Simulação computacional de eventos termo-hidraulicos transitorios em multicircuitos com multibombas

AUTOR(ES)
DATA DE PUBLICAÇÃO

2003

RESUMO

PANTERA-2 (from Programa para Análise Termo-hidráulica de Reatores a ÁguaProgram for Thermal-hydraulic Analysis of Water Reactors, Version 2), whose fundamentals are described in this work, is intended to carry out rod bundle subchannel analysis in conjunction with multiloop simulation. It solves simultaneously the conservation equations of mass, axial and lateral momentum, and energy for subchannel geometry coupled with the balance equations that describe the fluid flows in any number of coolant loops connected to A pressure vessel containing the rod bundle. As far as subchannel analysis is concemed, the basic computational strategy of P ANTERA-2 comes from COBRA codes, but an altemative implicit solution method oriented to the pressure field has been used to solve the finitedifference approximations for the balance laws. The results provided by the subchannel mode1 comprise the fluid density, enthalpy, flow rate, and pressure fields in the subchannels. The loop model predicts the individualloop flows, total flow through the pressure vessel, and pump rotational speeds as a function of time subsequent to the failure of any number of the coolant pumps. The flow transients in the loops may initiated by partial, total or sequentialloss of electric power to the operating pumps. Transient events caused by either shaft break or rotor locking may also be simulated. The changes in rotational speed of the pumps as a function of time are determined from a torque balance. Pump dynamic head and hydraulic torque are calculated as a function of rotational speed and volumetric flow from two polar homologous curves supplied to the code in the tabular form In order to illustrate the analytical capability of P ANTERA-2, three sample problems are presented and discussed. Comparisons between calculated and measured results indicate that the program reproduces with a good accuracy experimental data for subchannel exit temperatures and critical heat fluxes in 5x5 rod bundles. It is also observed a good correspondence between the theoretical curves predicted by P ANTERA-2 and measured values for pump rotational speeds and mass flow rates in the primary loops of Angra-2 nuclear power plant, when the four main coolant pumps are simultaneously switched off to simulate the flow decline evento

ASSUNTO(S)

escoamento bifasico subchannel analysis flow loop analysis reatores nucleares - dinamica dos fluidos subchannel codes pump failure accident. cobra codes

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