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Estimation of burnup in Taiwan research reactor fuel pins by using nondestructive techniques

Journal Article · · Nuclear Technology; (United States)
OSTI ID:6480731
 [1];  [2]
  1. Chung Cheng Inst. of Tech., Taoyuan (Taiwan, Province of China)
  2. Inst. of Nuclear Energy Research, Taoyuan (Taiwan, Province of China)
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.
OSTI ID:
6480731
Journal Information:
Nuclear Technology; (United States), Journal Name: Nuclear Technology; (United States) Vol. 102:3; ISSN 0029-5450; ISSN NUTYBB
Country of Publication:
United States
Language:
English