Calculating neutron flux in a finite medium using the SN method often yields values between 0 and 1, which may indicate normalization rather than actual flux. The real flux can be determined by multiplying the normalized values by the local or average power level, which can range from 1 W to 1 GW in a critical fission system. The actual neutron distribution is influenced by fission density and the relative fission cross-section compared to absorption cross-sections. Additionally, temperature effects, such as Doppler broadening and moderator density, must be considered for accurate calculations. Understanding these factors is essential for obtaining the real neutron flux in steady-state conditions.
