Abstract: MOX (mixed oxide) fuel plays a positive role in promoting sustainable development in nuclear energy. The quantitative determination of plutonium in mixed MOX fuel is also essential to nuclear security and safeguards. In this regard, neutron multiplicity measurement methods serve as crucial non-destructive testing techniques and play a vital role in quantitatively detecting plutonium in MOX fuel. In order to study the factors that influence the quantification process of plutonium, enhance measurement accuracy, and streamline the process, this study develops a comprehensive and specific simulation system for the multiplicity-based plutonium quantification method in MOX fuel. Based on the AWCC device as the basic model (simulation model with a detection efficiency of 23.89% and a die-away time of 45.42 μs), the combination of MCNP and MATLAB software is utilized to recombine the detector capture times of fission neutrons and (α, n) neutrons obtained from MCNP software with the particle emission time series obtained by MATLAB sampling, forming a complete pulse time series. After receiving the multiplicity moments of fission neutrons in MOX fuel (v_s1=2.157, v_s2=3.808, v_s3=5.283, v_i1=2.855, v_i2=6.953, v_i3=13.899), simulated pulse sequence acquisition, multiplicity analysis, and quantitative calculation are performed on four samples with different shapes and sizes. Each sample is measured for 300 s with a pre-delay time of 3 μs, a gate width of 54 μs, a long delay of 2 ms, and repeated three time’s measurements. Additionally, the simulated quantitative calculation results ²⁴⁰Pu are compared with the set values. The results indicate that the mass of ²⁴⁰Pu and the α-value obtained through simulated pulse analysis meet the requirement of less than 5% relative error compared to the actual input values. This work provides technical support for data interpretation in the quantitative measurement of plutonium in MOX fuel.