Radiation Transport Analysis in Chalcogenide-Based Devices and a Neutron Howitzer Using MCNP

abstract: As photons, electrons, and neutrons traverse a medium, they impart their energy in ways that are analytically difficult to describe. Monte Carlo methods provide valuable insight into understanding this behavior, especially when the radiation source or environment is too complex to simplify...

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Other Authors: Bowler, Herbert (Author)
Format: Dissertation
Language:English
Published: 2014
Subjects:
Online Access:http://hdl.handle.net/2286/R.I.27441
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spelling ndltd-asu.edu-item-274412018-06-22T03:05:43Z Radiation Transport Analysis in Chalcogenide-Based Devices and a Neutron Howitzer Using MCNP abstract: As photons, electrons, and neutrons traverse a medium, they impart their energy in ways that are analytically difficult to describe. Monte Carlo methods provide valuable insight into understanding this behavior, especially when the radiation source or environment is too complex to simplify. This research investigates simulating various radiation sources using the Monte Carlo N-Particle (MCNP) transport code, characterizing their impact on various materials, and comparing the simulation results to general theory and measurements. A total of five sources were of interest: two photon sources of different incident particle energies (3.83 eV and 1.25 MeV), two electron sources also of different energies (30 keV and 100 keV), and a californium-252 (Cf-252) spontaneous fission neutron source. Lateral and vertical programmable metallization cells (PMCs) were developed by other researchers for exposure to these photon and electron sources, so simplified PMC models were implemented in MCNP to estimate the doses and fluences. Dose rates measured around the neutron source and the predicted maximum activity of activation foils exposed to the neutrons were determined using MCNP and compared to experimental results obtained from gamma-ray spectroscopy. The analytical fluence calculations for the photon and electron cases agreed with MCNP results, and differences are due to MCNP considering particle movements that hand calculations do not. Doses for the photon cases agreed between the analytical and simulated results, while the electron cases differed by a factor of up to 4.8. Physical dose rate measurements taken from the neutron source agreed with MCNP within the 10% tolerance of the measurement device. The activity results had a percent error of up to 50%, which suggests a need to further evaluate the spectroscopy setup. Dissertation/Thesis Bowler, Herbert (Author) Holbert, Keith E (Advisor) Barnaby, Hugh J (Committee member) Clark, Lawrence T (Committee member) Arizona State University (Publisher) Electrical engineering Nuclear physics Chalcogenide MCNP Monte Carlo Radiation transport eng 101 pages Masters Thesis Electrical Engineering 2014 Masters Thesis http://hdl.handle.net/2286/R.I.27441 http://rightsstatements.org/vocab/InC/1.0/ All Rights Reserved 2014
collection NDLTD
language English
format Dissertation
sources NDLTD
topic Electrical engineering
Nuclear physics
Chalcogenide
MCNP
Monte Carlo
Radiation transport
spellingShingle Electrical engineering
Nuclear physics
Chalcogenide
MCNP
Monte Carlo
Radiation transport
Radiation Transport Analysis in Chalcogenide-Based Devices and a Neutron Howitzer Using MCNP
description abstract: As photons, electrons, and neutrons traverse a medium, they impart their energy in ways that are analytically difficult to describe. Monte Carlo methods provide valuable insight into understanding this behavior, especially when the radiation source or environment is too complex to simplify. This research investigates simulating various radiation sources using the Monte Carlo N-Particle (MCNP) transport code, characterizing their impact on various materials, and comparing the simulation results to general theory and measurements. A total of five sources were of interest: two photon sources of different incident particle energies (3.83 eV and 1.25 MeV), two electron sources also of different energies (30 keV and 100 keV), and a californium-252 (Cf-252) spontaneous fission neutron source. Lateral and vertical programmable metallization cells (PMCs) were developed by other researchers for exposure to these photon and electron sources, so simplified PMC models were implemented in MCNP to estimate the doses and fluences. Dose rates measured around the neutron source and the predicted maximum activity of activation foils exposed to the neutrons were determined using MCNP and compared to experimental results obtained from gamma-ray spectroscopy. The analytical fluence calculations for the photon and electron cases agreed with MCNP results, and differences are due to MCNP considering particle movements that hand calculations do not. Doses for the photon cases agreed between the analytical and simulated results, while the electron cases differed by a factor of up to 4.8. Physical dose rate measurements taken from the neutron source agreed with MCNP within the 10% tolerance of the measurement device. The activity results had a percent error of up to 50%, which suggests a need to further evaluate the spectroscopy setup. === Dissertation/Thesis === Masters Thesis Electrical Engineering 2014
author2 Bowler, Herbert (Author)
author_facet Bowler, Herbert (Author)
title Radiation Transport Analysis in Chalcogenide-Based Devices and a Neutron Howitzer Using MCNP
title_short Radiation Transport Analysis in Chalcogenide-Based Devices and a Neutron Howitzer Using MCNP
title_full Radiation Transport Analysis in Chalcogenide-Based Devices and a Neutron Howitzer Using MCNP
title_fullStr Radiation Transport Analysis in Chalcogenide-Based Devices and a Neutron Howitzer Using MCNP
title_full_unstemmed Radiation Transport Analysis in Chalcogenide-Based Devices and a Neutron Howitzer Using MCNP
title_sort radiation transport analysis in chalcogenide-based devices and a neutron howitzer using mcnp
publishDate 2014
url http://hdl.handle.net/2286/R.I.27441
_version_ 1718700605556391936