Analysis of A Nuclear Material Accounting Uncertainty for ACP Spent Fuel Standards: MCNPX and ORIGEN-S Code Simulations

作者：Lee Tae hoon^*; Joo June sik; Park Kwang june; Shin Hee sung; Kim Ho dong

来源：Journal of Nuclear Science and Technology, 2008, 410-413.

• neutron coincidence counter
• safeguards
• calibration curve
• MCNPX
• ORIGEN-S

The target nuclide to be measured by using a neutron coincidence counter is Cm-244 for the PWR spent fuel since the dominant neutron-emitting source is Cm-244 for the spent fuel. Spent fuel standards (SFS) are usually used to calibrate the neutron coincidence counter for nuclear material measurements of spent fuel. To make the best selection and minimize the measurement uncertainty of neutron coincidence counter for the SFS, expected measurement errors for several kinds of burnups were obtained by using the ORIGEN-S code and the expected singles and doubles count rates were also acquired by using MCNPX code. The expected doubles count errors for I pellet of spent fuel are less than 4% for a burnup from 27 to 40 GWD/MtU. Considering everything, a higher burnup of the spent fuel seems to be better to obtain more accurate data from a nuclear material measurement by using a neutron coincidence counter than that of a lower burnup.

• 出版日期2008-6