ESTIMATION OF BURNUP IN TAIWAN RESEARCH REACTOR-FUEL PINS BY USING NONDESTRUCTIVE TECHNIQUES

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% U-235) and enriched uranium (7.0% U-235)] by the Cs-134/Cs-137 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(a)(Cs-133), which leads to a Cs-134/Cs-137 ratio that would result in a burnup value approximately 24 to 35% lower than the measured data.