We performed measurements of neutron elastic scattering cross sections of carbon, silicon, iron, zirconium and lead at 75 MeV using a 7 Li(p,n) quasi-monoenergetic neutron source at TIARA of JAERI. Neutron spectra at 25 laboratory angles between 2.6° and 53.0° were measured by a time-of-flight method (TOF) using five liquid scintillation detectors. The data were corrected for the inelastic scattering neutrons and sample size effects. The experimental data were compared with the neutron cross section libraries (DLCI I 9/HIL086, LA 150) and systematics used in cascade/transport codes (HETC-KFA2, NMTC/JAERI). The DLC 119 data and sytematics show large discrepancy from the present data, while the LA 150 data are in fair agreement.
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
New algorithms and techniques are developed for calculating the iterated fission probability (I FP ) using a Monte Carlo eigenvalue calculation scheme. The proportionality of the calculated I FP to the adjoint flux is confirmed numerically. This I FP is then used with the MCNP code and its point-wise cross section data libraries to calculate kinetic parameters weighted by the adjoint flux. Experimental data for critical cores validate these theoretical estimates.
A calculation technique of the inverse reactor period of a perturbed state is investigated. That is achieved by solving the natural mode equation which expresses the neutron balance where the core power changes exponentially with time. The technique is implemented into the MCNP-5 code and tested for perturbed states. The convergence of the calculated inverse reactor period is observed and the ratio of the converged values to experimental data in literature ranges 0.93~1.28. The error would be mainly attributed to estimations of the neutron balance between production and annihilation.
Double differential (n, a) reaction cross sections of 58Ni and n"tNi were measured for 4.2-6.5 MeV neutrons with energy resolution good enough to separate a-particles from the low-lying levels of residual nuclei by using a gridded ionization chamber. Angular distribution and excitation functions were derived for ao, a1 and a , ?~ components (a-particles to the ground level, the 1st level and levels higher than the 2nd level, respectively). The experimental results were compared with these obtained from calculation based on HauserFeshbach model employing the optical potential and the level density parameters derived to reproduce the experimental values of total, (n, p ) and (n, a) cross sections. The calculation showed fair agreement with the experimental data while it underestimated the (n, a) cross section above 6MeV.
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