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Thesis (doctoral)--Delft University of Technology, 2007. / "Proefschrift ter verkrijging van de graad van doctor aan de Technische Universiteit Delft." Includes bibliographical references (p. 120-128) and index.

Thesis (doctoral)--Delft University of Technology, 2007. / "Proefschrift ter verkrijging van de graad van doctor aan de Technische Universiteit Delft." Includes bibliographical references (p.120-128) and index.

Thesis (doctoral)--Delft University of Technology, 2007. / "Proefschrift ter verkrijging van de graad van doctor aan de Technische Universiteit Delft." Includes bibliographical references (p.120-128) and index.

In the late 1970's, emphasis on non-proliferation forced suspension of all commercial spent-fuel reprocessing. The spent-fuel storage problem plaguing the nuclear industry can be alleviated by reprocessing. For commercial spentfuel reprocessing to again become a reality, a process is needed to reform reprocessing operations such that non-proliferation goals are satisfied. To satisfy these goals, the existing process which generates plutonium-nitrate solution must be altered to generate plutonium-uranium oxide powder. The COPRECAL process is designed to produce this solid. The COPRECAL process allows uranium and plutonium to be extracted from spent-fuel for reuse in commercial lightwater reactors. The COPRECAL process is unique in that no pure plutonium is ever present throughout the process, whether the COPRECAL process is intrinsically vulnerable to plutonium diversion is the object of this work. A simulation model of the COPRECAL process is presented which employs state-of-the-art instrumentation to measure in-process plutonium through the simulated passage of time. Plutonium diversion schemes are incorporated into the model. After simulated thefts, model output statistics are plotted on control charts and analyzed. Results show need for major design changes in the COPRECAL process. / Master of Science / incomplete_metadata

Thesis: M.S., Massachusetts Institute of Technology, Department of Nuclear Engineering, 1977 / Includes bibliographical references. / M.S. / M.S. Massachusetts Institute of Technology, Department of Nuclear Engineering

Call number: LD2668 .T4 1985 M66 / Master of Science

The goal of this research was to identify and fabricate zeolitic membranes that can separate radioisotope krypton-85 (half-life 10.72 years) and xenon gas released during spent nuclear fuel reprocessing. In spent nuclear fuel reprocessing, fissionable plutonium and uranium are recovered from spent nuclear fuel and recycled. During the process, krypton-85 and xenon are released from the spent nuclear fuel as process off-gas. The off-gas also contains NO, NO2, 129I, 85Kr, 14CO2, tritium (as 3H2O), and air and is usually vented to the atmosphere as waste without removing many of the radioactive components, such as 85Kr. Currently, the US does not reprocess spent nuclear fuel. However, as a member of the International Framework for Nuclear Energy Cooperation (IFNEC, formerly the Global Nuclear Energy Partnership), the United States has partnered with the international nuclear community to develop a “closed” nuclear fuel cycle that efficiently recycles all used nuclear fuel and safely disposes all radioactive waste byproducts. This research supports this initiative through the development of zeolitic membranes that can separate 85Kr from nuclear reprocessing off-gas for capture and long-term storage as nuclear waste. The implementation of an 85Kr/Xe separation step in the nuclear fuel cycle yields two main advantages. The primary advantage is reducing the volume of 85Kr contaminated gas that must be stored as radioactive waste. A secondary advantage is possible revenue generated from the sale of purified Xe. This research proposed to use a zeolitic membrane-based separation because of their molecular sieving properties, resistance to radiation degradation, and lower energy requirements compared to distillation-based separations. Currently, the only commercial process used to separate Kr and Xe is cryogenic distillation. However, cryogenic distillation is very energy intensive because the boiling points of Kr and Xe are -153 °C and -108 °C, respectively. The 85Kr/Xe separation step was envisioned to run as a continuous cross-flow filtration process (at room temperature using a transmembrane pressure of about 1 bar) with a zeolite membrane separating krypton-85 into the filtrate stream and concentrating xenon into the retentate stream. To measure process feasibility, zeolite membranes were synthesized on porous α-alumina support discs and permeation tested in dead-end filtration mode to measure single-gas permeance and selectivity of CO2, CH4, N2, H2, He, Ar, Xe, Kr, and SF6. Since the kinetic diameter of krypton is 3.6 Å and xenon is 3.96 Å, zeolites SAPO-34 (pore size 3.8 Å) and DDR (pore size 3.6 Å) were studied because their pore sizes are between or equal to the kinetic diameters of krypton and xenon; therefore, Kr and Xe could be separated by size-exclusion. Also, zeolite MFI (average pore size 5.5 Å) permeance and selectivity were evaluated to produce a baseline for comparison, and amorphous carbon membranes (pore size < 5 Å) were evaluated for Kr/Xe separation as well. After permeation testing, MFI, DDR, and amorphous carbon membranes did not separate Kr and Xe with high selectivity and high Kr permeance. However, SAPO-34 zeolite membranes were able to separate Kr and Xe with an average Kr/Xe ideal selectivity of 11.8 and an average Kr permeance of 19.4 GPU at ambient temperature and a 1 atm feed pressure. Also, an analysis of the SAPO-34 membrane defect permeance determined that the average Kr/Xe selectivity decreased by 53% at room temperature due to unselective defect permeance by Knudsen diffusion. However, sealing the membrane defects with polydimethylsiloxane increased Kr/Xe selectivity by 32.8% to 16.2 and retained a high Kr membrane permeance of 10.2 GPU at ambient temperature. Overall, this research has shown that high quality SAPO-34 membranes can be consistently fabricated to achieve a Kr/Xe ideal selectivity >10 and Kr permeance >10 GPU at ambient temperature and 1 atm feed pressure. Furthermore, a scale-up analysis based on the experimental results determined that a cross-flow SAPO-34 membrane with a Kr/Xe selectivity of 11.8 and an area of 4.2 m2 would recover 99.5% of the Kr from a 1 L/min feed stream containing 0.09% Kr and 0.91% Xe at ambient temperature and 1 atm feed pressure. Also, the membrane would produce a retentate stream containing 99.9% Xe. Based on the SAPO-34 membrane analysis results, further research is warranted to develop SAPO-34 membranes for separating 85Kr and Xe.

<p> Some experiments designed to test the validity of the edge-spread function (ESF) model for neutron radiographic image formation are described; in addition the experiments are meant to illustrate the application of ESF methods to two areas of practical concern. First, the prediction of optical density curves for specified material and geometric configurations is considered; then, the use of ESF methods in dimensioning irradiated reactor fuel elements is examined. Overall, the results indicate that within the framework of assumptions which ESF theory is based upon, the correlation between theory and experiment is excellent. The results also suggest that in situations which deviate from the theoretical ideal, the ESF method may serve as a good first approximation to more complex models. </p> / Thesis / Master of Engineering (ME)

One of the proposed solutions to the issue of nuclear waste volume is to transition from once through nuclear fuel cycle to advanced fuel cycles with used fuel recycling option. In any advanced fuel cycles with recycling options, the type and amount of separation technology deployed play a crucial role in the overall performance of the fuel cycle. In this work, a scenario study involving two advanced fuel cycles in addition to the once through fuel cycle were evaluated using VISION nuclear fuel cycle simulation code. The advanced fuel cycles were setup to transition completely to full recycling without any light water reactor by assuming all LWR currently in operation will have 20 years of operating life extension and no new LWR will be constructed thereafter. Several different separation capacities (1kT/yr, 2kT/yr and 4 kT/yr) were deployed and the overall impact of these capacities was analyzed in terms of resources utilization, used fuel and waste material generated and the amount of storage space required. Economic parameter (LCOE, LFCC, etc) analysis was also performed using VISION.ECON. Results presented in this work suggest that the need for LWR-UNF storage can be minimized if sufficient separation capacity is deployed early in the fuel cycle. It can also be concluded that a FuRe system without LEU will not be feasible, thus SFRs must be designed for optional use of LEU fuel. Otherwise LWRs must continue to be part of the mix to keep the near term cost of generating electricity competitive. It was observed that the higher amount of separation capacity deployed in the advanced fuel cycles led to higher LFCC and LCOE, but also translates into less environmental impact on both front and back end of the fuel cycle.