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

Author

Timothy Eugene Valentine

Date of Award

5-1995

Degree Type

Dissertation

Degree Name

Doctor of Philosophy

Major

Nuclear Engineering

Major Professor

John T. Mihalczo

Committee Members

P. N. Stevens, W. F. Lawkins, R. B. Perez

Abstract

The ²⁵²;Cf-source-driven noise analysis (CSDNA) measurement method was developed by J. T. Mihalczo and V. K. Paré 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(ω). 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) in order to calculate the noise measured parameters, time analysis quantities such as autocorrelation and cross-correlation functions, and the time distribution of counts after ²⁵²;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 ²⁵²;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 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(ω)was shown to be independent of detection efficiency and in some 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(ω) from the detector responses due 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 neutron detector and from the gamma ray detector responses, in this case, are the same. The ability to calculate the measured low-frequency value of R(ω) for a 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 code was able to adequately calculate the low-frequency value of R(ω) for a uranyl nitrate solution system at various heights. However, for some solution heights, the results of the calculations differed from the measurements. The results of the calculations were shown to be strongly dependent on the cross section data used in the analysis. The calculated noise parameters changed significantly when using different cross section data sets although K_3ff; changed slightly.This demonstrated the increased sensitivity in calculating the noise-measured parameters over calculating k_eff; when validating computational methods and cross section data. Thismore 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.

Recommended Citation

Valentine, Timothy Eugene, "MCNP-DSP : a neutron and gamma ray Monte Carlo calculation of source-driven noise-measured parameters. " PhD diss., University of Tennessee, 1995.
https://trace.tennessee.edu/utk_graddiss/10250