Estimation of Burnup in Taiwan Research Reactor Fuel Pins by Using Nondestructive Techniques

Abstract

A nondestructive measurement of spent fuel pins from the Taiwan Research Reactor has been performed at the Institute of Nuclear Energy Research. The analysis is based on a simplified balance equation for integrated flux and a series of one-group burnup-dependent microscopic cross-section libraries. A semiempirical test is used for evaluating the burnup values of two different kinds of spent fuel pins [natural uranium (0.7% [sup 235]U) and enriched uranium (7.0 % [sup 235]U)] by the [sup 134]Cs/ [sup 137]Cs activity ratio. Results are compared with radio-chemical burnup measurements. The agreement is within 3.8%, which verifies the accuracy of this method. The results are also compared with a theoretical estimation by the ORIGEN-II code. This indicates that the ORIGEN-II code's library might have an overestimated [sigma][sub a]([sup 133]Cs), which leads to a [sup 134]Cs/[sup 137]Cs ratio that would result in a burnup value [approximately]24 to 35% lower than the measured data.