Year
2009
Abstract
A passive neutron coincidence counter for the nuclear material measurement of the Advanced Spent Fuel Conditioning Process (ACP) has been developed by the Korea Atomic Energy Research Institute (KAERI) since 2003 and was deployed in a hot cell of the ACP Facility (ACPF) in 2005. 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, a hot test of the neutron counter was performed with several spent fuel rod-cuts in the ACPF 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 was a wide difference of around 30% in the slope of doubles rate calibration curve between the measurements and the MCNPX results. The chemical analysis results of the spent fuel samples were obtained in September 2008. Before having the chemical analysis results, the difference was considered coming from the uncertainty of source term calculated by using the ORIGEN-ARP code. However it is found that the difference between the measurements and MCNPX results is due to the error in the declared burnup since the chemical analysis burnups of the samples differ from the declared ones by around 10%. The expected burnup of each rod-cut was also obtained from the results of self-multiplication correction for the 244Cm mass of the rod-cuts and the difference between the expected burnup results and the chemical analysis results is less than 2%. This study shows the high performance of the neutron coincidence counter for 244 Cm measurements of spent fuel and also shows that the burnup of spent fuel samples can be obtained through a series of ORIGEN-ARP code simulations if it is possible to acquire the measurement data of neutron counting rates for 244Cm of the samples.