For keeping the consistency of calculation in serial and parallel Monte Carlo calculations,the random numbers used in the tracking of history are distributed by segmentation in the neutron transport simulation. The Uniformity and independence of the random number sequence was damaged when the stride was exceeded. In this paper,the energy deposition of fission production in a reactor was calculated by MCNP. In this case,the exceeding the stride does not result in wrong answers. It was consistent with MCNP Version 5 manual. In addition,the reasons for this result were also explained in the light of the principle of MCNP simulation in this work. The reasons for resulting in an underestimate of the variance were explained too.