Abstract: When the fission reaction happens in neutron transport calculation,the energy of the fission neutron can be usually determined by the fission spectrum.Since the coefficient of the spectrum type depends on the incident neutron energy,the multigroup neutron fission spectrum should strictly be a matrix form.In general transport calculation,the ²³⁵U vector fission spectrum is usually chosen as the standard fission spectrum.In order to make clear what kind of effect will be produced after this treatment,the Monte Carlo method and the multigroup P₃ Monte Carlo neutron transport code MCMG are used to analyze the difference between two types of fission spectrums.By comparison the results with MCNP code,it is certain that the difference exists,but it does not have any effect on the correction of the calculated results.At the same time,the calculation results of the different neutron cross section libraries have been compared.

Key words: matrix fission spectrum, vector fission spectrum, critical calculation, Monte Carlo