Calculation of neutron generation time with prompt neutron density decay method

The neutron flux density plays a very important role in reactor kinetics. The Monte Carlo code MCNP is used to calculate the neutron generation time of Xi'an Pulsed Reactor. First the decay of neutron flux density is simulated with MCNP and then the neutron generation time is calculated based on the point-reactor kinetics equation. In the slightly subcritical reactor, the effect of subcriticality, count region and distribution of the neutron source on the results are studied. The calculation indicates that the subcriticality, count region and distribution of the neutron source affect the neutron generation time very little. The dominant error is that the decay of prompt neutron density cannot be precisely described with the point-reactor kinetics equation. (authors)