A characterization method for reprocessed spent nuclear fuel through neutron and gamma-ray multiplicity counting

It is difficult to measure neutrons and gamma rays coming directly from uranium and plutonium atoms in pyroprocessed ingot, which is of great interest to nuclear safeguards. Most neutrons of the ingot are emitted from 244 Cm isotope and gamma rays from its fission products. Therefore, it is necessary to identify spent fuel characteristics such as initial enrichment, burnup and cooling time from measurement and estimate plutonium mass from calibration curve or burnup calculations. In this study, we explored the possibility of characterizing pyroprocessed ingots using neutron and gamma-ray multiplicity counting. We particularly focused on finding the possible correlation between neutron and gamma-ray multiplicity properties (singles, doubles, and triples) and spent fuel properties for U/TRU/RE preprocessed ingot. We examined neutron and gamma-ray correlation for spent fuel, using the Cascading Gamma-ray Multiplicity with Fission (CGMF) and the Fission Reaction Event Yield Algorithm (FREYA) library in MCNP6.2. Due to high spontaneous fission rate, neutron and gamma-ray multiplicity parameters did not strongly correlate with multiplication factor from the point model but was proportional to 244 Cm mass. The results based on the CGMF and FREYA fission library showed the similar trends for neutron and gamma-ray cases. The sum of two multiplicity data yielded best linearity with the 244 Cm effective mass. Although total neutron counting ( S ) alone was sufficient for effectively characterizing spent fuel due to its strong linearity, this feasibility study confirmed that, in principle, it will be more effective to calculate the sum of neutron and gamma-ray multiplicity properties to directly estimate the mass of 244 Cm, rather than relying on the conventional method of calculating the multiplication factor.