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Abstract
Thermal-hydraulic analysis of plate-type fuel has great importance to the establishment of safety criteria, also to the licensing of the future nuclear reactor with the objective of propelling the Brazilian nuclear submarine. In this work, an analysis of a single plate-type fuel surrounding by two water channels was performed using the RELAP5 thermalhydraulic code. To realize the simulations, a plate-type fuel with the meat of uranium dioxide sandwiched between two Zircaloy-4 plates was proposed. A partial loss of flow accident was simulated to show the behavior of the model under this type of accident. The results show that the critical heat flux was detected in the central region along the axial direction of the plate when the right water channel was blocked.
Keywords
RELAP5, thermal-hydraulic, critical heat flux, partial loss of flow accident, plate-type fuel
Date received: 20 July 2015; accepted: 30 November 2015
Academic Editor: Bo Yu
(ProQuest: ... denotes formulae omitted.)
Introduction
Currently, plate-type fuel is widely used in research reactors (RRs) as well as in the core of nuclear submarines (SSN).1-4 An analysis to assess the trade-offs involved in the use of highly enriched uranium (HEU) versus low enriched uranium (LEU) as SSN reactor fuel, regarding to factors as core life, core size, and reactor safety was conducted by Ippolito Jr.1 It was accomplished by modeling one HEU and two LEU reactor cores for comparison. Besides, in this study, it was determined that the plate-type fuel with the fuel device made of uranium dioxide (UO2) sandwiched into two Zircaloy-4 plates is more qualified to support the high degree of requirements to power SSNs.
According to Duderstadt,5 a more critical limitation is the heat flux that can be transferred from the clad to the coolant in water-cooled reactors, for example, pressurized water reactor (PWR) or boiling water reactor (BWR). This thermal limitation, known as critical heat flux (CHF), is of primary concern in water-cooled reactors cores in which the coolant temperature is allowed to approach the boiling point. The CHF has been the subject of research in the field of boiling heat transfer by nuclear engineers for many decades.6
As evidenced by Collier and Thome,7 the critical heat flux condition (CHFC) is characterized by a sharp reduction of the local heat...