow, is simulated by using a simplified mathematical model of reactor core and primary coolant.  An improved lumped parameter formulation for transient heat conduction in fuel rod is used for core heat transfer modelling.  Transient temperature response of fuel, cladding and coolant is analyzed.

Keywords: Reactor thermohydraulics, Transient heat conduction, Lumped parameter analysis, Point kinetics, Reactor neutronics