Spent nuclear fuel is received at a storage facility in heavily shielded casks transported by either rail or by truck. Once at the storage facility, the casks are inspected, emptied, decontaminated, and reshipped. Allied-General Nuclear Services' (AGNS) nuclear fuel reprocessing plant in Barnwell, South Carolina, is constructed but not yet licensed for spent nuclear fuel storage or reprocessing. Recently, however, AGNS was granted funds by the Department of Energy to prepare the necessary procedural and regulatory paperwork in order that the Fuel Receiving and Storage Station (FRSS) of the plant can be licensed by 1985. In this paper, the activities involved in the receiving an unloading of casks at the Barnwell FRSS is simulated by computer using IBM's program software package, General Purpose Simulation System (GPSS). The GPSS model is developed and verified, and steady-state output statistics are achieved. Also, several sensitivity analyses are performed such as, changes in expected arrival schedules and decision policies, and changes to the physical characteristics of the existing FRSS to monitor the effect of these changes in the existing system.

Various fuel cycles for a sodium-cooled, subcritical, fast reactor with a fusion neutron source for the transmutation of light water reactor spent fuel have been analyzed. All fuel cycles were 4-batch, and all but one were constrained by a total fuel residence time consistent with a 200 dpa clad and structure materials damage limit. The objective of this study was to achieve greater than 90% burn up of the transuranics from the spent fuel.

The United States' fleet of Light Water Reactors (LWRs) produces a large amount of spent fuel each year; all of which is presently intended to be stored in a fuel repository for disposal. As these LWRs continue to operate and more are built to match the increasing demand for electricity, the required capacity for these repositories grows. Georgia Tech's Subcritical Advanced Burner Reactor (SABR) has been designed to reduce the capacity requirements for these repositories and thereby help close the back end of the nuclear fuel cycle by burning the long-lived transuranics in spent nuclear fuel. SABR's design is based heavily off of the Integral Fast Reactor (IFR). It is important to understand whether the SABR design retains the passive safety characteristics of the IFR. A full safety analysis of SABR's transient response to various possible accident scenarios needs to be performed to determine this. However, before this safety analysis can be performed, it is imperative to model all components of the reactivity feedback mechanism in SABR. The purpose of this work is to develop a calculational model for the fuel bowing reactivity coefficients that can be used in SABR's future safety analysis. This thesis discusses background on fuel bowing and other reactivity coefficients, the history of the IFR, the design of SABR, describes the method that was developed for calculating fuel bowing reactivity coefficients and its validation, and presents an example of a fuel bowing reactivity calculation for SABR.

Spent nuclear fuel (SNF) resultant from the generation of nuclear power is a chemically and radiologically diverse system which is advantageous to chemically process prior to geologic disposal. Hydrometallurgy is the primary technology for chemical processing for light water reactor spent fuels, where spent fuel is dissolved in an acid for liquid based separations. The primary means for recovery of desired metals from the SNF solution is liquid-liquid extraction which is based on distribution (partitioning) of the metal ions between two immiscible phases based on thermodynamic favorability. One of the means of increasing this favorability is by designing extractant molecules to be either "harder" or "softer" bases, which will more preferentially extract harder or softer metal ions respectively. This technique is used in designing extractant molecules for targeted extraction as actinides are slightly softer than lanthanides, and precious metals produced in significant quantities from the fission process are especially soft metals. The work performed in this thesis involved the synthesis of a novel soft electron donor organic extractant molecule for testing of targeted soft metal extraction. The molecule synthesized was bis-dibutanethiolthiophosphinato-methane, or S6, a bidentate neutral extractant molecule with significant thiolysis for a softer electron environment. The synthesis technique was refined and the molecule composition and structure was confirmed by ¹H NMR, ³¹P NMR, and elemental analysis. Two metal groups, f-elements (actinides and lanthanides) and soft transition metals were tested for their extractability from nitric acid solutions into an S6 solution in n-dodecane. Aqueous solutions of nitric acid and n-dodecane as an organic diluent are typical liquid-liquid extraction conditions in spent nuclear fuel reprocessing. As extraction experiments were performed with radiotracers, for the soft metal extraction experiment, a mixture of the selected metals was neutron-activated in the OSU TRIGA reactor, as was europium to create a lanthanide radiotracer. Actinides and lanthanides were not seen to effectively extract into the organic or form a precipitate at all, making their partitioning with this extractant seemingly ineffective. Through gamma spectroscopy of an irradiated metal solution post-extraction, it is seen that only silver and palladium preferentially complex in the mixed metal samples into an insoluble organic ligand, dropping out of solution. This effect was more pronounced at higher acid concentrations, but silver was seen to slightly extract to the organic phase at all acid concentrations as well. This testing has shown that the S6 extractant can be used to recover silver and palladium from a mixed metal aqueous solution, such as one resultant from advanced spent nuclear fuel reprocessing operations. This result shows promise for future development of sulfur based organophosphate ligands for targeted extraction of precious metals from solutions. / Graduation date: 2013

A series of fuel cycle simulations were performed using CEA's reactor physics code ERANOS 2.0 to analyze the transmutation performance of the Subcritical Advanced Burner Reactor (SABR). SABR is a fusion-fission hybrid reactor that combines the leading sodium cooled fast reactor technology with the leading tokamak plasma technology based on ITER physics. Two general fuel cycles were considered for the SABR system. The first fuel cycle is one in which all of the transuranics from light water reactors are burned in SABR. The second fuel cycle is a minor actinide burning fuel cycle in which all of the minor actinides and some of the plutonium produced in light water reactors are burned in SABR, with the excess plutonium being set aside for starting up fast reactors in the future. The minor actinide burning fuel cycle is being considered in European Scenario Studies. The fuel cycles were evaluated on the basis of TRU/MA transmutation rate, power profile, accumulated radiation damage, and decay heat to the repository. Each of the fuel cycles are compared against each other, and the minor actinide burning fuel cycles are compared against the EFIT transmutation system, and a low conversion ratio fast reactor.

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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.

The purpose of this research is to develop a methodology for a thorium fuel recycling analysis that provides results for isotopics and radio-toxicity evaluation and analysis. This research is motivated by the need to reduce the long term radiological hazard in spent nuclear fuel, which mitigates the mixing hazard (radiotoxicity and chemical toxicity) and decay heat load on the repository. The first part of the thesis presents comparison of several once-through cases with uranium and thorium fuels to show how transuranics build up as fuel is depleted. The once-through analysis is performed for the following pairs of comparison cases: low enriched uranium dioxide (UOX) vs. thorium dioxide with 233UOX (233U-ThOX), natural uranium dioxide mixed with transuranic oxides (U-TRUOX) vs. thorium dioxide mixed with transuranic oxides (Th-TRUOX), natural uranium dioxide mixed with weapons grade plutonium dioxide (U-WGPuOX) vs. thorium dioxide mixed with weapons grade plutonium dioxide (Th-WGPuOX), natural uranium dioxide mixed with reactor grade plutonium dioxide (U-RGPuOX) vs. thorium mixed with reactor grade plutonium dioxide (Th-RGPuOX). The second part of the research evaluates the thorium fuel equilibrium cycle in a pressurized water reactor (PWR) and compares several recycling cases with different partitioning schemes. Radio-toxicity results of the once-through cycle and multi-recycle calculations demonstrate advantages for thorium fuel and reprocessing with respect to long term nuclear waste management.