Thesis (Ph.D)--Mechanical Engineering, Georgia Institute of Technology, 2010. / Committee Chair: Hertel, Nolan; Committee Member: Kulp, William David; Committee Member: Lee, Eva; Committee Member: Pagh, Richard; Committee Member: Petrovic, Bojan; Committee Member: Rahnema, Farzad; Committee Member: Smith, Eric; Committee Member: Wang, Chris. Part of the SMARTech Electronic Thesis and Dissertation Collection.

There are a variety of uses for reference neutron fields including detector response and dosimeter studies. The Georgia Institute of Technology has a 252Cf spontaneous fission source and an AmBe (α, n) source available for use in its research programs. In addition, it has iron, lead, beryllium, tantalum, heavy water, and polyethylene spheres to modify the neutron energy distributions from these neutron sources. This research characterized the neutron leakage spectra from the source inside spherical shells using a Bonner sphere spectrometer. All the neutron fields measured were also computed with a Monte Carlo code to determine the neutron fluence rate and ambient dose equivalent rate. The comparison of experimental data and calculations are used to provide further insight into the neutron spectra as modified by the spheres. The characterization of these modified sources will provide data to assist in using the resulting neutron fields in other research activities. To measure each neutron field combination, one of the two sources was placed in the center of an attenuating sphere. The neutron field was first measured at a variety of source-to-detector distances with a Bonner Sphere System. The spectrometer measurements, specifically the count rates of the different Bonner spheres, as a function of distance from the source is fitted to obtain corrections for room-scatter and air-scatter of neutrons using the Eisenhauer, Schwartz, and Johnson method. Using these corrections, the count rates free of room return is obtained at 1 m from the source and unfolded using the BUMS software to obtain the reported fluence and dose equivalent rates. These results are compared to those generated by the Monte Carlo Neutral Particle (MCNP) code. Models were made in MCNP for each of the source and moderating sphere combinations. The neutron fluence and dose rates were tallied during the MCNP simulation. The unfolded experimental data and the MCNP calculations showed good agreement for most of source-attenuating sphere combinations, thereby reinforcing the experimental results.

The goal of a network interdiction problem is to model competitive decision-making between two parties with opposing goals. The simplest interdiction problem is a bilevel model consisting of an 'adversary' and an interdictor. In this setting, the interdictor first expends resources to optimally disrupt the network operations of the adversary. The adversary subsequently optimizes in the residual interdicted network. In particular, this dissertation considers an interdiction problem in which the interdictor places radiation detectors on a transportation network in order to minimize the probability that a smuggler of nuclear material can avoid detection. A particular area of interest in stochastic network interdiction problems (SNIPs) is the application of so-called prioritized decision-making. The motivation for this framework is as follows: In many real-world settings, decisions must be made now under uncertain resource levels, e.g., interdiction budgets, available man-hours, or any other resource depending on the problem setting. Applying this idea to the stochastic network interdiction setting, the solution to the prioritized SNIP (PrSNIP) is a rank-ordered list of locations to interdict, ranked from highest to lowest importance. It is well known in the operations research literature that stochastic integer programs are among the most difficult optimization problems to solve. Even for modest levels of uncertainty, commercial integer programming solvers can have difficulty solving models such as PrSNIP. However, metaheuristic and large-scale mathematical programming algorithms are often effective in solving instances from this class of difficult optimization problems. The goal of this doctoral research is to investigate different methods for modeling and solving SNIPs (optimization) and PrSNIPs (prioritization via optimization). We develop a number of different prioritized and unprioritized models, as well as exact and heuristic algorithms for solving each problem type. The mathematical programming algorithms that we consider are based on row and column generation techniques, and our heuristic approach uses adaptive tabu search to quickly find near-optimal solutions. Finally, we develop a group of hybrid algorithms that combine various elements of both classes of algorithms. / text

In the event of a radioactive dispersal device (RDD), the assessment of the internal contamination level of victims is necessary to determine if immediate medical follow-up is necessary. Thermo Scientific's TPM-903B Portal Monitor was investigated to determine if it is a suitable first cut screening tool for internal contamination assessment of victims. A portal monitor was chosen for this study because they are readily accessible, transportable, easy to assemble, and provide whole body count rates due to the detector size. The TPM-903B was modeled in Monte Carlo N-Particles Transport Code Version 5 (MCNP). This computational model was validated against the portal monitor's response to a series of measurements made with four point sources in a polymethyl methacrylate (PMMA) slab box. Using the validated MCNP5 model and models of the MIRD male and female anthropomorphic phantoms, the response of the portal monitor was simulated for the inhalation and ingestion radionuclides from an RDD. Six representative phantoms were considered: Reference Male, Reference Female, Adipose Male, Adipose Female, Post-Menopausal Adipose Female, and 10-Year-Old Child. The biokinetics via Dose and Risk Calculation Software (DCAL) was implemented using both the inhalation and ingestion pathways to determine the radionuclide concentrations in the organs of the body which were then used to determine the count rate of the portal monitor as a function of time. Dose coefficients were employed to determine the count rate of the detector associated with specific dose limits. These count rates were then compiled into procedure sheets to be used by first responders during the triaging of victims following an RDD.

In the event of a nuclear power plant accident, fission products could be released into the atmosphere potentially affecting the health of local citizens. In order to triage the possibly large number of people impacted, a detection device is needed that can acquire data quickly and that is sensitive to internal contamination. The portal monitor TPM-903B was investigated for use in the event of a fission product release. A list of fission products released from a Pressurized Water Reactor (PWR) was generated and separated into two groups--Group 1 (gamma- and beta-emitting fission products) and Group 2 (strictly beta-emitting fission products.) Group one fission products were used in the previously validated Monte Carlo N-Particle Transport Code (MCNP) model of the portal monitor. Two MIRD anthropomorphic phantom types were implemented in the MCNP model--the Adipose Male and Child phantoms. Dose and Risk Calculation software (DCAL) provided inhalation biokinetic data that were applied to the output of the MCNP modeling to determine the radionuclide concentrations in each organ as a function of time. For each phantom type, these data were used to determine the total body counts associated with each individual gamma-emitting fission product. Corresponding adult and child dose coefficients were implemented to determine the total body counts per 250 mSv. A weighted sum of all of the isotopes involved was performed. The ratio of dose associated with gamma-emitting fission products to the total of all fission products was determined based on corresponding dose coefficients and relative abundance. This ratio was used to project the total body counts corresponding to 250mSv for the entire fission product release inhalation--including all types of radiation. The developed procedure sheets will be used by first response personnel in the event of a fission product release.

The accurate and efficient simulation of coupled neutron-photon problems is necessary for several important radiation detection applications. Examples include the detection of nuclear threats concealed in cargo containers and prompt gamma neutron activation analysis for nondestructive determination of elemental composition of unknown samples. In these applications, high-resolution gamma-ray spectrometers are used to preserve as much information as possible about the emitted photon flux, which consists of both continuum and characteristic gamma rays with discrete energies. Monte Carlo transport is the most commonly used modeling tool for this type of problem, but computational times for many problems can be prohibitive. This work explored the use of coupled Monte Carlo-deterministic methods for the simulation of neutron-induced photons for high-resolution gamma-ray spectroscopy applications. A method was developed for the implementation of coupled neutron-photon problems into RAdiation Detection Scenario Analysis Toolbox (RADSAT), a computer code that couples the complementary strengths of discrete-ordinate and Monte Carlo approaches to obtain high-resolution detector responses. Central to this work was the development of a method for generating multi-group neutron-photon cross-sections in a way that separates the discrete and continuum photon emissions so that the key signatures in neutron activation analysis (i.e., the characteristic line energies) are preserved. The mechanics of the cross-section preparation method are described and contrasted with standard neutron-gamma cross-section sets. These custom cross-sections were then applied to several benchmark problems using the method developed in this work. Multi-group results for neutron and photon flux are compared to MCNP results. Finally, calculated responses of high-resolution spectrometers were compared. The added computational efficiency of the coupled Monte Carlo-deterministic method and the positive agreement achieved in the code-to-code verification make the integration of the coupled neutron-photon method into RADSAT a promising endeavor.

The detonation of a radiological dispersion device (RDD) may result in a situation where many individuals are exposed to contamination due to the inhalation of radioactive materials. Assessments of contamination may need to be performed by emergency response personnel in order to triage the potentially exposed public. The feasibility of using readily available standard 2x2-inch sodium-iodide detectors to determine the committed effective dose to a patient following the inhalation of a radionuclide has been investigated. The 2x2-NaI(Tl) detector was modeled using the Monte Carlo simulation code, MCNP-5, and was validated via a series of experimental benchmark measurements using a polymethyl methacrylate (PMMA) slab phantom. Such validation was essential in reproducing an accurate detector response. Upon verification of the detector model, six anthropomorphic phantoms, based on the MIRD-V phantoms, were modeled with nuclides distributed to simulate inhaled contamination. The nuclides assessed included Am-241, Co-60, Cs-137, I-131, and Ir-192. Detectors were placed at four positions on the phantoms: anterior right torso, posterior right torso, anterior neck, and lateral left thigh. The detected count-rate varied with respect to detector position, and the optimal detector location was determined on the body. The triage threshold for contamination was set at an action level of 250-mSv of intake. Time dependent biokinetic modeling was employed to determine the source distribution and activity in the body as a function of post-inhalation time. The detector response was determined as a function of count-rate per becquerel of activity at initial intake. This was converted to count-rate per 250-mSv intake for triage use by first responders operating the detector to facilitate triage decisions of contamination level. A set of procedure sheets for use by first responders was compiled for each of the phantoms and nuclides investigated.