Accurately quantifying the uranium in uranium yellow cake material is the key to selecting the subsequent processing technology. As an essential nondestructive testing method for uranium-containing materials, the active multiplicity method is proposed to quantify uranium by recording and analyzing 238U fission information induced by neutron sources. However, the quantitative results are biased owing to the neutron self-shielding of the uranium yellow cake material itself and differences in water content between samples.
This study aims to rapidly measure the uranium content of uranium yellow cake material using active multiplicity method and further improve of measurement accuracy.
First of all, following the comparison of the excitation effects of different neutron sources on a sample using the MCNP (Monte Carlo N-Particle Transport) program, a 241Am-Be source was selected to simulate the sample measurement process and optimized using MATLAB programming combined with MCNP. Then, the curve of multiplication factor M versus the uranium mass was obtained by simulation, and an appropriate M was selected according to the net content of the sample. Finally, the quantitative error caused by the difference between neutron absorption and water content in the process was investigated, and the double rate was corrected using the relationship between S0/Si and D0/Di and then calculated.
The simulation results of a series of samples with different masses and water contents show that a large gap is found between M and the leakage multiplication factor ML caused by neutron self-shielding. The error in uranium quantification is less than 5%; neutron self-shielding due to the change in water content affects the single, double, and triple counting rates (S/D/T). The relative error of uranium quantification can be controlled at around 10%.
This study has significance and important reference value for further research on the application of the active multiplicity method in the production and measurement of uranium yellow cake.
