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Conference Article: Progress in the measurement of fuel temperature coefficients in CAGRs

Abstract

The fuel temperature coefficient of reactivity in the Advanced Gas Cooled Reactor varies significantly with fuel burnup. The initial negative coefficient due to the doppler effect in 238U is counteracted by an increasing positive component arising from thermal neutron spectrum effects on 239Pu fission rate. The core-average net coefficient remains negative but is small and its prediction is subject to relatively large uncertainty. A means of measurement of the fuel temperature reactivity feedback was therefore devised which could be applied with minimum disturbance to full power operation. The method has now been applied to all the CEGB's Commercial AGR's. It involves a relatively small withdrawal of a bank of rods, a hold period of approximately one minute and reinsertion to the original level. The resulting power transient is measured by neutron flux detectors outside the core and is used as input to two calculations. Firstly, an inverse neutron kinetics programme to calculate the core reactivity as a function of time during the whole transient, and secondly, a thermal hydraulics calculation to calculate the transient in mean fuel temperature. The resulting core reactivity versus fuel temperature plot gives a measure of fuel temperature coefficient of reactivity at the times when rods are stationary. Analysis of the results to date show the method to be robust, giving repeatable and reliable results

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Reference:

Specialists' meeting on safety and accident analysis for gas-cooled reactors. Oak Ridge, TN (USA). 13-15 May 1985

International Atomic Energy Agency, Vienna (Austria)

IAEA-TECDOC--358, pp:325-336