Mountain Scholar :: Browsing by Author "Brandl, Alexander, advisor"

Browsing by Author "Brandl, Alexander, advisor"

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Open Access
A comparison of air samples at ground level and aerial gamma count rates from the response to the Fukushima Dai-Ichi Nuclear Power Plant accident
(Colorado State University. Libraries, 2014) Sublett, Sarah Miriam, author; Brandl, Alexander, advisor; Biedron, Sandra, committee member; Johnson, Thomas, committee member
March 11, 2011 will be a day that will never be forgotten in the minds of the thousands of Japanese people whose lives were forever changed by a series of natural disasters, including a 9.0 earthquake and subsequent tsunami that triggered the Fukushima Dai-Ichi Nuclear Power Plant (FDNPP), located on the eastern coast of Japan, to become cripplingly damaged. The FDNPP nuclear accident resulted in the emission of radionuclides into the environment and those released nuclides, specifically 134Cs, 137Cs, and 131I and their measurement by ground and air based means, are the subject of this research project. Within the first few days following the start of the disaster, numerous US federal agencies responded and deployed to Japan to help characterize and measure the release of radionuclides from FDNPP. Over the course of approximately a two-month span, over 500,000 measurements were obtained and analyzed, including air and soil samples and in situ spectra. The core of this research project was to analyze and compare ground air samples to aerial gamma count rate measurements obtained in Fukushima Prefecture within the first two months following the disaster. The results of this project estimate the ground deposition of radionuclides in Fukushima Prefecture, which accounted for 99.4% of the measured aerial net gamma count rate. Another finding of this project is the estimated ground deposition of nuclides based upon aerial gamma count rates in areas where there was measurable ground air contamination was 14.25 times higher than in areas where there was no measurable ground air contamination as determined by the evaluated air samples in this project. Of the samples evaluated in this project, ground deposition averaged 5.4E6 Bq/m2 in areas where there was measurable ground air contamination, versus 3.79E5 Bq/m2 in areas where there was no measurable ground air contamination.
Open Access
Analytical methods to enhance detection of anthropogenic radionuclides in environmental matrices
(Colorado State University. Libraries, 2016) Rosenberg, Brett L., author; Brandl, Alexander, advisor; Borch, Thomas, committee member; Henry, Charles, committee member; Pinder, John, committee member; Steinhauser, Georg, committee member
The efficacy of methods that are used to detect radionuclides is dependent on the properties of the radionuclides and the matrices being analyzed. Gamma spectroscopy is an excellent tool for detecting very low quantities of a short-lived gamma-emitting radionuclide. However, as the probability of gamma ray emission decreases and the half-life increases, greater quantities of a radionuclide are required for detection by gamma spectroscopy. Since most transuranic actinides are usually not present in such quantities or concentrations in the environment, mass spectrometry is the preferred tool. For tritium, 90Sr, and other lower-Z elements that emit no easily detectable gamma rays, liquid scintillation counting is commonly used to measure the beta particles they emit. However, this methodology requires radiochemical extraction procedures to ensure a maximized ratio between signal and background. Nondestructive gamma spectroscopy was used to evaluate radiocesium content in soil and vegetation samples collected from the Fukushima prefecture exclusion zone in 2013 and 2014. Liquid scintillation counting was used for quantifying 3H in samples collected in 2013 and 90Sr in samples collected in 2013 and 2014. The radiocesium and 90Sr activities were found to have decreased from 2013 to 2014. Although 3H activities could be quantified in most samples, a sample from Chimeiji had a specific activity that statistically exceeded background (1.2 ± 1.6 Bq mL-1); further investigation is required to ascertain if 3H is present within that sample. Reports generated by TEPCO were also evaluated; radiocesium ratios and 131I/132Te ratios calculated from the reports reveal the importance of considering counting statistics and spectroscopic interference when drawing conclusions about the presence of anthropogenic radionuclides in environmental samples. Gamma spectroscopy was then applied to explore radiochemical separation techniques that can enhance detection of anthropogenic radionuclides, especially gamma-emitting actinides like 239Np shortly after a nuclear event. Ion specific extraction chromatography was found to be effective at minimizing spectroscopic interference from fission products, and addition of stable iodide carrier and a precipitating agent facilitated decreasing radioiodine activity within environmental samples. Extraction chromatography was found to reduce 131I interference by at least one order of magnitude, making it preferred for reducing 131I activity within an environmental sample. Extraction chromatography also avoids the potential of precipitating any analyte. The separation and measurement techniques utilized herein have effectively enhanced the ability to detect low-activity anthropogenic radionuclides; supplemental measurements gathered from the exclusion zone confirm the observed trends and prove the necessity of minimizing interference.
Open Access
Characteristics of gamma radiation fields in subterranean structures for radiation protection and decision making
(Colorado State University. Libraries, 2022) Parker, Alex R., author; Brandl, Alexander, advisor; Johnson, Thomas E., committee member; Lindsay, James, committee member
The threat of a CBRN attack or accident within subterranean space presents unique challenges for decision makers and emergency planners due to the operational constraints imposed by the physical environment. Radiological exposure device threat scenarios have not previously been explored for vulnerable subsurface infrastructure, like mass transit tunnels. It is important to investigate the impact that radiation exposure devices could have in these types of environments because radiation fields from gamma ray emitting sources behave peculiarly in well shielded and confined spaces where radiation scattering is substantial. This project began with benchmark measurements of a Cs-137 source in several different well shielded, small scale tunnel geometries as a proof of concept study demonstrating the complexity of radiation fields in such environments. Follow on calculations utilizing the radiation transport code, MCNP®, confirmed that the high scattering environment results in apparent radiation streaming down the length of the tunnel, where the calculated dose rates are higher than observed in a free field at equivalent distances. The tunnel material also proved important for the impact of radiation scattering, implying that there is an optimization between the probability of scattering and self-absorption in walls. Other simulated tunnel geometries confirmed the presence of scatter and increased radiation dose rates beyond the line-of-sight of the source, where virtually no transmission is expected through the tunnel materials. The final part of this project was implementing the modeling techniques into real world threat scenarios for subterranean infrastructure. Two models of a full sized underground roadway were completed using two known radiological exposure device threat materials, Cs-137 and Co-60. Both models resulted in similarly shaped radiation fields and confirmed that the near wall of a crossing roadway offered lower radiation dose rates than the far wall upon approach. These findings could prove useful to decision makers facing a subsurface infrastructure CBRN incident and could lead to the development of tools that can be implemented into emergency preparedness framework.
Open Access
Characterization of mixed linear energy transfer environments utilizing tissue-equivalent proportional counters
(Colorado State University. Libraries, 2023) Fehrman, Joseph M., author; Brandl, Alexander, advisor; Johnson, Thomas, advisor; Carlson, Kenneth, committee member
There is currently great interest in the biological impact of radiological space exposures due to manned space missions (e.g., moon) where astronauts will face the challenge of living on board spacecraft for long periods of time. Cosmic radiation of many types exists in space and creates a unique mixed linear energy transfer (LET) environment. A tissue equivalent proportional counter (TEPC) was used to produce dose-equivalent measurements and quantify neutron exposures. Three neutron sources were used to simulate high energy, mixed LET environments: californium-252, a plutonium-beryllium source, and a deuterium-tritium neutron generator. TEPC dose-equivalent measurements can be utilized for basic research, and regulatory or clinical purposes for correlation with observable health effects. The main study was to quantify and compare TEPC dose equivalent rates in microdosimetric volumes to determine if significant dosimetric differences exist between mixed LET environments generated by photons and neutrons. The findings from this experiment showed that mixed LET environments where both photons and neutrons interacted with the TEPC had lower average LET values than neutron-only exposures, and produced varying dose equivalent rates that were dependent on the source characteristics. In summary, the TEPC was capable of monitoring in a mixed-LET environment and was successful at measuring the absorbed dose of high-energy photon and neutron interactions in space-like settings.
Open Access
Characterization of radiation response for first responder instrumentation
(Colorado State University. Libraries, 2019) Fabian, Ryan, author; Brandl, Alexander, advisor; Johnson, Tom, committee member; Gudmestad, Robert, committee member
Radiation exposures of first responders can present health concerns. Some situations where such radiation exposures of first responders may exist include emergencies at nuclear power plants or facilities that house nuclear materials, coming into contact with lost or stolen nuclear material or nuclear weapons, and a dirty bomb or nuclear weapon detonation. As a result, it is important for first responders to be outfitted with equipment appropriate for each of these emergency situations. One such piece of equipment is an electronic personal dosimeter (EPD). EPDs are capable of providing real-time dose and dose rate data, and can also produce audible alarms triggered when preset dose rate levels are exceeded; however, electronic personal dosimeters do not discriminate natural background radiation contributing to their signal. This can result in the first responders becoming distracted/confused by dose or dose rate readouts from natural background radiation, or in missed real radiation exposure threats from unknown sources. An algorithm which would allow for the discrimination of natural background radiation from an EPD's signal was developed in the course of this project. If applied and implemented in a commercial EPD, this algorithm will allow first responders to distinguish directly between a readout due to naturally occurring radon progeny in the environment and contributions from other sources of ionizing radiation.
Open Access
Comparison of MCNP modeling against empirical data for the measurement of gamma fields due to actinide oxides in a glovebox
(Colorado State University. Libraries, 2012) Adams, David Wright, author; Brandl, Alexander, advisor; Johnson, Thomas E., committee member; Lindsay, James E., committee member
Los Alamos National Laboratory (LANL) is a facility that conducts research in the fields of global security, astrophysics, nuclear energy, and materials science. At Technical Area 55 (TA-55) actinide oxides used for experimental nuclear fuel research are placed in gloveboxes to be manipulated by glovebox operators. The actinide oxides are radioactive and emit gamma rays which impact the glovebox operator. Although proper precautions protect workers from unnecessary dose, the measurement and characterization of the gamma fields are useful in deciding if multiple dosimetry may be necessary as per standards at the lab and national standards. Experimental measurements were made by radiation protection personnel of TA-55 at the glovebox containing different actinide oxides. Thermoluminescent dosimeters (TLDs) were utilized to determine the dose to workers at the glovebox. A lead apron covered some of the TLDs while others were left unshielded to simulate shielded and unshielded portions of the body. Monte Carlo N-Particle Code Version 5 (MCNP5) was used to model and simulate the experimental setup at TA-55 to determine the efficacy of the lead apron and to determine the spatiality of the dose distribution as required to determine whether multiple dosimetry is necessary or not. Multiple dosimetry was found to not be required given that the TLD's are worn on the chest, at the location of highest dose.
Open Access
⁶⁴Cu-ATSM internal dosimetry in a computational tumor-bearing mouse model
(Colorado State University. Libraries, 2021) Witter, Paige, author; Brandl, Alexander, advisor; Leary, Del, committee member; Stallones, Lorann, committee member
The radiopharmaceutical Copper-64 diacetyl-bis(N4-methylthiosemicarbazones) (64Cu-ATSM) has growing clinical interest as a theragnostic agent; that is, one that can be used as both a diagnostic tool and a treatment for certain tumors. The ATSM ligand selectively distributes to hypoxic tissues, traditionally difficult targets for chemotherapy and radiotherapy cancer treatments. The copper radionuclide dissociates in the reducing environment, at which point the unique decay scheme of 64Cu offers multiple imaging and clinical pathways. Using an anatomically accurate voxelized phantom of a cancer-bearing mouse, a Monte Carlo N-Particle (MCNP) radiation transport code assessment of absorbed fractions and absorbed dose to major organs was carried out. In decay scenarios in which the organ of interest was the source organ, approximately 60% of the total energy transferred from a 64Cu decay came from the beta decay, approximately 30% from the positron decay, and approximately 1% from the Auger electrons. These results were used in conjunction with 64Cu-ATSM biodistribution data from two published studies to estimate organ and whole-body absorbed doses from administered activities. The differences in organ absorbed doses between the studies shows the mouse model variances that can impact dose and lead to variability in dose-response curves down the line. An MCNP model allows for a better understanding of the organ-specific doses and should be considered in the development of future studies assessing the theragnostic properties of 64Cu-ATSM in translational models while sparing these models' healthy organs and tissues.
Open Access
Decontaminating cobalt-60 from wounds
(Colorado State University. Libraries, 2013) Mannis, Daniel, author; Brandl, Alexander, advisor; Johnson, Thomas, committee member; Lindsay, James, committee member
Removing radionuclide contamination from wounds in tissue is essential to minimizing incorporation and dose to an individual. This experiment compared the effectiveness of decontaminating wounds inflicted in pig tissue that were contaminated with cobalt-60. The process was established to compare three decontamination methods consisting of: commercially available, non-prescription, surfactant based, non-ionic wound cleanser spray; physiologic saline solution spray; physiologic saline solution pour. Three wound types were used: smooth incision, jagged cut, and blunt force trauma wounds. The cleanser spray and the saline spray were more effective at decontaminating all three wounds than the saline pour. The difference between the cleanser spray and saline spray was not statistically significant, but the cleanser spray did decontaminate the wound to a lower mean value. The spray pressure used for the saline and cleanser sprays produced the most noticeable impact in the decontamination process.
Open Access
Design of a Compton scatter based radiation tracking system
(Colorado State University. Libraries, 2015) Healy, Heather, author; Brandl, Alexander, advisor; Johnson, Thomas, advisor; Lindsay, James, committee member
Gamma spectroscopy is one of the most common techniques used for the detection of radiologic materials. This technology is deployed in a variety of scenarios such as emergency response, monitoring, and the recovery of lost, stolen, or otherwise unaccounted radiologic material. In most practical scenarios, it is useful to know the location of a source in relation to a detector, in addition to the classic output from gamma spectrometers such as decay rate and energy peak information. In collaboration with the Remote Sensing Laboratory (RSL) at Andrew's Air Force Base, a novel detector design by RSL, which utilizes a 360° detectable range in order to increase the probability of remote detection, was investigated for the possibility to recreate source location information from Compton scattering events within the detector. A recreation of this novel detector is simulated using Geant4 to determine the optimal dimensions of sodium iodide detectors that produce the most single Compton scattering events in order to facilitate source location through the back-projection of Compton scattering angles. The optimal detector dimensions are determined by maximizing the number of single Compton scatter events and minimizing the percentage of Compton events that undergo multiple successive scatters in detectors of varying thicknesses and lengths. The optimal detector thickness was chosen to be 1.88 in, and the optimal detector length was chosen to be 4 to 4.5 in. In future projects, these optimized detectors can be used to apply suggested back-projection algorithms in order to determine the feasibility and functionality of this detector design for the purpose of radiologic source location.
Open Access
Detection and analysis of low level tritium in rainwater for a proposed environmental monitoring program
(Colorado State University. Libraries, 2014) Gillis, Jessica McDonnel, author; Brandl, Alexander, advisor; Whicker, Ward, committee member; Collett, Jeffrey L., committee member
Radioactive tritium, an isotope of hydrogen, is present at low levels in the atmosphere and can be deposited by precipitation. Tritium is produced naturally by the interaction of cosmic rays with gaseous atoms in the atmosphere, but the primary contributors to atmospheric concentrations are residues from past nuclear weapons testing and releases of tritium produced at nuclear facilities. The National Atmospheric Deposition Program (NADP) is a nationwide program that manages and analyzes rain and snow samples from networks of precipitation monitoring stations. The NADP and Savannah River National Laboratory have jointly proposed a monitoring program for tritium in rainwater in order to demonstrate the use of existing sampling locations in the NADP's National Trends Network and characterize the deposition of radionuclides in the United States. This research investigates the feasibility of measuring tritium concentrations in rainwater samples given the proposed laboratory detection range of 0.6-1.2 Bq/L. Rainwater samples were analyzed using Colorado State University's liquid scintillation counter (LSC), and minimum detectable activity concentrations on the LSC were investigated based on background count rate, count duration, and detection efficiency. To achieve the analytical capabilities and throughput proposed, count times of several hours and comparison with tritium-depleted blanks were determined to be necessary. Detection efficiencies for tritium in rainwater were affected by quench in the samples, optimization of the counting window, and LSC vial type.
Open Access
Development of a Bayesian linear regression model for the detection of a weak radiological source from gamma spectra measurements
(Colorado State University. Libraries, 2021) Meengs, Matthew, author; Brandl, Alexander, advisor; Johnson, Thomas E., committee member; Sudowe, Ralf, committee member; Kokoszka, Piotr, committee member
The detection of radiation requires the use of statistical tools due to the probabilistic nature of the emission and the interaction properties of radiation. Frequentist statistical methods are traditionally employed towards this end – the most common being the "traditional" method which calculates a decision threshold above which a source is determined to be present. The decision threshold is calculated from a predetermined false positive rate (typically 5%) and is used as a decision parameter. The decision parameter is a statistical tool by which it is determined whether or not a source other than background is present. In radiological conditions where a source is both improbable and weak, and where counting time is limited, the detection of a source becomes progressively more challenging using this traditional method. The detection of clandestine fissile materials presents such a challenge, and with the increasing risk of nuclear proliferation, there exists a growing motivation to research more optimal methods of detection, especially where a missed detection is of such high consequence. Previous research has been conducted on using a Bayesian model to develop a decision parameter for weak source detection. The use of a Bayesian model has been shown in laboratory settings to outperform the traditional frequentist method. However, the model tested was designed for gross counts only. In the present study, a Bayesian algorithm is being developed and tested that uses the entirety of the gamma spectrum. Specifically, several Bayesian linear regressions are developed and tested which compared different energy ranges in the spectrum. The parameters generated from these linear regressions are tested for their efficacy as decision parameters. With the additional information presented from the entire spectrum, it is theoretically possible that even further improvements in the detection of a weak source can be achieved. The results of this research have shown that regressor coefficients via a Bayesian method are effective as decision parameters. The best results, however, were shown only to match the efficacy of the more traditional, frequentist method of detection.
Open Access
Development of a decision threshold for radiological source detection utilizing Bayesian statistical techniques applied to gross count measurements
(Colorado State University. Libraries, 2018) Brogan, John, author; Brandl, Alexander, advisor; Johnson, Thomas E., committee member; Leary, Del, committee member; Kokoszka, Piotr, committee member
Numerous studies have been published using Bayesian statistics in source localization and identification, characterization of radioactive samples, and uncertainty analysis; but there is a limited amount of material specific to the development of a decision threshold for simple gross count measurements using Bayesian statistics. Radiation detection in low fidelity systems is customarily accomplished through the measurement of gross counts. Difficulties arise when applying decision techniques to low count rate data, which are restricted by the fact that decisions are being made on individual gross count measurements alone. The investigation presented demonstrates a method to develop a viable Bayesian model to detect radiological sources using gross count measurements in low fidelity systems. An integral component of the research is the process required to validate a Bayesian model both statistically and operationally in Health Physics. The results describe the necessary model development, validation steps, and application to the detection of radiological sources at low signal-to-background ratios by testing the model against laboratory data. The approach may serve as a guideline for a series of requirements to integrate Bayesian modeling (specifically, an interaction model) with radiation detection using gross counts in low fidelity systems.
Open Access
Dose profile surrounding a waste repository
(Colorado State University. Libraries, 2014) Parson, Jenelle, author; Brandl, Alexander, advisor; Johnson, Thomas, committee member; Woody, Robert, committee member
The waste repository analyzed is an interim storage facility that houses low and intermediate level conditioned radioactive waste. In total, it contains 9996 200-L waste barrels. The barrels are arranged in a crate geometry to ensure easy access to each barrel. The repository walls are 0.2 m thick with extra shielding (0.7 m) present on the west, north, and east sides of the repository. Instead of extra shielding the south side of the repository has a 5.25 m gap from the crates to the wall, allowing for crane maneuverability. The dose rate profile was analyzed using Monte Carlo N-Particle Transport eXtended (MCNPX) for the south, west, and north sides. The east side was not explicitly analyzed, because of the symmetry between the west and east sides. The dose rate was analyzed using f5 detector tallies and fluence rate to dose rate conversion factors from ICRP 21. Here, contributions due to skyshine and other wall effects are analyzed in detail. For the west and north side (where shielding was present), it was found that as distance from the source increases the dose rate initially increases logarithmically to a maximum and subsequently falls off following an exponential function. The initial increase in dose rate is significant with a peak dose rate as much as 300% of the dose rate at the wall and remaining elevated until approximately 60 m from the waste repository. A similar dose rate increase is not observed for the southern side of the repository; instead, the dose rate falls off with a power function corresponding to a function between that for an infinite plane and a point source. The dose rate profile was analyzed with and without repository structures, and the initial increase is only present with the repository structure. This indicates that the initial rise is due to the presence of walls and/or roof. The shape of the dose rate profile closely follows observed skyshine functions at accelerator facilities. Skyshine and wall effects have been analyzed extensively for medical accelerator facilities, but are generally not considered for a waste repository; the work suggests that skyshine and wall effects may be more significant than previously thought and should be considered in the design and construction of waste repositories.
Open Access
Gross count multi-spectra analysis of weak activity sources amid elevated background and various shielded conditions
(Colorado State University. Libraries, 2018) Fischer, Joseph Christian, author; Brandl, Alexander, advisor; Johnson, Thomas E., committee member; Lindsay, James, committee member
Current methods to detect low activity radiological sources in areas with elevated or changing background may be insufficient for assessing real-time events. Background radiation fluctuates throughout the day, which thwarts the effectiveness of decision thresholds (y*) extrapolated from long background count times. However, decision thresholds calculated from a string of spectral measurements can be used to optimize the decision threshold (y*) at which count rates statistically exceed background. The goal of this project is to apply a frequentist string data analysis tool to real-time gross count sodium iodide (NaI) spectra under varying background radiation and shielding conditions. A sodium iodide scintillator is utilized with the ProSpect acquisition software to acquire gross count data for analysis of the source detection efficiency under varying background and shielding conditions. A series of measurements is taken using short count times, and assessed using Receiver Operating Characteristics (ROC) curves. The best combination of N measurements, with n events individually exceeding a discriminator threshold, is determined using true positive rates from the generated ROC curves. Ideally, the optimized combination of n exceedances of the discriminator threshold in N measurements discerns weak signals from elevated background, while minimizing false positive events.
Open Access
Improved detection of radioactive material using a series of measurements
(Colorado State University. Libraries, 2016) Mann, Jenelle, author; Brandl, Alexander, advisor; Johnson, Thomas, committee member; Kokoszka, Piotr, committee member; Leary, Del, committee member
The goal of this project is to develop improved algorithms for detection of radioactive sources that have low signal compared to background. The detection of low signal sources is of interest in national security applications where the source may have weak ionizing radiation emissions, is heavily shielded, or the counting time is short (such as portal monitoring). Traditionally to distinguish signal from background the decision threshold (y*) is calculated by taking a long background count and limiting the false negative error (α error) to 5%. Some problems with this method include: background is constantly changing due to natural environmental fluctuations and large amounts of data are being taken as the detector continuously scans that are not utilized. Rather than looking at a single measurement, this work investigates looking at a series of N measurements and develops an appropriate decision threshold for exceeding the decision threshold n times in a series of N. This methodology is investigated for a rectangular, triangular, sinusoidal, Poisson, and Gaussian distribution.
Open Access
Internal dosimetric evaluation of Cu-64-ATSM in canine cancer bearing patients
(Colorado State University. Libraries, 2015) Bell, Justin James, author; Brandl, Alexander, advisor; Jones, Brian, committee member; Leary, Del, committee member
The assessment of the novel radiopharmaceutical Copper-64 diacetyl-bis(N4-methylthiosemicarbazones) (64Cu-ATSMs) ability to selectively accumulate in hypoxic tumor tissue has been subject to ongoing research effort at Colorado State Universitie's (CSU) Veterinary Teaching Hospital (VTH). Due to the unique decay scheme of 64Cu and internal distribution when bound to ATSM, the radiopharmaceutical has clinical importance as a possible Positron Emission Tomography/Computed Tomography (PET/CT) imaging agent with cancer theranostic potential. In this study, the PET/CT images of two cancer bearing canine patients treated with 64Cu-ATSM were used to create patient specific voxelized phantoms that were compatible with the Monte Carlo N Particle extended (MCNPX) radiation transport code which was emplyed to retrospectively assess the internal radiation dose each patient received. By defining specific regions of interest (ROI) within the models, the dose to major organs was estimated. The derived models were designed as to dynamically approximate the relative dose each patient would receive to their tumor and normal tissue if the administered activity of 64Cu-ATSM was altered. From these results, it was identified that the patient's malignances received 64Cu-ATSM uptake at least during the acquisition of the patients' images. Organs such as the liver also had relatively high amounts of 64Cu-ATSM uptake and were used to project what ROI might be dose limiting.
Open Access
Investigation of tritium atom exchange in plastic liquid scintillation vials
(Colorado State University. Libraries, 2018) Wang, John Jen-Chiang, author; Brandl, Alexander, advisor; Sudowe, Ralf, committee member; Walrond, John, committee member
Tritium is a naturally occurring radionuclide; it is an analyte of interest in many air, soil, and water samples. It has been shown that long term storage and study of tritium samples results in a reduction in tritium activity not attributed to the natural radioactive decay. Several explanations have been offered through past literatures including diffusion, LSC cocktail degradation, and change in quenching effects. Another possible explanation for the decrease in activity is that tritium may have been organically bound to the plastic possibly due to exchangeable hydrogen atoms along the plastic carbon chain. The hypothesis that tritium can be incorporated into the plastic, interchanging the 1H atoms in the plastic with 3H atoms, was experimentally tested. The experiment consisted of adding deionized water into a previously used plastic vial which had contained tritium to determine if the deionized water had now become tritiated. The results showed that the longer the tritiated water is stored in the vials, the greater the loss of tritium activity in plastic vials is compared to glass vials. An increase in the time that the tritiated water is stored also increases the activity of the tritium found in the deionized water in plastic vials but not in the glass vials. The combination of these two observations supports the hypothesis that tritium exchange may have occurred between the tritiated water and the hydrogen within the plastic vials.
Open Access
Monte Carlo determination of detection efficiency for portal monitoring
(Colorado State University. Libraries, 2024) Blair, Noah Jeffrey, author; Brandl, Alexander, advisor; Sudowe, Ralf, committee member; Jeckel, Kimberly, committee member
At ports of entry into the country, at high security events such as political or athletic gatherings of crowds, and at high risk locations, portal monitors are used to detect the presence of ionizing radiation and ensure radiological/nuclear materials do not fall out of regulatory control or are utilized with malicious intent. This work uses computer models to determine the probability of photon radiation being detected in polyvinyl toluene, a plastic scintillator material, for a range of source energies, truck positions relative to the detector, and cargo materials. These results are used to develop models for operators of portal monitors to predict the activity of a radioactive source given the measured count rate and integrated count measurements. A linear model of the detection eﬀiciency produces an analytic expression of the measured count rate given the emission spectrum of a source which allows for calculation of net counts between two points on the truck's trajectory and the time derivative of the count rate. The time derivative of the count rate has a similar characteristic behavior along the truck's trajectory and has the potential to be an indicator of weak radiological sources. A MARS model allows for accurate prediction of the net count rate for truck positions and source energies not directly modeled. These models best predict detection eﬀiciency for higher energy photons and are primarily useful for the prediction of detection of fission and activation products.
Open Access
Neutron production and transport at a medical linear accelerator
(Colorado State University. Libraries, 2014) Allardice, Amber, author; Brandl, Alexander, advisor; Custis, James, committee member; Lindsay, James, committee member
The Colorado State University Veterinary Teaching Hospital (VTH) uses a Varian Trilogy™ linear accelerator (linac) for radiation oncology treatment. The high-energy electron beam is used to treat superficial tumors (deep tissues are spared with this modality) or is accelerated against a target to produce high-energy photons that are used to treat deep seated tumors (skin is spared with this modality). Either application might exceed the neutron production threshold for various materials. This study evaluates neutron production and transport in the environment surrounding the VTH trilogy through MCNP modeling and physical measurements of the 10 MV photon and 18 MeV electron beam modalities. MCNP modeling was accomplished in two phases. The first phase involved simulating the linear accelerator and determining the relevant parameters for neutron production for both modalities. This was accomplished by using various target specifications and replicating the geometry of the machine. In the second phase, MCNP modeling of the accelerator vault as well as other locations of interest within the treatment suite was conducted. This phase determined measurable neutron fluence and dose rates at the test locations where physical measurements were taken. The MCNP results (for neutron energies between 0.2 to 10 MeV) were compared with the physical measurements. Physical measurements were performed with a BF3 detector (responsive to energies between 0.2 and 10 MeV) and taken at the same test locations. For both modalities, MCNP and physical measurements demonstrated neutron production. Large uncertainties were associated with the physical measurements for both modalities. For the photon mode, MCNP modeling resulted in neutron equivalent doses per photon Gy up to 0.112 mrem/photon Gy, and physical measurements up to 0.133 mrem/photon Gy. For the electron mode, MCNP modeling resulted in measureable neutron equivalent doses per electron Gy up to 14.88 mrem/electron Gy, and physical measurements up to 3.83 x 10-04 mrem/electron Gy. Taking the entire neutron spectrum into account, MCNP results showed neutron doses up to 347.81 mrem/ photon Gy at the isocenter for the photon beam, and up to 1.77 x 105 mrem/electron Gy at the isocenter for the electron beam. These numbers could not be compared to the physical measurements because the BF3 detector used in this experiment only responded to neutron energies between 0.2 and 10 MeV. The conclusion made from this research is that neutrons are generated at various locations in and outside the room. For the photon modality, the neutron dose to the patient can be considered negligible when compared with the treatment dose. Neutron production does not appear to exceed the tolerance for workers in appropriate locations surrounding the VTH linac vault. Further research is recommended for an accurate analysis of both modalities.
Open Access
Neutron/muon correlation functions to improve neutron detection capabilities outside nuclear facilities
(Colorado State University. Libraries, 2016) Ordinario, Donald Thomas, author; Brandl, Alexander, advisor; Johnson, Thomas, committee member; Wilson, Robert, committee member
The natural neutron background rate is largely due to cosmic ray interactions in the atmosphere and the subsequent neutron emission from the interaction products. The neutron background is part of a larger cosmic radiation shower that also includes electrons, gamma rays, and muons. Since neutrons interact much differently than muons in building materials, the muon and neutron fluence rates in the natural background can be compared to the measured muon and neutron fluence rate when shielded by common building materials. The simultaneous measurement of muon and neutron fluence rates might allow for an earlier identification of man-made neutron sources, such as hidden nuclear materials. This study compares natural background neutron rates to computer simulated neutron rates shielded by common structural and building materials. The characteristic differences between neutrons and muons resulted in different attenuation properties under the same shielded conditions. Correlation functions between cosmic ray generated neutrons and muons are then used to predict neutron fluence rates in different urban environments.