Direct calculation methods of kinetic parameters are proposed based on the continuous energy Monte Carlo method. In the proposed methods, the effective delayed neutron fraction eff and the neutron generation time à are estimated using eigenvalue calculations. The expected number of fission neutrons in the next generation is newly applied to the proposed methods instead of the adjoint flux that has been conventionally used. The algorithms to estimate the kinetic parameters are established and incorporated into the continuous energy Monte Carlo transport calculation code MCNP-4C, which is versatile for eigenvalue calculations of nuclear reactor cores with various types of neutron energy spectrum and geometry. The proposed methods were validated since the calculated values agreed with the experimental data of eff and eff =à for the critical cores within accuracies of 4.5% and 10%, respectively.KEYWORDS: kinetic parameter, effective delayed neutron fraction eff , neutron generation time Ã, continuous energy Monte Carlo Method, expected number of fission neutrons in the next generation