Eventos Anais de eventos
ENCIT 2016
16th Brazilian Congress of Thermal Sciences and Engineering
THERMAL-HYDRAULIC ANALYSIS CODE FOR PLATE-TYPE FUEL NUCLEAR REACTORS
Submission Author:
Duvan Castellanos , SP
Co-Authors:
PEDRO CARAJILESCOV, José Rubens Maiorino
Presenter: Duvan Castellanos
doi://10.26678/ABCM.ENCIT2016.CIT2016-0142
Abstract
Plate-type fuel assemblies is mostly associated with research and material test reactors but currently such fuels have also been considered for research and naval propulsion reactors. This research work presents a computer code model written in FORTRAN language to perform thermal-hydraulic analysis of nuclear plate-type fuel elements, named COTENP. The code solves the conservation equation for mass, momentum and energy for a sub-channel based on geometric and thermal-hydraulic conditions. It calculates minimum DNBR for the hottest channel. The code uses the chain or cascade method for two stages in order to facilitate the whole analysis. In the first stage, we divide the core into channels with size equivalent to a fuel assembly. In the second stage, we divide the hottest fuel assembly into sub-channels with size equivalent to one actual coolant channel. For the code validation, we considered two different problems. The first was the CARR research reactor with low pressure and temperature conditions, and the second was the LABGENE reactor, a small PWR prototype with high temperature and pressure conditions. The code yields detailed information such as static pressure in the channel, mass flow rate distribution among the channels, coolant temperature axial distribution, quality and local and critical fluxes. The COTENP code reproduced well the CARR reactor results, but presented important discrepancy regarding the temperature axial distribution for the LABGENE reactor results. The DNBR estimation for both problems were accurate.
Keywords
Thermal-hydraulic, Reactor, Research, Plate-type fuel, naval propulsion
