Active Test of the ACP Safeguards Neutron Counter Using Spent Fuel Rod Cuts

A neutron coincidence counter for the nuclear material measurement and control of the Advanced spent fuel Conditioning Process (ACP) has been developed by the Korea Atomic Energy Research Institute (KAERI) since 2002. The most dominant neutron source among the spontaneous fission nuclides contained in spent fuel is 244Cm. To obtain the neutron counting rates of the singles, doubles, and triples coincidences of the neutron counter with an increment of the 244Cm mass, an active test of the neutron counter was performed with several spent fuel rod-cuts in the ACP hot cell in 2007. The source term of the spent fuel rod-cuts was obtained by using the ORIGEN-ARP burnup simulation code and a set of preliminary calibration curves of the neutron counter for 244Cm was generated. The calibration curves were also obtained from the results of an MCNPX code simulation, but there were some large differences in the neutron counting rates between the measurements and the MCNPX code simulation results. These differences seem to originate from the errors in the 244Cm mass obtained by the ORIGEN-ARP code simulation with inexact burnup information of the spent fuel rod-cuts. These differences were corrected by using a self-multiplication correction for the measurement results for the spent fuel rod-cuts. With those corrected counting rates, the corrected 244Cm mass was also obtained by using the point model equations of the singles and doubles rates for a non-multiplying sample. A series of burnup simulations for spent fuel rod-cuts were performed with a variation of the average burnup of a spent fuel assembly by using the ORIGEN-ARP code, and an expected burnup for each rod-cut was inferred from the simulation results.