Browsing by Subject "radioactive"
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Item External detection and measurement of inhaled radionuclides using thermoluminescent dosimeters(Texas A&M University, 2007-04-25) Prause, Christopher AlvinMany radiation detection programs use bio-assays, whole-body counters, or air sampling to estimate internal doses. This study examines the possibility of using a common external thermoluminescent dosimeter (TLD) badge as a device for detecting inhaled radionuclides through radiation those radionuclides emit which escape the body. The three common radionuclides chosen for modeling due to their varying decay modes and use or production in the nuclear industry were Cs-137, U-238, and Sr-90. These three radionuclides were modeled for biological and radiological removal in the dynamic systems modeling program of STELLA II and modeled for TLD dose per organ in the geometry and radiation simulation program of MCNP. The results show that none of the nuclides in the study can be detected at air concentrations below regulatory limits for acute inhalation exposures. To achieve a detectable dose from an 8-hour work exposure, with a 90-day wait until the TLD is read, the airborne concentrations for the inhalation classes that produced the most dose per Bq would be 37.9 kBq/m3, 146 MBq/m3, and 1.67 MBq/m3 for Cs-137, U-238, and Sr-90, respectively.Item Radioactive Flow Characterization for Real-Time Detection Systems in UREX+ Nuclear Fuel Reprocessing(2011-02-22) Hogelin, Thomas RussellThe reprocessing of used nuclear fuel requires the dissolution and separation of numerous radioisotopes that are present as fission products in the fuel. The leading technology option in the U.S. for reprocessing is a sequence of processing methods known as UREX+ (Uranium Extraction ). However, an industrial scale facility implementing this separation procedure will require the establishment of safeguards and security systems to ensure the protection of the separated materials. A number of technologies have been developed for meeting the measurement demands for such a facility. This project focuses on the design of a gamma detection system for taking measurements of the flow streams of such a reprocessing facility. An experimental apparatus was constructed capable of pumping water spiked with soluble radioisotopes under various flow conditions through a stainless steel coil around a sodium iodide (NaI) detector system. Experiments were conducted to characterize the impact of flow rate, pipe air voids, geometry, and radioactivity dilution level on activity measurements and gamma energy spectra. Two coil geometries were used for these experiments, using 0.5 in stainless steel pipe wound into a coil with a 6 inch diameter; the first coil was 5.5 revolutions tall and the second coil was 9.5 revolutions tall. The isotopes dissolved in the flowing water were produced at the Texas A&M Nuclear Science Center via neutron activation of chromium, gold, cerium, and ytterbium nitrate salts. After activation, the salts were dissolved in distilled water and inserted into the radioactive flow assembly for quantitative measurements. Flow rate variations from 100 to 2000 ml/min were used and activity dilution levels for the experiments conducted were between 0.02 and 1.6 ?Ci/liter. Detection of system transients was observed to improve with decreasing flow rate. The detection limits observed for this system were 0.02 ?Ci/liter over background, 0.5% total activity change in a pre-spiked system, and a dilution change of 2% of the coil volume. MCNP (Monte Carlo N-Particle Transport) models were constructed to simulate the results and were used to extend the results to other geometries and piping materials as well as simulate actual UREX stream material in the system. The stainless steel piping for the flow around the detector was found to attenuate key identifying gamma peaks on the low end of the energy spectrum. For the proposed schedule 40 stainless steel pipe for an actual reprocessing facility, gamma rays below 100 keV in energy would be reduced to less than half their initial intensities. The exact ideal detection set up is largely activity and flow stream dependant. However, the characteristics best suited for flow stream detection are: 1) minimize volume around detector, 2) low flow rate for long count times, and 3) low attenuation piping material such as glass.Item Radon (Rn-222) and thoron (Rn-220) emanation fractions from three separate formations of oil field pipe scale(Texas A&M University, 2004-11-15) Fruchtnicht, Erich HaroldOver the course of normal oil well operations, pipes used downhole in the oil and petroleum industry tend to accumulate a mineral deposit on their interior, which restricts the flow of oil. This deposit, termed scale, will eventually occlude the interior diameter of the pipe making removal from service and descaling a cost effective option. The pipes are sent to cleaning yards where they remain until descaling can be performed. This storage period can potentially create a health concern not only because of the external radiation exposure but also because of the radon gas emissions, both of which are due to the radioactive minerals contained in the scale. It was believed that the structure of the scale is formed tightly enough to prevent much of the radon from becoming airborne. The goal of this research was to determine the emanation fractions for the rattled scale samples from three formations. A high purity germanium detector was used to measure the activities of the parents and progeny of radon, and electret ion chambers were used to measure the concentration of radon emanated from the scale. The emanation fractions of between 4.9x10-5 and 1.08x10-3 for radon were a factor of approximately 100 smaller than previous research results. For thoron, the fractions were and 5.72x10-8 and 4.92x10-7 for thoron with no previous research to compare. However, information that pertains to the temperature dependence of emanation was included in this research and was not available for previous, similar research. Therefore, differences in the environment (e.g., temperature, humidity, etc.) in which the previous experiments were conducted, as well as differences in the scale formation types used, could account for the discrepancy. In addition, measuring the emanation fractions of the rattled scale was a method of determining whether surface to volume ratio dependence existed. After acquiring the emanation fractions, insufficient evidence of any surface to volume ratio dependence could be found.Item Sintered Bentonite Ceramics for the Immobilization of Cesium- and Strontium-Bearing Radioactive Waste(2010-07-14) Ortega, Luis H.The Advanced Fuel Cycle Initiative (AFCI) is a Department of Energy (DOE) program, that has been investigating technologies to improve fuel cycle sustainability and proliferation resistance. One of the program's goals is to reduce the amount of radioactive waste requiring repository disposal. Cesium and strontium are two primary heat sources during the first 300 years of spent nuclear fuel's decay, specifically isotopes Cs-137 and Sr-90. Removal of these isotopes from spent nuclear fuel will reduce the activity of the bulk spent fuel, reducing the heat given off by the waste. Once the cesium and strontium are separated from the bulk of the spent nuclear fuel, the isotopes must be immobilized. This study is focused on a method to immobilize a cesium- and strontium-bearing radioactive liquid waste stream. While there are various schemes to remove these isotopes from spent fuel, this study has focused on a nitric acid based liquid waste. The waste liquid was mixed with the bentonite, dried then sintered. To be effective sintering temperatures from 1100 to 1200 degrees C were required, and waste concentrations must be at least 25 wt%. The product is a leach resistant ceramic solid with the waste elements embedded within alumino-silicates and a silicon rich phase. The cesium is primarily incorporated into pollucite and the strontium into a monoclinic feldspar. The simulated waste was prepared from nitrate salts of stable ions. These ions were limited to cesium, strontium, barium and rubidium. Barium and rubidium will be co-extracted during separation due to similar chemical properties to cesium and strontium. The waste liquid was added to the bentonite clay incrementally with drying steps between each addition. The dry powder was pressed and then sintered at various temperatures. The maximum loading tested is 32 wt. percent waste, which refers to 13.9 wt. percent cesium, 12.2 wt. percent barium, 4.1 wt. percent strontium, and 2.0 wt. percent rubidium. Lower loadings of waste were also tested. The final solid product was a hard dense ceramic with a density that varied from 2.12 g/cm3 for a 19% waste loading with a 1200 degrees C sintering temperature to 3.03 g/cm3 with a 29% waste loading and sintered at 1100 degrees C. Differential Scanning Calorimetry and Thermal Gravimetric Analysis (DSC-TGA) of the loaded bentonite displayed mass loss steps which were consistent with water losses in pure bentonite. Water losses were complete after dehydroxylation at ~650 degrees C. No mass losses were evident beyond the dehydroxylation. The ceramic melts at temperatures greater than 1300 degrees C. Light flash analysis found heat capacities of the ceramic to be comparable to those of strontium and barium feldspars as well as pollucite. Thermal conductivity improved with higher sintering temperatures, attributed to lower porosity. Porosity was minimized in 1200 degrees C sinterings. Ceramics with waste loadings less than 25 wt% displayed slump, the lowest waste loading, 15 wt% bloated at a 1200 degrees C sintering. Waste loading above 25 wt% produced smooth uniform ceramics when sintered >1100 degrees C. Sintered bentonite may provide a simple alternative to vitrification and other engineered radioactive waste-forms.