MCNP-DSP: A neutron and gamma ray Monte Carlo calculation of source-driven noise-measured parameters

The {dollar}sp{lcub}252{rcub}{dollar}Cf-source-driven noise analysis (CSDNA) measurement method was developed by J.T. Mihalczo and V.K. Pare to determine the subcriticality of a fissile assembly. The technique provides measured parameters that can be used for verification of calculation models and cross section data. A certain ratio of spectral densities, R({dollar}omega{dollar}), obtained from the measured spectra is independent of detection efficiency and in some cases source intensity. MCNP-DSP was developed from the Monte Carlo code MCNP4A (MCNP is a trademark of the Regents of the University of California, Los Alamos National Laboratory) to calculate the frequency analysis parameters, time analysis quantities such as autocorrelation and cross-correlation functions, and the time distribution of counts after {dollar}sp{lcub}252{rcub}{dollar}Cf fission for both neutrons and gamma rays.; Several calculations were performed to validate MCNP-DSP. MCNP-DSP successfully calculated the neutron and photon time distributions after {dollar}sp{lcub}252{rcub}{dollar}Cf fission for simple configurations. The data processing algorithms were validated by successfully calculating problems with known analytical solutions. In the calculations, the frequency spectra can be obtained directly by Fourier transforming the detector response or indirectly by taking the Fourier transform of the autocorrelation and cross-correlation functions. The direct and indirect Fourier processing methods were shown to produce the same frequency spectra, and the convergence of the estimates of the frequency spectra are the same. The calculated value of R({dollar}omega{dollar}) was shown to be independent of detection efficiency and in some cases source intensity.; Several calculations were performed for systems that have been measured using the CSDNA technique. MCNP-DSP was able to adequately calculate the low-frequency value of R({dollar}omega{dollar}) from the detector responses due to neutrons and gamma rays for various uranium systems. The results of the calculations were shown to be strongly dependent on the cross section data sets used in the analysis. The calculated noise parameters changed significantly when using different cross section data sets although k{dollar}sb{lcub}rm eff{rcub}{dollar} changed slightly. This demonstrated the increased sensitivity in calculating the noise-measured parameters over calculating k{dollar}sb{lcub}rm eff{rcub}{dollar} when validating computational methods and cross section data. This more general neutron and gamma ray treatment provides the most comprehensive calculation of the measured parameters and is useful for planning experiments and analyzing the results of measurements.