Uncertainty assessment of the experimental thermal-hydraulic parameters for CDTN TRIGA research reactor

Abstract

Experimental studies have been performed in the TRIGA Nuclear Reactor at Nuclear Technology Development Centre (CDTN), Brazil, to find out its thermal hydraulic parameters. Fuel to coolant heat transfer patterns must be evaluated as function of the reactor power. The heat generated by nuclear fission in the reactor core is transferred from fuel elements to the cooling system through the fuel-cladding (gap), and the cladding to coolant interfaces. As the reactor core power increases the heat transfer regime from the fuel cladding to the coolant changes. This paper presents the uncertainty analysis in the results of the thermal hydraulics experiments performed. The uncertainty analysis on thermal hydraulics parameters is determined, basically, by the uncertainty of the reactor's thermal power.