Graduation date: 1966

Graduation date: 1966

Graduation date: 1963

Fusion reactions with radioactive neutron-rich projectiles have been the subject of much recent theoretical and experimental interest. Predictions of enhancement of the cross section due to the use of a neutron-rich projectile may have implications for synthesis of heavy nuclei. In this work, the fusion-fission excitation functions were measured for the [superscript 32,38]S + �������Ta reactions. The radioactive �����S beam was produced by projectile fragmentation. In the ����S-induced reaction, an incomplete fusion component was observed at high energies, with average momentum transfer corresponding to escape of an alpha particle. Angular distribution data were used to estimate the quasifission component of the stable-beam reaction. The excitation functions were analyzed using classical and coupled-channels methods; the deduced interaction barriers were 130.7 �� 0.3 MeV and 124.8 �� 0.3 MeV for the ����S- and �����S-induced reactions, respectively. No evidence of any additional mechanism beyond a simple shift in the Coulomb barrier was observed. Taking into account the difference in reaction Q-values, the net lowering of the compound nucleus excitation energy at the barrier is about 12 MeV due to the use of the radioactive neutron-rich projectile; this could significantly affect survival probabilities of heavy nuclei. / Graduation date: 2000

More and more, modern society is incorporating the use of radioactive materials into everyday uses. And with society using more radioactive materials, the odds of it being accidentally disposed of into the solid waste stream increases. There are several radiation systems available which market themselves as being complete and "ready to go". While it is true that a person could purchase one of these systems and would have coverage of the landfill, such a system does not provide the necessary education, response and liability protection programs. Indeed, it would be feasible to foresee a scenario where installing a systems could lead to an increase in liability and employee problems. As a result, Coffin Butte Landfill worked with the author to establish a complete radiation monitoring program. This program encompasses everything from installment of the system to employee education and training. It also examined the myriad and murky depths of federal and state regulation dealing with solid and radioactive waste to help the landfill set an acceptance policy and minimize liability. This led the author to the belief that the combination of federal and state regulations imply a requirement for landfills to have a working radiation monitoring program. Future government action remains uncertain as pertaining to a requirement for landfills to maintain a radiation monitoring system. Indeed, current state regulations are out of sync with federal regulations regarding acceptable public exposures. It is hoped that, with this study's help, Coffin Butte Landfill and Oregon State University will continue with the established relationship and be prepared to respond to regulation changes. / Graduation date: 1999

Graduation date: 1996

The International Commission on Radiological Protection (ICRP) has modeled twelve reference animal and plant (RAP) species using simple geometric shapes in Monte���Carlo (MCNP) based simulations. The focus has now shifted to creating voxel phantoms of each RAP to advance the understanding of radiation interactions in nonhuman biota. The work contained herein presents results for the voxel phantom of the Dungeness crab, Metacarcinus magister, the Sand Dab, Limanda limanda, and the brown seaweed, Fucus vesiculosus, and details a generalized framework for creating voxel phantoms of the other RAPs. Absorbed fractions (AFs) for all identified organs were calculated at several discrete initial energies: 0.01, 0.015, 0.02, 0.03, 0.05, 0.1, 0.2, 0.5, 1.0, 1.5, 2.0, and 4.0 MeV for photons and 0.1, 0.2, 0.4, 0.5, 0.7, 1.0, 1.5, 2.0 and 4.0 MeV for electrons. AFs were then tabulated for each organ as a source and target at each energy listed above. AFs whose error exceeded 5% are marked with an underline in the data tables; AFs whose error was higher than 10% are shown in the tabulated data as a dashed line. The AF���s were highly dependent on organ mass and geometry. For photons above 0.5 MeV and electrons above 0.2-0.4 MeV a nontrivial amount of energy escapes the source organ. / Graduation date: 2013

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.

The 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 μCi/liter. Detection of system transients was observed to improve with decreasing flow rate. The detection limits observed for this system were 0.02 μCi/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.

Thesis (Ph. D.)--Lehigh University, 2002. / Includes vita. Includes bibliographical references (leaves 128-134).