Research on Correction Method for Uranium Interference in Passive Neutron Coincidence Counting Method for Pu Measurement in MOX Fuel

The closed cycle requires the recovery of valuable fuels such as uranium and plutonium after spent fuel reprocessing and recycling in different forms like MOX fuel. With the improvement of spent fuel reprocessing capabilities and the advancement of integrated fast reactors, the measurement of MOX fuel characteristics has become increasingly important. In the production process of MOX fuel, the mass of uranium and plutonium must be accurately measured for nuclear material accounting requirements. Passive neutron coincidence counting method can perform a non-destructive analysis of the plutonium mass in the material, reduce the influence of environmental and other geometric factors through time-correlated neutron signal analysis, and calculate the plutonium mass with the abundance information obtained by the gamma-ray spectroscopy method. Combined with the U/Pu ratio obtained by the X-ray fluorescence analysis method, the mass of uranium and plutonium in MOX fuel can be calculated. In the neutron measurement, there is a large induced fission reaction cross-section for uranium in the energy range of spontaneous fission neutrons of plutonium. The induced fission neutrons caused by ²³⁵U affect the spontaneous fission neutron signal measurement and must be corrected. Based on the method, confirmatory experiments were conducted through PuO₂ powder and U₃O₈ powder standard samples to confirm the significant impact and a geometric model of MCNP was established. The deviation of the neutron counting rate obtained by comparative analysis of the experiment and simulation was within the acceptable range. Multiplicity shift register logic programs were used to analyze the simulation results of samples with different U/Pu ratios and abundances. The R+A and A signals were obtained by calculating the number of neutron pulses within the gates of the front and rear of the long delay. Coincident count rates can be calculated by combining the real signals and simulation time. MCNP simulated a series of MOX fuel powder samples with different PuO₂ mass contents, densities, and total masses to discuss the effects of interference under different circumstances. The interference signals from uranium are identified by neutron information provided in the MCNP output file PTRAC, and the coincidence count rates before and after eliminating the interference signal were simulated to confirm the proportion of the uranium interference signal. According to the simulation results, least squares fitting was performed on a case-by-case basis and a fitting formula with a correlation coefficient better than 0.99 was obtained. Correction formulas for the uranium interference signals under different circumstances were established, and a passive neutron measurement correction method for Pu mass in MOX fuel was established, which can be applied to the measurement of different MOX fuels.