Investigation of Critical Heat Flux for Plutonium-Based Mixed Oxide Advanced Fuel Bundle Design

Abstract

The critical heat flux (CHF) performance of an advanced plutonium-based mixed oxide (MOX) fuel for potential use in a pressure tube heavy water reactor (PT-HWR) has been studied experimentally at Canadian Nuclear Laboratories with an electrically heated string simulator of 43-element fuel bundles. The fuel simulator has a uniform axial power profile and a radial power profile that is representative of the plutonium-based MOX fuel. The measurements of CHF caused by liquid-film dryout were made in the MR-3 heat transfer loop facility using R-134a refrigerant as the working fluid. The test matrix included system pressures from 1.47 to 2.11 MPa, mass flow rates from 12.7 to 14.7 kg/s and inlet temperatures from 31 to 59 °C, which are representative of the water-equivalent reactor operating conditions of 9 to 12.5 MPa pressure, 13.5 to 21.3 kg/s mass flow rate and the desired inlet subcoolings. Compared to conventional natural uranium fuel, the radial power profile of a MOX fuel exhibits a steeper and less even distribution across the fuel element rings, with a relatively higher power in the outer ring. It was found that CHF values of the MOX fuel are significantly lower than those of the natural uranium fuel. Based on the experimental data, a correlation has been derived to account for the effect of radial power profile on CHF. This correlation can be used to evaluate the relative CHF values of advanced or nonconventional fuel designs with radial power profiles deviating from that of natural uranium fuel.