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Hot-Test Results of the Advanced Spent Fuel Conditioning Process Safeguards Neutron Counter for PWR Spent Fuel Rods

Tae-Hoon Lee, Young Soo Kim, Hee-Sung Shin, Ho-Dong Kim

Nuclear Technology / Volume 176 / Number 1 / October 2011 / Pages 147-154

Radiation Measurements and General Instrumentation

A passive neutron coincidence counter for 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 in 2007 with several spent fuel rod-cuts in the ACPF hot cell. The source term of the spent fuel rod-cuts was obtained 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 [approximately]30% in the slope of the double-rate calibration curve between the measurements and the MCNPX results. Chemical analysis results of the spent fuel samples were obtained in September 2008, and it was found that the difference between the measurements and the MCNPX results is due to an error in the declared burnup since the chemical analysis burnups of the samples differ from the declared ones by [approximately]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 <2%. This study shows high performance of the neutron coincidence counter for 244Cm 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.

 
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