Abstract: The purpose of canceling the secondary neutron source for Daya Bay Nuclear Power Station was introduced. The necessity of the usability analyses for the source range detector was explained. On the condition of some important data absented, how to calculate the neutron count rates exactly by numerical method was a tough problem. Then, a method was proposed. By using this method, an approximate coefficient was introduced to amend the MCNP calculated results. Some neutron count rates of several fuel assemblies were calculated and compared with the experimental data. The analyses indicate that the method is effective and the calculated results match up to the experiment results.