Estimating spent fuel burnup with Neutron measurements: A Practical Rule of Thumb Equation

Abstract

We present a concise equation correlating the burnup of spent nuclear fuel (SF) with the neutron count rate, developed through comprehensive data analysis from Origen-ARP and MCNP codes. This equation is applicable to Fork detectors, commonly used for verifying SF assemblies before their transfer to new storage sites. The detector is assumed to be a helium-3 detector. The reaction rate (RR) in the 3He detector is influenced by Total Neutron Source Intensity (TNSI), net neutron multiplication, and neutron capture during transit to the detector. TNSI emerges as the most influential factor. Two scenarios were explored: one involving pure water and the other with water containing 2000 ppm of boron. The characteristics of the concise equation are also analyzed.

Keywords

• Spent nuclear fuel
• Burnup
• Origen-ARP
• MCNP
• Total neutron source intensity
• Fork detector