Abstract: The ex-core nuclear instrument equipment consists of the source range detector, the start-up and work range detector and the refueling monitoring range detector in Tianwan nuclear power plant (NPP). The neutron flux density of the reactor is monitored by the refueling monitoring range detector, which is a temporary device，installed into the measurement tubes before refueling and dismounted after refueling manually. It will increase radiation risk to worker, prolong overhaul period and have an impact on the economic benefits due to installing and dismounting the refueling monitoring range detector. To cope with the deficiency of neutron flux density monitoring method during refueling in Tianwan NPP, an optimization scheme was investigated to replace the refueling monitoring range detector with the source range detector to complete the neutron flux density monitoring during refueling. How to calculate source range detector neutron counting rate exactly was critical to this scheme. Firstly, the neutron emission rate versus burnup of spent fuel assemblies with initial enrichment of 2.4%, 3.6%, 4.0% and 4.9% were calculated using the ORIGENS program, mainly considering the spontaneous fission neutron and (α, n) neutron. Secondly, based on the actual core structure，the axial distribution of the neutron emission rate and the neutron emission spectrum of the spent fuel assembly calculated by the ORIGENS program，a Monte Carlo program MCNP was used to simulate the calculation of the neutron flux density where the source range detector was located. Finally, the neutron counting rates of the source range detector were calculated with considering about the deviation coefficient K and the conservative factor B and compared with the experimental data. In this paper，the calculation of the neutron counting rates of the source range detector were performed according to the two core refueling methods of all-in and all-out and core feeding. The calculation results show that: 1) The neutron counting rate of the 6th channel source range detector is affected mostly by the spent fuel assembly located at (15, 24); 2) The neutron counting rate of the source range detector is 0.5 s^-1 when the burnup of the spent fuel assembly near the source range detector reachs a certain one, and which is in good agreement with the measured value. According to the fuel management strategy of Tianwan NPP unit 3, the burnup of the spent fuel assembly near the source range detector will reach 44 000 MW·d·tU^-1 since the third cycle, which will ensure the neutron counting rate meets the monitoring requirement during refueling. It shows that the method of the source range detector can be used for the monitoring of the neutron flux density replacing the monitoring range detector during refueling.