Accession Number:

ADA238738

Title:

High Altitude Neutral Particle Transport Using the Monte Carlo Simulation Code MCNP with Variable Density Atmosphere

Descriptive Note:

Master's thesis

Corporate Author:

AIR FORCE INST OF TECH WRIGHT-PATTERSONAFB OH SCHOOL OF ENGINEERING

Personal Author(s):

Report Date:

1991-03-01

Pagination or Media Count:

138.0

Abstract:

The Monte Carlo transport code, MCNP was modified for purposes of two dimensional neutron-photon transport modelling in a variable density atmosphere. Calculations were performed using cylindrical r, z geometry and a point isotropic neutron-photon source at an elevation of 40 kilometers. Neutron and photon fluence results were computed using ring detectors for field points at constant range versus source elevation angle and for field points at constant altitude versus radius. Comparisons between the modified and unmodified versions of MCNP showed a decrease in run time by a factor of two was possible. Only a fraction of the spatial cells previously required were used. Results showed streaming of the neutron and photon radiation at very close ranges. Buildup from multiple scattering contributions proved to be the dominant effect at intermediate ranges. Attenuation from successive downscatter in energy was shown to be the dominant effect at ranges greater than about 60 kilometers. Results obtained with MCNP were compared to those obtained with the computer code SMAUG- II, which uses the mass-integral scaling approximation. The comparison showed SMAUG-II to be accurate at points close to the source, but serious errors were encountered at ranges exceeding 9 kilometers as the mass range increased.

Subject Categories:

  • Radioactivity, Radioactive Wastes and Fission Products

Distribution Statement:

APPROVED FOR PUBLIC RELEASE