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Turkish Journal of Physics

MEHMET TÜRKMEN

DOI

10.3906/fiz-1609-8

Abstract

Radial temperature distribution in a fuel rod is a parabolic function. Neutronics calculations are in general performed over a volume-averaged temperature by ignoring this distribution. Such an assumption results in an uncertainty in reactor design parameters. In this study, the magnitude of this uncertainty is estimated by solving the heat equation with a temperature-dependent conductivity coefficient coupled with a reactor physics code. The effect of radial fuel temperature distribution is investigated by representing the fuel region as multiregional. Uncertainty is investigated for various Pu contents of (U-Pu)O$_{2}$ fuel, a mix of depleted U and reactor-grade Pu. The effect of Pu content on the uncertainty is studied. The PWR TMI-1 unit cell case from UAM test problems is used in the calculations. Results are obtained by using the discrete ordinate method. From the results, it was observed that uncertainties in reactor parameters ($k_{\infty }$ and Doppler coefficient) due to the use of no temperature gradient and in $k_{\infty }$ due to cross-sections decrease as Pu content increases. Moreover, it was calculated that uncertainty due to uniform temperature inside the nuclear fuel is about 7% of uncertainty due to cross-sections.

Keywords

(U-Pu)O$_{2}$ fuel, SCALE6, uncertainty analysis, temperature gradient, Doppler feedback, cross-section

First Page

113

Last Page

123

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