CANDU Fuel Attribution Through the Analysis of Delayed Neutron Temporal Behaviour
36th Annual CNS-CNA Student Conference - 2012 June 12

Presented at:
36th Annual CNS-CNA Student Conference
2012 June 12
Saskatoon, Canada
Session Title:
CNS/CNA Student Conference 2012

Madison Sellers (Royal Military College of Canada )
Emily Corcoran (Royal Military College of Canada )
David Kelly (Royal Miltiary College of Canada )


Delayed Neutron Counting (DNC) is an established technique in the Canadian nuclear industry as it is used for the detection of defective fuel in several CANDU reactors and the assay of uranium in geological samples. This paper describes the possible expansion of DNC to the discipline of nuclear forensics analysis. The temporal behaviour of experimentally measured delayed neutron spectra were used to determine the relative contributions of 233U and 235U to the overall fissile content present in mixtures with average absolute errors of ±4 %. The characterization of fissile content in current and proposed CANDU fuels (natural UO2, thoria and mixed oxide (MOX) based) by DNC analysis is evaluated through Monte Carlo simulations.

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