Validation of the Monte Carlo code MCNP-DSP

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Several calculations were performed to validate MCNP-DSP, which is a Monte Carlo code that calculates all the time and frequency analysis parameters associated with the 252Cf-source-driven time and frequency analysis method. The frequency analysis parameters are obtained in two ways: directly by Fourier transforming the detector responses and 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 convergence, thus verifying the way to obtain the frequency analysis parameters from the time sequences of detector pulses MCNP-DSP was verified by calculations of configurations of sources and detectors with accurately known theoretical answers. The calculated value of the ratio of spectral densities, R( ω), was shown to be independent of detection efficiency and in some cases source intensity as theoretically predicted. The neutron and gamma ray time distributions after 252Cf fission from MCNP-DSP for simple configurations of source and detectors in air and with a small sample of beryllium between the source and detector were compared with measurements. In addition, calculations were performed for measurements using the 252Cf-source-driven noise analysis technique. MCNP-DSP adequately calculated the measured low-frequency values of R( ω) from the detector responses due to neutrons and gamma rays for the unmoderated, unreflected uranium metal cylinders using the ENDF/B-IV cross-sections. These calculations also showed that the frequency dependence of the spectra obtained from the netron detector and from the gamma-ray detector responses, in this case, is the same. The ability to calculate the measured low-frequency value of R( ω) for two tightly coupled uranium metal cylinders separated by a borated plaster disk, using detectors sensitive only to gamma rays, further demonstrated that the code could accurately calculate the detector response due to gamma rays. The results of the calculations were shown to be strongly dependent on the cross-section data sets used in the analysis. The calculated frequency analysis parameters changed significantly when using different cross-section data sets, although the neutron multiplication factor k changed slightly. This demonstrated the known increased sensitivity in calculating the noise-measured parameters rather than calculating k 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 from a 252Cf-source-driven time and frequency analysis measurement and is useful for planning experiments, analyzing the results of measurements, and verifying Monte Carlo neutron and gamma-ray transport methods. MCNP-DSP can be used in the planning and analysis of measurements by the source-driven time and frequency analysis method for nuclear weapons and/or component identification, subcriticality measurements, nuclear material control and accountability, process monitoring and control, and quality assurance of fissile components, such as reactor fuel elements. In addition, it can be used to obtain the calculational bias in the calculation of the neutron multiplication factor from subcritical experiments, a quantity that is essential to the criticality safety specialist for criticality safety assessments. This general Monte Carlo model eliminates the need for the limited point kinetics models for interpreting subcriticality measurements by this method.