Avaliação do comportamento térmico e de segurança de um reator de alta temperatura

Abstract

Fourth-generation reactors (GEN-IV) operate at much higher temperatures than third-generation reactors and are candidates for the next generation of nuclear reactors, according to the IAEA. Predicting the thermo-hydraulic performance of high temperature reactors is an important contribution to the development of technology. Therefore, special attention is paid to the behavior of materials and heat transfer in these reactors. In this way, the investigation and evaluation of the operational and safety aspects of the GEN-IV reactors have been the object of numerous studies by the international community and also in Brazil. In this work, a methodology is proposed for the thermo-hydraulic study of prismatic VHTR (Very High Temperature Reactor) reactors from thermo-hydraulic modeling through parametric studies, changing the turbulence model, the generation profile of energy in the fuel blocks and the influence of changes in the geometry itself, both in stationary operation and in transient situations. Two analyzes were performed, one for the complete core using the RELAP5-3D thermo-hydraulic analysis code to evaluate global parameters and another for a part of the reactor core for a more detailed analysis of local parameters with computational fluid dynamics tools (CFD). In the methodology of the complete core of the VHTR reactor, modeling and simulations with the code RELAP5-3D are presented, as well as their verifications for stationary calculations, considering cores of high temperature reactors cooled with helium and liquid salt, and for transient calculations of flow loss considering the effect of the gap (interstitial spaces between fuel elements). On the other hand, the methodology for the three-dimensional analysis with CFD tools was divided into 4 parts: (1) thermo-hydraulic analysis of a single cooling channel; (2) analysis of some aspects of heat transfer in the fuel element consisting of a unit cell; (3) evaluation of a baseline method for the phenomenon of flow and heat transfer of a sector equivalent to 1/12 of a standard fuel block column and (4) evaluation of a method to investigate bypass flow in 1 /12 of a standard fuel block column and especially how it is affected by various parameters. The ability to predict the main thermo-hydraulic parameters from different computational models with the complete core methodology under steady-state and transient conditions for the selected reactor was proven. The analyzes of the main thermo-hydraulic parameters: temperature of the fuel elements, of the coolant, of the structural elements, speeds and pressures were carried out with the CFD methodology from comparative studies with semi-empirical correlations. The prediction of the thermo-hydraulic parameters of the computational models of the hot channel in 2D, 3D and of the unit cell, present a lower use of computational resources, and together with 1/12 of the section of the prismatic block, it allows to obtain an acceptable description of the thermo-hydraulics of the VHTR reactor. In the study, the temperature in compact fuels always remains below the reactor design limits of 1250°C in normal operation and without reaching the limit temperatures (1600°C). The main result is that the methodologies presented in this work can be adopted to simulate any type of high temperature prismatic nuclear reactor.