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At least 1,153 records · Page 64
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Electrochemical behavior of uranyl in anhydrous polar organic media

Abstract Weak complexes between pentavalent and hexavalent actinyl cations have been reported to exist in acidic, non-complexing high ionic strength aqueous media. Such “cation-cation complexes” were first identified in the context of actinide-actinide redox reactions in acidic aqueous media relevant to solvent extraction-based separation systems, hence their characterization is of potential interest for advanced nuclear fuel reprocessing. This chemistry could be relevant to efforts to develop advanced actinide separations based on the upper oxidation states of americium, which are of current interest. In the present study, the chemical behavior of pentavalent uranyl was examined in non-aqueous, aprotic polar organic solvents (propylene carbonate and acetonitrile) to determine whether UO

Chemistry↗
DOE OSTI2025

Hestia-SWIFL: hourly anthropogenic fossil fuel CO2 and heat on the 2km WRF grid, version 1.1

The Hestia-SWIFL version 1.1 anthropogenic heat (AH) and fossil fuel CO2 (FFCO2) emissions data product represent emissions due to the combustion of fossil fuel and cement production within the state of Arizona from 2019 to 2022. This product was developed as part of the Southwest Urban Corridor Integrated Field Laboratory (SW-IFL) project, which aims to provide new knowledge and tools that address extreme heat, air quality, climate change and related urban environmental issues by integrating high-resolution observations, modeling, and civic engagement. The emissions are generated using a bottom-up/engineering approach and are tied to results generated by the Vulcan Project version 4, an effort to quantify space/time-resolved FFCO2 & AH emissions for the entire United States landscape. A large number of data sources are combined to best estimate the emissions at fine scales such as air quality emissions data, traffic flow data, building information, sociodemographic information, and fuel statistics. The AH product provides emissions for two emissions sources (transportation and point source emissions) in units of Watts per hour per square meter (W/m2) per year (annual files) or per hour (hourly files). The FFCO2 product provides emissions from nine individual emission sectors as well as the total, and in units of tons of carbon (tC) per grid cell per year or per hour. The output made available here places the native spatial resolution of the Hestia FFCO2 & AH emissions data product (points, lines, and polygons) into a regularized 2km x 2km grid at hourly and annual temporal resolutions, and stored in netCDF files. The exact spatial extent is defined by the ASU Weather Research Forecast (WRF) simulation grid. All data are processed using R/Python pm high-performance computing system. 2-27-2026 updates: Bugs in airport hourly profile (both AH and FFCO2) and building spatial patterns (FFCO2 only) were fixed. Hourly emissions are reprocessed for all years to reflect those changes.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2019

Method Development for High Temperature In-Situ Neutron Diffraction Measurements of Glass Crystallization on Cooling from Melt

A glass-ceramic borosilicate waste form is being considered for immobilization of waste streams of alkali, alkaline-earth, lanthanide, and transition metals generated by transuranic extraction for reprocessing used nuclear fuel. Waste forms are created by partial crystallization on cooling, primarily of oxyapatite and powellite phases. In-situ neutron diffraction experiments were performed to obtain detailed information about crystallization upon cooling from 1200°C. The combination of high temperatures and reactivity of borosilicate glass with typical containers used in neutron experiments, such as vanadium and niobium, prevented their use here. Thus, methods using sealed thick-walled silica ampoules were developed for the in-situ studies. Unexpectedly, high neutron absorption, low crystal fraction, and high silica container background made quantification difficult for these high temperature measurements. As a follow-up, proof-of-concept measurements were performed on different potential high-temperature container materials, emphasizing crystalline materials so that residual glass in the waste form sample could be more easily analyzed. Room temperature measurements were conducted with a pre-crystallized sample in ‘ideal’ containers stable at low temperatures (i.e., vanadium and thin-wall silica capillaries) and compared to the same measurements in containers stable at high temperatures (i.e, platinum, single crystal sapphire, and thick-walled silica ampoules). Results indicated that Pt is probably the best choice if suitably sealed to prevent contamination from the sample after neutron activation.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Process optimization of caustic scrubber and iodine-129 immobilization in sodalite-based waste forms

Caustic scrubbers are proposed to remove vaporized iodine from nuclear fuel reprocessing plants. Resulting caustic slurries of radioactive iodine ( 129 I) and other anions can be captured and immobilized into the cage structures of sodalite-type minerals. Here, mixed anion sodalites were hydrothermally synthesized from a simulated caustic scrubber solution containing hydroxide/water, carbonate, chlorine, bromine, iodine, nitrate, and nitrite ions with added kaolinite. Experiments were conducted with twenty-five different sets of process conditions. Resulting powders were characterized by quantitative X-ray diffraction (XRD) and multi-functional thermal analysis, including gravimetry, calorimetry, and mass spectrometry. Using statistical analysis, process variables were tailored to maximize conversion of kaolinite into cage silicates, immobilize caustic scrubber anions, and minimize amorphous content. Further, chemical analysis showed the incorporation of all the targeted ions into the solids. Detailed investigation of a batch characterized after washing (creating secondary waste) and one without washing but subsequent calcination is given. This work contributes to defining the constraints required for subsequent consolidation process steps of the waste form to avoid volatilization.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Analysis of Microseismicity and Reactivated Fault Size to Assess the Potential for Felt Events by CO 2 Injection in the Illinois Basin

The results of monitoring of carbon dioxide (CO 2 ) injection at the Illinois Basin—Decatur Project (IBDP) and the companion Illinois Industrial Carbon Capture and Sequestration Sources (IL-ICCS) project—have shown that reservoir response to fluid pressure changes can vary significantly at different injection locations within the same reservoir. Predrill reservoir characterization is important to identify potentially seismogenic faults. However, interpretations of newly reprocessed 3D seismic reflection data illustrate the challenges related to their identification in a region dominated by faulting with small vertical offsets. Faults interpreted in the 3D seismic volume range from ~300 to 1200 m wide and are in the same size range as faults that could have been the source of historical events up to Mw 2.7 in central Illinois. The array of monitoring sensors that was installed for the IBDP continues to collect data, as injection operates in IL-ICCS, the second injection well. CO 2 injection rates for the IL-ICCS well are on average 1.7 times the rates injected in the IBDP well, but a significantly reduced rate of induced seismicity is observed. This article presents results of passive seismic monitoring for the duration of the project to date, integrating active and passive seismic data to develop a new interpretation of the subsurface structure at the Decatur site that explicitly identifies pathways for fluid flow into the basement leading to induced seismicity, and provides a geological explanation for the sharp reduction of induced seismicity during injection at higher rates into the second well. The use of seismic moment to estimate the length of seismogenic slip planes in the local subsurface suggests that faults large enough to produce felt seismicity are unlikely to be present at or near the Decatur site.

Geochemistry & Geophysics↗
DOE OSTI2018

Standardized beamline particulate characterization analysis: Initial application to CEBAF and LCLS-II cryomodule components

Despite continuously evolving efforts to minimize and manage the occurrence of particulates in operational SRF accelerator systems, the presence of electron field emission from contaminating particulates on SRF surfaces with high surface electric fields remains a consistent challenge. Jefferson Lab has recently initiated a standardized particulate sampling and characterization practice in order to gain more specific and systematic knowledge of the particulates actually present. It is expected that patterns emerging from such sampling will strengthen source attribution and guide improvement efforts. Initial samples were gathered from a cryomodule and adjoining warm girders that were removed from the CEBAF tunnel for reprocessing. The collection and analysis techniques were also used to characterize particulates on the inside of LCLS-II string components. Samples are transferred to clean industry-standard forensic GSR carbon tape spindles and examined via automated cleanroom SEM scanning for particle localization trends and sizing. The particulates are subsequently analyzed with EDS for elemental composition. A catalogue of found particle types is being accumulated. The methods used and results obtained from these initial applications will be presented.

Reece, Charles↗
DOE OSTI2018

Why natural monazite never becomes amorphous: Experimental evidence for alpha self-healing

Monazite, a common accessory rare-earth orthophosphate mineral in the continental crust widely used in U-Pb geochronology, holds promise for (U-Th)/He thermochronology and for the immobilization of Pu and minor actinides (MA) coming from spent nuclear fuel reprocessing. Previous results obtained on natural and plutonium-doped monazite have demonstrated the ability of this structure to maintain a crystalline state despite high radiation damage levels. However, the low critical temperature (180 °C), above which amorphization cannot be achieved in natural monazite under ion irradiation, does not explain this old and unsolved paradox: why do natural monazites, independent of their geological history, remain crystalline even when they did not experience any thermal event that could heal the defects? This is what the present study aims to address. Synthetic polycrystals of LaPO 4 -monazite were irradiated sequentially and simultaneously with α particles (He) and gold (Au) ions. Here, our results demonstrate experimentally for the first time in monazite, the existence of the defect recovery mechanism, called α-healing, acting in this structure due to electronic energy loss of α particles, which explains the absence of amorphization in natural monazite samples. This mechanism is critically important for monazite geo- and thermochronology and to design and predictively model the long-term behavior of ceramic matrices for nuclear waste conditioning.

36 MATERIALS SCIENCE↗
DOE OSTI1993

Nuclear Safety [Vol. 34, No. 1, January-March 1993]

Nuclear Safety is a review journal that covers significant developments in the field of nuclear safety. Its scope includes the analysis and control of hazards associated with nuclear energy, operations involving fissionable materials, and the products of nuclear fission and their effects on the environment. Primary emphasis is on safety in reactor design, construction, and operation; however, the safety aspects of the entire fuel cycle, including fuel fabrication, spent-fuel processing, nuclear waste disposal, handling of radioisotopes, and environmental effects of these operations, are also treated. Table of Contents for this issue follows. GENERAL SAFETY CONSIDERATIONS: 1 An Integrated Regional Approach To Risk Management of Industrial Systems, S. Chakraborty and R. Stratton; 9 Annual Technical Meeting of the NRC Cooperative Severe Accident Research Program, E. G. Silver; ACCIDENT ANALYSIS: 13 Deterministic Severe Accident Criteria as Severe Accident Design Criteria and Policy for the New Production Reactor-Heavy Water Reactor, P. T. Rhoads; 20 Proposed Deterministic Severe Accident Criteria for the Heavy Water Reactor-New Production Reactor Containment, K. D. Bergeron, S. E. Slezak, and C. E. Leach; 33 Quasi-Static Core Liquid Level Depression and Long-Term Core Uncovery During a PWR LOCA, Y. Kukita, R. R. Schultz, H. Nakamura, and J. Katayama; 49 LIRA: An Advanced Containment System to Minimize the Accidental Radioactivity Releases, A. Turricchia; 63 Errata to “A Review of Hydrogen Production During Melt/Water Interaction in LWRs,” Vol. 33, No. 4; CONTROL AND INSTRUMENTATION: 64 The Nuplex 80+™ Advanced Control Complex from ABB Combustion Engineering, F. Ridolfo, D. Harmon, and K. Scarola; DESIGN FEATURES: 76 A Scheme for Passive Isolation of the Containment of a Reactor, A. K. Ghosh, V. V. Raj, and K. Kakodkar; 84 Utility Requirements for Safety in the Passive Advanced Light-Water Reactor, T. U. Marston, W. H. Layman, and G. Bockhold, Jr.; WASTE AND SPENT FUEL MANAGEMENT: 91 Activities Related to Waste and Spent Fuel Management, Compiled by M. D. Muhlheim and E. G. Silver; OPERATING EXPERIENCES: 103 Testing Deficiencies in Auxiliary Systems Feedwater Systems, J. D. Kueck; 110 Reactor Shutdown Experience, Compiled by J. W. Cletcher; 113 Selected Safety-Related Events, Compiled by G. A. Murphy; 115 Operating U.S. Power Reactors, Compiled by M. D. Muhlheim and E. G. Silver; RECENT DEVELOPMENTS: 133 General Administrative Activities, Compiled by M. D. Muhlheim and E. G. Silver; 145 Reports, Standards, and Safety Guides, D. S. Queener; 150 Proposed Rule Changes as of Sept. 30, 1992; ANNOUNCEMENTS: 75 Harvard Short Course on Risk: Science, Assessment, and Management; 102 International Conference on Reactor Physics and Reactor Computations; 102 The Fourth International Conference on Nuclear Waste Reprocessing and Waste Management [RECOD ’94]; 132 International Symposium on the Radiological Exposure of the Population of the European Community from Radioactivity in the Mediterranean Sea [MARINA-MED]; 162 International High-Level Radioactive Waste Conference ‘Technology Integration—Above and Below”; 154 The Authors; 158 Indexes to Nuclear Safety, Volume 33.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

High Sodium Simulant Testing To Support SB8 Sludge Preparation

Scoping studies were completed for high sodium simulant SRAT/SME cycles to determine any impact to CPC processing. Two SRAT/SME cycles were performed with simulant having sodium supernate concentration of 1.9M at 130% and 100% of the Koopman Minimum Acid requirement. Both of these failed to meet DWPF processing objectives related to nitrite destruction and hydrogen generation. Another set of SRAT/SME cycles were performed with simulant having a sodium supernate concentration of 1.6M at 130%, 125%, 110%, and 100% of the Koopman Minimum Acid requirement. Only the run at 110% met DWPF processing objectives. Neither simulant had a stoichiometric factor window of 30% between nitrite destruction and excessive hydrogen generation. Based on the 2M-110 results it was anticipated that the 2.5M stoichiometric window for processing would likely be smaller than from 110-130%, since it appeared that it would be necessary to increase the KMA factor by at least 10% above the minimum calculated requirement to achieve nitrite destruction due to the high oxalate content. The 2.5M-130 run exceeded the DWPF hydrogen limits in both the SRAT and SME cycle. Therefore, testing of this wash endpoint was halted. This wash endpoint with this minimum acid requirement and mercury-noble metal concentration profile appears to be something DWPF should not process due to an overly narrow window of stoichiometry. The 2M case was potentially processable in DWPF, but modifications would likely be needed in DWPF such as occasionally accepting SRAT batches with undestroyed nitrite for further acid addition and reprocessing, running near the bottom of the as yet ill-defined window of allowable stoichiometric factors, potentially extending the SRAT cycle to burn off unreacted formic acid before transferring to the SME cycle, and eliminating formic acid additions in the frit slurry.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2012

MSC – DFEM Model for the Sintering of Ceramic Nuclear Fuels

Ceramic nuclear fuel pellets represent the fuel of choice for most of the civilian nuclear reactors worldwide. Therefore, it is reasonable to expect the use of ceramic uranium-oxide pellets will continue for many years to come, because an extensive infrastructure is in place to support this strategic component of the energy industry, the technology is well-established and profitable, and the manufacturing process is flexible with the potential to support fuel reprocessing within the existing infrastructure. A better understanding of the properties of the starting materials, improving the control of the processes that are used to fabricate fuel pellets and understanding how the properties of these pellets change in service, represent important aspects that are being studied via experiments, models and simulations. A close integration of these approaches is essential if we are to find new ways to optimize both the fuel composition and structure for the purpose of improving performance and reducing costs.

Material Science↗
DOE OSTI2012

The Aqueous Corrosion Behavior of Technetium – Iron Alloy Materials

If reprocessed, one by-product of spent nuclear fuel is technetium. Metal waste forms are currently being considered for the long-term storage of technetium. Technetium may be alloyed with stainless steel in order to enhance processing. As an initial step in understanding the corrosion behavior of these waste forms, the corrosion of binary technetium alloys has been studied. Technetium was alloyed with iron across a broad range of compositions. It was found that both chemical and microstructural effects play a role in corrosion behavior with corrosion minimized at intermediate technetium concentrations. Preferential dissolution of iron was found to occur at the extremes of the concentration ranges.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Corrosion and Stress Corrosion Cracking: Recommendations for Mitigation and Advanced Detection

Corrosion of stainless steel used fuel canisters in dry storage could, under certain conditions, lead to breaches of the canisters, possibly resulting in releases to the environment and introduction of moisture and oxygen into the canister, enabling possible cladding failures. With the possibility that dry cask storage may be extended to several hundred years before used fuel is disposed of in a geologic repository or reprocessed, the need to look at corrosion of the canisters containing the used fuel becomes more important. This increased importance is due to (a) corrosion mechanisms which are time dependent, with corrosion progression (not necessarily the rate) increasing as a function of time, (b) some corrosion mechanisms such as stress corrosion cracking may exhibit an incubation period, and (c) there will simply be more canisters subjected to corrosion increasing the likelihood of observing outliers. With the possibility that during long term storage corrosion could lead to dry cask canister breach analysis is warranted to determine which corrosion mechanisms are possible and what their likelihood and consequences are, and what possible mitigation and early detection methods are available and warrant further study.

Engineering(42)↗
DOE OSTI2012

LWR First Recycle of TRU with Thorium Oxide for Transmutation and Cross Sections

Thorium has been considered as an option to uranium-based fuel, based on considerations of resource utilization (thorium is approximately three times more plentiful than uranium) and as a result of concerns about proliferation and waste management (e.g. reduced production of plutonium, etc.). Since the average composition of natural Thorium is dominated (100%) by the fertile isotope Th-232, Thorium is only useful as a resource for breeding new fissile materials, in this case U-233. Consequently a certain amount of fissile material must be present at the start-up of the reactor in order to guarantee its operation. The thorium fuel can be used in both once-through and recycle options, and in both fast and thermal spectrum systems. The present study has been aimed by the necessity of investigating the option of using reprocessed plutonium/TRU, from a once-through reference LEU scenario (50 GWd/ tIHM), mixed with natural thorium and the need of collect data (mass fractions, cross-sections etc.) for this particular fuel cycle scenario. As previously pointed out, the fissile plutonium is needed to guarantee the operation of the reactor.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Zone Freezing Study for Pyrochemical Process Waste Minimization

Pyroprocessing technology is a non-aqueous separation process for treatment of used nuclear fuel. At the heart of pyroprocessing lies the electrorefiner, which electrochemically dissolves uranium from the used fuel at the anode and deposits it onto a cathode. During this operation, sodium, transuranics, and fission product chlorides accumulate in the electrolyte salt (LiCl-KCl). These contaminates change the characteristics of the salt overtime and as a result, large volumes of contaminated salt are being removed, reprocessed and stored as radioactive waste. To reduce the storage volumes and improve recycling process for cost minimization, a salt purification method called zone freezing has been proposed at Korea Atomic Energy Research Institute (KAERI).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Preliminary Neutronics Design and Analysis of D 2 O Cooled High Conversion PWRs

This report presents a neutronics analysis of tight-pitch D 2 O-cooled PWRs loaded with MOX fuel and focuses essentially on the Pu breeding potential of such reactors as well as on an important safety parameter, the void coefficient, which has to be negative. It is well known that fast reactors have a better neutron economy and are better suited than thermal reactors to breed fissile material from neutron capture in fertile material. Such fast reactors (e.g. sodium-cooled reactors) usually rely on technologies that are very different from those of existing water-cooled reactors and are probably more expensive. This report investigates another possibility to obtain a fast neutron reactor while still relying mostly on a PWR technology by: (1) Tightening the lattice pitch to reduce the water-to-fuel volume ratio compared to that of a standard PWR. Water-to-fuel volume ratios of between 0.45 and 1 have been considered in this study while a value of about 2 is typical of standard PWRs, (2) Using D 2 O instead of H 2 O as a coolant. Indeed, because of its different neutron physics properties, the use of D 2 O hardens the neutron spectrum to an extent impossible with H 2 O when used in a tight-pitch lattice. The neutron spectra thus obtained are not as fast as those in sodium-cooled reactor but they can still be characterized as fast compared to that of standard PWR neutron spectra. In the phase space investigated in this study we did not find any configurations that would have, at the same time, a positive Pu mass balance (more Pu at the end than at the beginning of the irradiation) and a negative void coefficient. At this stage, the use of radial blankets has only been briefly addressed whereas the impact of axial blankets has been well defined. For example, with a D 2 O-to-fuel volume ratio of 0.45 and a core driver height of about 60 cm, the fissile Pu mass balance between the fresh fuel and the irradiated fuel (50 GWd/t) would be about -7.5% (i.e. there are 7.5% fewer fissile Pu isotopes at the end than at the beginning of the irradiation) and the void coefficient would be negative. The addition of 1 cm of U-238 blanket at the top and bottom of the fuel would bring the fissile Pu mass balance from -7.5% to -6.5% but would also impact the void coefficient in the wrong way. In fact, it turns out that the void coefficient is so sensitive to the presence of axial blanket that it limits its size to only a few cm for driver fuel height of about 50-60 cm. For reference, the fissile Pu mass balance is about -35% in a standard PWR MOX fuel such as those used in France. In order to reduce the fissile Pu deficit and potentially reach a true breeding regime (i.e. a positive Pu mass balance), it would be necessary to make extensive use of radial blankets, both internal and external. Even though this was not addressed in detail here, it is reasonable to believe that at least as much U-238 blanket subassemblies as MOX driver fuel subassemblies would be necessary to breed enough Pu to compensate for the Pu deficit in the driver fuel. Hence, whereas a relatively simple D 2 O-cooled PWR core design makes it possible to obtain a near-breeder core, it may be necessary to more than double the mass of heavy metal in the core as well as the mass of heavy metal to reprocess per unit of energy produced in order to breed the few percents of Pu missing to reach a true breeding regime. It may be interesting to quantify these aspects further in the future.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

OSPREY Model

The absence of industrial scale nuclear fuel reprocessing in the U.S. has precluded the necessary driver for developing the advanced simulation capability now prevalent in so many other countries. Thus, it is essential to model complex series of unit operations to simulate, understand, and predict inherent transient behavior and feedback loops. A capability of accurately simulating the dynamic behavior of advanced fuel cycle separation processes will provide substantial cost savings and many technical benefits. The specific fuel cycle separation process discussed in this report is the off-gas treatment system. The off-gas separation consists of a series of scrubbers and adsorption beds to capture constituents of interest. Dynamic models are being developed to simulate each unit operation involved so each unit operation can be used as a stand-alone model and in series with multiple others. Currently, an adsorption model has been developed within Multi-physics Object Oriented Simulation Environment (MOOSE) developed at the Idaho National Laboratory (INL). Off-gas Separation and REcoverY (OSPREY) models the adsorption of off-gas constituents for dispersed plug flow in a packed bed under non-isothermal and non-isobaric conditions. Inputs to the model include gas, sorbent, and column properties, equilibrium and kinetic data, and inlet conditions. The simulation outputs component concentrations along the column length as a function of time from which breakthrough data is obtained. The breakthrough data can be used to determine bed capacity, which in turn can be used to size columns. It also outputs temperature along the column length as a function of time and pressure drop along the column length. Experimental data and parameters were input into the adsorption model to develop models specific for krypton adsorption. The same can be done for iodine, xenon, and tritium. The model will be validated with experimental breakthrough curves. Customers will be given access to OSPREY to used and evaluate the model.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Selective Extraction of Uranium from Liquid or Supercritical Carbon Dioxide

Current liquid-liquid extraction processes used in recycling irradiated nuclear fuel rely on (1) strong nitric acid to dissolve uranium oxide fuel, and (2) the use of aliphatic hydrocarbons as a diluent in formulating the solvent used to extract uranium. The nitric acid dissolution process is not selective. It dissolves virtually the entire fuel meat which complicates the uranium extraction process. In addition, a solvent washing process is used to remove TBP degradation products, which adds complexity to the recycling plant and increases the overall plant footprint and cost. A liquid or supercritical carbon dioxide (l/sc-CO 2 ) system was designed to mitigate these problems. Indeed, TBP nitric acid complexes are highly soluble in l/sc-CO 2 and are capable of extracting uranium directly from UO 2 , UO 3 and U 3 O 8 powders. This eliminates the need for total acid dissolution of the irradiated fuel. Furthermore, since CO 2 is easily recycled by evaporation at room temperature and pressure, it eliminates the complex solvent washing process. In this report, we demonstrate: (1) A reprocessing scheme starting with the selective extraction of uranium from solid uranium oxides into a TBP-HNO 3 loaded Sc-CO 2 phase, (2) Back extraction of uranium into an aqueous phase, and (3) Conversion of recovered purified uranium into uranium oxide. The purified uranium product from step 3 can be disposed of as low level waste, or mixed with enriched uranium for use in a reactor for another fuel cycle. After an introduction on the concept and properties of supercritical fluids, we first report the characterization of the different oxides used for this project. Our extraction system and our online monitoring capability using UV-Vis absorbance spectroscopy directly in sc-CO 2 is then presented. Next, the uranium extraction efficiencies and kinetics is demonstrated for different oxides and under different physical and chemical conditions: l/sc-CO 2 pressure and temperature, TBP/HNO 3 complex used, reductant or complexant used for selectivity, and ionic liquids used as supportive media. To complete the extraction and recovery cycle, we then demonstrate uranium back extraction from the TBP loaded sc-CO 2 phase into an aqueous phase and the characterization of the uranium complex formed at the end of this process. Another aspect of this project was to limit proliferation risks by either co-extracting uranium and plutonium, or by leaving plutonium behind by selectively extracting uranium. We report that the former is easily achieved, since plutonium is in the tetravalent or hexavalent oxidation state in the oxidizing environment created by the TBP-nitric acid complex, and is therefore co-extracted. The latter is more challenging, as a reductant or complexant to plutonium has to be used to selectively extract uranium. After undertaking experiments on different reducing or complexing systems (e.g., AcetoHydroxamic Acid (AHA), Fe(II), ascorbic acid), oxalic acid was chosen as it can complex tetravalent actinides (Pu, Np, Th) in the aqueous phase while allowing the extraction of hexavalent uranium in the sc-CO 2 phase. Finally, we show results using an alternative media to commonly used aqueous phases: ionic liquids. We show the dissolution of uranium in ionic liquids and its extraction using sc-CO 2 with and without the presence of AHA. The possible separation of trivalent actinides from uranium is also demonstrated in ionic liquids using neodymium as a surrogate and diglycolamides as the extractant.

carbone dioxide↗
DOE OSTI2013

Nuclear Energy Return on Energy Investment

This report provides a methodology and requisite data to assess the potential Energy Return on (Energy) Investment (EROI) for nuclear fuel cycle alternatives, and applies that methodology to an example “once-through” fuel cycle using low enrichment uranium (LEU) in conventional light water cooled reactors (LWRs). Within the DOE Office of Nuclear Energy Science and Technology - Fuel Cycle Technologies program, the Fuel Cycle Options Campaign “performs integrating analyses of nuclear energy and fuel cycle systems to inform fuel cycle R&D, programmatic decisions, strategy formulation, and policy development”. Campaign objectives include development of relevant fuel cycle metrics, and development of tools and associated data for analysis of fuel cycle systems. This study represents an extension of a prior evaluation of EROI as a metric for fuel cycle facilities, processes and technologies. That prior study [Simon 2011] addressed the energy return on the addition of fuel recycle to an existing nuclear energy system. Limited to just the addition of fuel recycle, that study did not include all the energy investments required to create, operate and decommission the underlying nuclear fuel cycle, such as uranium mining, fuel fabrication, reactor construction, and used fuel disposition. This extension of the prior work adds these remaining pieces of the fuel cycle to provide a basic evaluation framework and initial data to enable evaluation of EROI for nuclear energy in general. It is intended as a basis for evaluation of alternative fuel cycle options in the future with the addition of pertinent details for other fuel cycle scenarios. In the prior work, a spreadsheet tool was constructed for the specific purpose of this energy input/output analysis of nuclear fuel cycle facilities. This tool allows the user to enter the parameters of the nuclear fuel cycle and assumptions about energy use in reprocessing. In the current analysis, energy consumption for an entire nuclear energy enterprise is considered, including facility construction, materials, facility operation, and decommissioning. The analysis tool has been extended to include the entire fuel cycle for a representative scenario of a once-through LWR nuclear energy system. Energy content data have been developed for an initial representation. In many cases where there is a valid range of plausible energy content, the range is discussed and a representative value selected for demonstration of the evaluation methodology. In this analysis, energy produced is the energy output of power reactors within the fuel cycle. The spreadsheet calculates both Primary EROI and Final EROI for the fuel cycle entered. The EROI is the ratio of the output energy divided by the consumed energy. The intent for this study was to develop the methodology and analysis tool for a complete fuel cycle. It is not the intent of the study to fully explore the wide range of potential energy intensities, or to reconcile the disparate values found in the literature. Representative numbers were used in this demonstration. The methodology and tool do provide a framework for future exploration of the key energy intensity values and for conducting sensitivity studies on specific values, either to assess improved understanding of the values, or to explore the potential for alternative technologies to impact EROI. The Primary and Final Energy EROI values calculated for the representative scenario were ~ 52 and 24, respectively.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗