Determination of the 243,246,248Cm thermal neutron induced fission cross sections

Serot, O.; Wagemans, C.; Vermote, S.; Heyse, J.; Soldner, T.; Geltenbort, P.
November 2005
AIP Conference Proceedings;2005, Vol. 798 Issue 1, p182
Academic Journal
The minor actinide waste produced in nuclear power plants contains various Cm-isotopes, and transmutation scenarios require improved fission cross section data. The available thermal neutron induced fission cross section data for 243Cm, 246Cm and 248Cm are not very accurate, so new cross section measurements have been performed at the high flux reactor of the ILL in Grenoble (France) under better experimental conditions (highly enriched samples, very intense and clean neutron beam). The measurements were performed at a neutron energy of 5.38 meV, yielding fission cross section values of (1240±28)b for 243Cm, (25±47)mb for 246Cm and (685±84)mb for 248Cm. From these results, thermal fission cross section values of (572±14)b; (12±25)mb and (316±43)mb have been deduced for 243Cm, 246Cm and 248Cm, respectively. © 2005 American Institute of Physics


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