Development of NDA Measurement Method to Determine the Fissile Material Contents for DUPIC Fuel

Abstract

Neutron measurement method by NDA is being developed and simulated for the possible use in determination of fissile contents of DUPIC (Direct Use of Spent PWR Fuel in CANDU) fuel. This method could effectively be applied to DUPIC fuel design, fabrication and its safeguard implementation under the condition of very high radiation environment. The change of neutron count between the induced and non-induced fission by Cm-244 spontaneous fission neutron in spent fuel was analyzed. Results from MCNP calculation model for the two-parameter(singles and doubles) are compared with NDA measurements using PWR spent fuel rod-cuts at KAERI. It shows that the measured neutron count ratio versus quantity of spent fuel material is reasonably well agreed with the calculated values.