Evaluation of the effect of minor actinide nuclides on neutron non-destructive assay for plutonium quantification in MA-containing fuel

Eguchi, Aya; Sagara, Hiroshi*; Mitsuboshi, Natsumi; Nagatani, Taketeru 

This research aims to develop a measurement method that enables Pu quantification under conditions that include measurement inhibitors such as MA and FP. Monte Carlo simulations of neutron transport to simulate neutron coincidence counting of MA containing MOX fuel assembly were performed in MCNP6.2 code with ENDF/B-VII.0 nuclear data library. Multiple He-3 tubes with Polyethylene moderator were modeled as a neutron detector. Fuel composition was derived from equilibrium status core of a large-scale sodium cooled fast reactor. The simulation results showed the correlation between Pu fissile content and neutron leakage multiplication, estimate Pu fissile content by DDSI method. This paper shows the results of an evaluation of the effects of MA nuclides contained in MA-containing fuel on the neutron coincidence and DDSI methods.