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Data Package for Groundwater Monitoring Well 299-W19-46 at the 200-UP-1 Operable Unit

Description: One new groundwater monitoring well was constructed at the 200-UP-1 Operable Unit in November 2002. This document provides the information on drilling and construction. One new groundwater monitoring well was constructed in the 200-UP-1 Operable Unit in November 2002. The purpose of the well is to monitor the concentration of technetium-99 and uranium in groundwater associated with the pump-and-treat systems in the 200-UP-1 Operable Unit. The well name is 299-W19-46 and the corresponding well number is C3958. Well 299-W19-46 is located off the southwest corner of the 216-U-17 crib and east of Beloit Avenue. It is a replacement well for well 299-W19-38, which has gone dry. The location of the well is shown on Figure 1. Well 299-W19-46 was drilled in response to the recommendations of a data quality objectives process that indicated a need for additional monitoring wells in the area (BHI-01576). The new well was constructed to the specifications and requirements described in Washington Administrative Code (WAC) 173-160 and WAC 173-303, the Data Quality Objectives document (BHI-01576), and the description of work for well drilling and construction (Fruchter 2002). This document compiles information on the drilling and construction, geophysical logging, and sediment and groundwater sampling applicable to the installation of well 299-W19-46. The information on drilling and construction, well development, and pump installation is summarized from CP-14265. Appendix A contains the Well Summary Sheets (as-built diagrams), the Well Construction Summary Reports, and the geologist's logs; Appendix B contains results of physical properties testing; Appendix C contains the analytical results from groundwater samples obtained during drilling; and Appendix D contains borehole geophysical logs. Additional documentation concerning well construction can be found in CP-14265 and is on file with Fluor Hanford, Inc., Richland, WA.
Date: April 17, 2003
Creator: Horton, Duane G.
Partner: UNT Libraries Government Documents Department

Summary of the Hanford Site Environmental Report for Calendar Year 2002

Description: This summary booklet is designed to briefly (1) describe the Hanford Site and its mission; (2) describe environmental programs at the Hanford Site; (3) discuss estimated radionuclide exposures to the public from 2002 Hanford Site activities; (4) summarize the status of compliance with environmental regulations; and (5) present information on environmental monitoring and surveillance and groundwater protection and monitoring.
Date: September 26, 2003
Creator: Hanf, Robert W.; Morasch, Launa F.; O'Connor, Georganne P. & Poston, Ted M.
Partner: UNT Libraries Government Documents Department

Summary of Hanford Site Groundwater Monitoring for Fiscal Year 2001

Description: This booklet summarizes a more detailed report, Hanford Site Groundwater Monitoring for Fiscal Year 2001. This summary booklet is designed to briefly (1) describe the highlights for fiscal year 2001; (2) identify emerging issued in groundwater monitoring; (3) discuss groundwater flow and movement; and (4) provide an overview of current contamination in the Hanford Site groundwater and vadose zone.
Date: October 1, 2002
Creator: Hartman, Mary J.; Morasch, Launa F. & Webber, William D.
Partner: UNT Libraries Government Documents Department

From the Lab to the real world : sources of error in UF {sub 6} gas enrichment monitoring

Description: Safeguarding uranium enrichment facilities is a serious concern for the International Atomic Energy Agency (IAEA). Safeguards methods have changed over the years, most recently switching to an improved safeguards model that calls for new technologies to help keep up with the increasing size and complexity of today’s gas centrifuge enrichment plants (GCEPs). One of the primary goals of the IAEA is to detect the production of uranium at levels greater than those an enrichment facility may have declared. In order to accomplish this goal, new enrichment monitors need to be as accurate as possible. This dissertation will look at the Advanced Enrichment Monitor (AEM), a new enrichment monitor designed at Los Alamos National Laboratory. Specifically explored are various factors that could potentially contribute to errors in a final enrichment determination delivered by the AEM. There are many factors that can cause errors in the determination of uranium hexafluoride (UF{sub 6}) gas enrichment, especially during the period when the enrichment is being measured in an operating GCEP. To measure enrichment using the AEM, a passive 186-keV (kiloelectronvolt) measurement is used to determine the {sup 235}U content in the gas, and a transmission measurement or a gas pressure reading is used to determine the total uranium content. A transmission spectrum is generated using an x-ray tube and a “notch” filter. In this dissertation, changes that could occur in the detection efficiency and the transmission errors that could result from variations in pipe-wall thickness will be explored. Additional factors that could contribute to errors in enrichment measurement will also be examined, including changes in the gas pressure, ambient and UF{sub 6} temperature, instrumental errors, and the effects of uranium deposits on the inside of the pipe walls will be considered. The sensitivity of the enrichment calculation to these various parameters will then be …
Date: March 1, 2012
Creator: Lombardi, Marcie L.
Partner: UNT Libraries Government Documents Department

Monthly Technical Progress Report

Description: Progress is reported on work on carbon reduction of uranium oxide; uranium slug and slug canning development; preparation and handling of fine non-pyrophoric uranium powder; separation of alloys; reduction of thorium oxide; and preparation of beryllium. Work to investigate and develop methods, by means of which flat plate fuel elements approximately 14 feet long can be manufactured economically by powder metallurgical processes and to investigate methods for producing tubular fuel elementsis reported. Progress is reported on work on hydrostatic pressing of metal powders and slip casting of metal powders. Further development work is reported on dimensionally stable uranium alloys, wire fuel elements, and perforated wafer fuel elements, as well as investigation of the mechanism of dimensional instability of U under irradiation and the fundamentals of sintering and of diffusional bonding.
Date: September 13, 1955
Partner: UNT Libraries Government Documents Department

Experimental Verification of a Cracked Fuel Mechanical Model

Description: This report describes the results of a series of laboratory experiments conducted to independently verify a model that describes the nonlinear mechanical behavior of cracked fuel in pelletized UO{sub 2}/Zircaloy nuclear fuel rods under normal operating conditions. After a brief description of the analytical model, each experiment is discussed in detail. Experiments were conducted to verify the general behavior and numerical values for the three primary independent modelling parameters (effective crack roughness, effective gap roughness, and total crack length), and to verify the model predictions that the effective Young's moduli for cracked fuel systems were substantially less than those for solid UO{sub 2} pellets. In general, the model parameters and predictions were confirmed, and new insight was gained concerning the complexities of cracked fuel mechanics.
Date: December 1, 1982
Creator: Williford, R. E.
Partner: UNT Libraries Government Documents Department

Reactivity Initiated Accident Test Series RIA Scoping Test Quick Look Report

Description: The Reactivity Initiated Accident Scoping Test (RIA-ST) was successfully completed August 30, 1978. The test was introductory to the RIA Series 1 tests and was designed to investigate and resolve several anticipated problem areas prior to performance of the first test of the series, Test RIA 1-1. The RIA Scoping Test, as performed, consisted of four separate single-rod experiment phases. The first three phases were performed with shrouded fuel rods of 5.8 wt.% enrichment. They were subjected to power bursts resulting in total fuel surface energies ranging from 205 to 261 cal/q at the axial peak elevation. The fourth phase consisted of a 20 wt.% enriched, shrouded fuel rod which was subjected to a power hurst that deposited a total radially averaged energy of 527 cal/g. The primary objectives of the Scoping Test were defined as follows: (1) Determine the applicability of extrapolating low-power steady state calorimetric measurements and self-powered neutron detector (SPND) output to determine fuel rod energy depositions during a power burst. (2) Determine the enerqy deposition failure threshold for unirradiated fuel rods at BWR hot-startup coolant conditions. (3) Determine the magnitudes of oossible pressure pulses resulting from rod failure. (4) Determine the sensitivity of the test instrumentation to high transient radiation exposures. In general, the energy deposition values for the Scoping Test derived from the SPND output were 25% higher than those obtained from the core ion chamber data. Determining which values are correct will require radiochemical analysis of the fuel rods which will take several months. At present, it apoears that the SPND derived energies are in error because of excellent agreement between the calculated and measured power calibration results and the agreement between the predicted failure threshold and that seen using the core ion chamber derived energies. Meeting the second objective was accomplished during the …
Date: September 1978
Creator: Martinson, Z. R.; Semken, R. S.; Inabe, T.; Smith, R. H.; Cook, T. F. & Appelhans, A. D.
Partner: UNT Libraries Government Documents Department

ION EXCHANGE PERFORMANCE OF TITANOSILICATES, GERMANATES AND CARBON NANOTUBES

Description: This report presents a summary of testing the affinity of titanosilicates (TSP), germanium-substituted titanosilicates (Ge-TSP) and multiwall carbon nanotubes (MWCNT) for lanthanide ions in dilute nitric acid solution. The K-TSP ion exchanger exhibited the highest affinity for lanthanides in dilute nitric acid solutions. The Ge-TSP ion exchanger shows promise as a material with high affinity, but additional tests are needed to confirm the preliminary results. The MWCNT exhibited much lower affinities than the K-TSP in dilute nitric acid solutions. However, the MWCNT are much more chemically stable to concentrated nitric acid solutions and, therefore, may candidates for ion exchange in more concentrated nitric acid solutions. This technical report serves as the deliverable documenting completion of the FY13 research milestone, M4FT-13SR0303061 – measure actinide and lanthanide distribution values in nitric acid solutions with sodium and potassium titanosilicate materials.
Date: April 24, 2013
Creator: Alsobrook, A. & Hobbs, D.
Partner: UNT Libraries Government Documents Department

Multilaboratory analytical quality control for the hydrochemical and stream sediment reconnaissance

Description: For the first time data received from LLL has been incorporated in the quality assurance report. LASL has indicated that their results on the water standard B2 have been consistently low. They suspect that this may be resulting from uranium plating out in the polyethylene containers. Such an observation has not been made by either ORGDP, SRL or LLL. To evaluate these suspect observations LASL has been sent two sets of standards, one set contained in teflon and the second in the usual polyethylene containers. LASL results for June will be carefully evaluated.
Date: May 31, 1978
Creator: D'Silva, A. P.; Haas, W. J., Jr. & Floyd, M. A.
Partner: UNT Libraries Government Documents Department

Criticality Experiments with Mixed Plutonium and Uranium Nitrate Solution at a Plutonium Fraction of 0.4 in Slab and Cylindrical Geometry

Description: A series of critical experiments was completed with mixed plutonium-uranium solutions having Pu/(Pu + U) ratios of approximately 0.4. These experiments were a part of the Criticality Data Development Program between the United States Department of Energy (USDOE), and the Power Reactor and Nuclear Fuel Development Corporation (PNC) of Japan. A complete description of, and data from, the experiments are included in this report. The experiments were performed with mixed plutonium-uranium solutions in cylinqrical and slab geometries and included measurements with a water reflector, a concrete reflector, and without an added reflector. The concentration was varied from 105 to 436 g (Pu + U)/liter. The ratio of plutonium to total heavy metal (plutonium plus uranium) was 0.4 for all experiments.
Date: April 1, 1988
Creator: Lloyd, RC
Partner: UNT Libraries Government Documents Department

Criticality Experiments with Subcritical Clusters of 2.35 Wt% and 4.31 Wt% 235U Enriched U02 Rods in Water with Uranium or Lead Reflecting Walls Undermoderated Water-to-Fuel Volume Ratio of 1.6

Description: A series of criticality experiments with undermoderated (1.6 water-to-fuel volume ratio) 2.35 wt% and 4.31 wt% {sup 235}U enriched UO{sub 2} rods in water were performed to provide data on the reactivity effects of lead and depleted uranium reflecting walls. This data furnishes well defined benchmarks for use in validating calculational techniques employed in analyzing fuel shipping and storage systems having lead or uranium biological shields. For each fuel enrichment, the critical separation between three subcritical fuel clusters was observed to increase as either 77mm thick depleted uranium or 102mm thick lead reflecting walls were moved towards the fuel. A maximum critical separation was observed for both the lead and the depleted uranium reflecting walls with a water gap between the fuel clusters and the reflecting walls. For both fuel enrichments, this optimum water gap was about 25mm for the depleted uranium walls and about lOmm for the lead walls.
Date: December 1, 1981
Creator: Bierman, S. R.; Durst, B. M. & Clayton, E. D.
Partner: UNT Libraries Government Documents Department

Radon Diffusion Through Uranium Mill Tailings and Cover Defects

Description: Research was conducted at Pacific Northwest Laboratory to define the effects of cover defects on the emission of radon gas from covered uranium mill tailings piles. This report describes the results from the analysis of four geometrically simplified cover defects.
Date: December 1, 1981
Creator: Mayer, D. W. & Zimmerman, D. A.
Partner: UNT Libraries Government Documents Department

Environmental Assessment Methodology for the Nuclear Fuel Cycle

Description: This report describes the methodology for determining where environmental control technology is required for the nuclear fuel cycle. The methodology addresses routine emission of chemical and radioactive effluents, and applies to mining, milling, conversion, enrichment, fuel fabrication, reactors (LWR and BWR) and fuel reprocessing. Chemical and radioactive effluents are evaluated independently. Radioactive effluents are evaluated on the basis of maximum exposed individual dose and population dose calculations for a 1-year emission period and a 50-year commitment. Sources of radionuclides for each facility are then listed according to their relative contribution to the total calculated dose. Effluent, ambient and toxicology standards are used to evaluate the effect of chemical effluents. First, each chemical and source configuration is determined. Sources are tagged if they exceed existirrg standards. The combined effect of all chemicals is assessed for each facility. If the additive effects are unacceptable, then additional control technology is recommended. Finally, sources and their chemicals at each facility are ranked according to their relative contribution to the ambient pollution level. This ranking identifies those sources most in need of environmental control.
Date: July 1, 1977
Creator: Brenchley, D. L.; Soldat, J. K.; McNeese, J. A. & Watson, E. C.
Partner: UNT Libraries Government Documents Department

SAVANNAH RIVER SITE'S H-CANYON FACILITY: RECOVERY AND DOWN BLEND URANIUM FOR BENEFICIAL USE

Description: For over fifty years, the H Canyon facility at the Savannah River Site (SRS) has performed remotely operated radiochemical separations of irradiated targets to produce materials for national defense. Although the materials production mission has ended, the facility continues to play an important role in the stabilization and safe disposition of proliferable nuclear materials. As part of the US HEU Disposition Program, SRS has been down blending off-specification (off-spec) HEU to produce LEU since 2003. Off-spec HEU contains fission products not amenable to meeting the American Society for Testing and Material (ASTM) commercial fuel standards prior to purification. This down blended HEU material produced 301 MT of ~5% enriched LEU which has been fabricated into light water reactor fuel being utilized in Tennessee Valley Authority (TVA) reactors in Tennessee and Alabama producing economic power. There is still in excess of ~10 MT of off-spec HEU throughout the DOE complex or future foreign and domestic research reactor returns that could be recovered and down blended for beneficial use as either ~5% enriched LEU, or for use in subsequent LEU reactors requiring ~19.75% enriched LEU fuel.
Date: May 27, 2013
Creator: Magoulas, V.
Partner: UNT Libraries Government Documents Department

Preliminary Outline for Book: Engineering for Nuclear Reactor Fuel Reprocessing

Description: This document outlines a book on the subject of reactor fuel reprocessing that is still in the planning stages, representing the authors' thinking as of the arbitrary cut-off date of October 15, 1957. The subject matter that was intended for inclusion was: special considerations in radiochemical processing; chemical processes and operations; mechanical operations; fluid flow; heat transfer operations; solvent extraction; other mass diffusion operations; instrumentation; auxiliary equipment; plant design and operation; and fuel processing economics.
Date: November 15, 1957
Creator: Long, J.T.; Carter, W.L. & Rom, A.M.
Partner: UNT Libraries Government Documents Department

Criticality Experiments with Mixed Plutonium and Uranium Nitrate Solution at a Plutonium Fraction of 0.5 in Annular Cylindrical Geometry

Description: A series of critical experiments was completed with mixed plutonium-uranium solutions having Pu/(Pu + U) ratios of approximately 0.5. These experiments were a part of the Criticality Data Development Program between the United States Department of Energy (USDOE), and the Power Reactor and Nuclear Fuel Development Corporation (PNC) of Japan. A complete description of, and data from, the experiments are included in this report. The experiments were performed with mixed plutonium-uranium solutions in annular cylindrical geometry. The measurements were made with a water reflector. The central region included a concrete annular cylinder containing B{sub 4}C. Interior to the concrete insert was a stainless steel bottle containing plutonium-uranium solution. The concentration of the solution in the annular region was varied from 116 to 433 g (Pu + U)/liter. The ratio of plutonium to total heavy metal (plutonium plus uranium) was 52% for all experiments.
Date: April 1, 1988
Creator: Lloyd, RC
Partner: UNT Libraries Government Documents Department

Solubility Classification of Airborne Uranium Products from LWR-Fuel Plants

Description: Airborne dust samples were obtained from various locations within plants manufacturing fuel elements for light-water reactors, and the dissolution rates of uranium from these samples into simulated lung fluid at 37°C were measured. These measurements were used to classify the solubilities of the samples in terms of the lung clearance model proposed by the International Commission on Radiological Protection. Similar evaluations were performed for samples of pure uranium compounds expected as components in plant dust. The variation in solubility classifications of dust encountered along the fuel production lines is described and correlated with the process chemistry and the solubility classifications of the pure uranium compounds.
Date: August 1, 1980
Creator: kalkwarf, D. R.
Partner: UNT Libraries Government Documents Department

Quantify Water Extraction by TBP/Dodecane via Molecular Dynamics Simulations

Description: The purpose of this project is to quantify the interfacial transport of water into the most prevalent nuclear reprocessing solvent extractant mixture, namely tri-butyl- phosphate (TBP) and dodecane, via massively parallel molecular dynamics simulations on the most powerful machines available for open research. Specifically, we will accomplish this objective by evolving the water/TBP/dodecane system up to 1 ms elapsed time, and validate the simulation results by direct comparison with experimentally measured water solubility in the organic phase. The significance of this effort is to demonstrate for the first time that the combination of emerging simulation tools and state-of-the-art supercomputers can provide quantitative information on par to experimental measurements for solvent extraction systems of relevance to the nuclear fuel cycle. Results: Initially, the isolated single component, and single phase systems were studied followed by the two-phase, multicomponent counterpart. Specifically, the systems we studied were: pure TBP; pure n-dodecane; TBP/n-dodecane mixture; and the complete extraction system: water-TBP/n-dodecane two phase system to gain deep insight into the water extraction process. We have completely achieved our goal of simulating the molecular extraction of water molecules into the TBP/n-dodecane mixture up to the saturation point, and obtained favorable comparison with experimental data. Many insights into fundamental molecular level processes and physics were obtained from the process. Most importantly, we found that the dipole moment of the extracting agent is crucially important in affecting the interface roughness and the extraction rate of water molecules into the organic phase. In addition, we have identified shortcomings in the existing OPLS-AA force field potential for long-chain alkanes. The significance of this force field is that it is supposed to be optimized for molecular liquid simulations. We found that it failed for dodecane and/or longer chains for this particular solvent extraction application. We have proposed a simple way to circumvent …
Date: May 16, 2013
Creator: Khomami, Bamin; Cui, Shengting; de Almeida, Valmor F. & Felker, Kevin
Partner: UNT Libraries Government Documents Department

RESEARCH AND DEVELOPMENT ACTIVITIES AT SAVANNAH RIVER SITE'S H CANYON FACILITY

Description: The Savannah River Site's (SRS) H Canyon Facility is the only large scale, heavily shielded, nuclear chemical separations plant still in operation in the U.S. The facility's operations historically recovered uranium-235 (U-235) and neptunium-237 (Np-237) from aluminum-clad, enriched-uranium fuel tubes from Site nuclear reactors and other domestic and foreign research reactors. Today the facility, in conjunction with HB Line, is working to provide the initial feed material to the Mixed Oxide Facility also located on SRS. Many additional campaigns are also in the planning process. Furthermore, the facility has started to integrate collaborative research and development (R&D) projects into its schedule. H Canyon can serve as the appropriate testing location for many technologies focused on monitoring the back end of the fuel cycle, due to the nature of the facility and continued operation. H Canyon, in collaboration with the Savannah River National Laboratory (SRNL), has been working with several groups in the DOE complex to conduct testing demonstrations of novel technologies at the facility. The purpose of conducting these demonstrations at H Canyon will be to demonstrate the capabilities of the emerging technologies in an operational environment. This paper will summarize R&D testing activities currently taking place in H Canyon and discuss the possibilities for future collaborations.
Date: July 9, 2013
Creator: Sexton, L. & Fuller, Kenneth
Partner: UNT Libraries Government Documents Department

Aerosols Generated by Free Fall Spills of Powders and Solutions in Static Air

Description: Safety assessments and environmental impact statements for nuclear fuel cycle facilities require an estimation of potential airborne releases. Aerosols generated by accidents are being investigated to develop the source terms for these releases. The lower boundary accidental release event would be a free fall spill of powders or liquids in static air. Experiments measured the mass airborne and particle size distribution of these aerosols for various source sizes and spill heights. Two powder and liquid sources were used: Ti02 and uo2; and aqueous uranine (sodium fluorescein) and uranyl nitrate solutions. Spill height and source size were significant in releases of both powders and liquids. For the source powders used (l "m uo2 and 1.7 "m Ti0 2, quantities from 25 g to 1000 g, and fall heights of 1 m and 3m), the maximum source airborne was 0.12%. The maximum source airborne was an order of magnitude less for the liquids (with source quantities ranging from 125 to 1000 cc at the same fall heights). The median aerodynamic equivalent diameters for collected airborne powder ranged from 6 to 26.5 "m; liquids ranged from 4.1 to 34 "m. All of the spills produced a significant fraction of respirable particles 10 ~m and less.
Date: December 1, 1981
Creator: Sutter, S. L.; Johnston, J. W. & Mishima, J.
Partner: UNT Libraries Government Documents Department

Simulations of the Thermodynamic and Diffusion Properties of Actinide Oxide Fuel Materials

Description: Spent nuclear fuel from commercial reactors is comprised of 95-99 percent UO{sub 2} and 1-5 percent fission products and transuranic elements. Certain actinides and fission products are of particular interest in terms of fuel stability, which affects reprocessing and waste materials. The transuranics found in spent nuclear fuels are Np, Pu, Am, and Cm, some of which have long half- lives (e.g., 2.1 million years for {sup 237}Np). These actinides can be separated and recycled into new fuel matrices, thereby reducing the nuclear waste inventory. Oxides of these actinides are isostructural with UO{sub 2}, and are expected to form solid solutions. This project will use computational techniques to conduct a comprehensive study on thermodynamic properties of actinide-oxide solid solutions. The goals of this project are to: Determine the temperature-dependent mixing properties of actinide-oxide fuels; Validate computational methods by comparing results with experimental results; Expand research scope to complex (ternary and quaternary) mixed actinide oxide fuels. After deriving phase diagrams and the stability of solid solutions as a function of temperature and pressure, the project team will determine whether potential phase separations or ordered phases can actually occur by studying diffusion of cations and the kinetics of potential phase separations or ordered phases. In addition, the team will investigate the diffusion of fission product gases that can also have a significant influence on fuel stability. Once the system has been established for binary solid solutions of Th, U, Np, and Pu oxides, the methodology can be quickly applied to new compositions that apply to ternaries and quaternaries, higher actinides (Am, Cm), burnable poisons (B, Gd, Hf), and fission products (Cs, Sr, Tc) to improve reactivity.
Date: April 16, 2013
Creator: Becker, Udo
Partner: UNT Libraries Government Documents Department

SULFUR HEXAFLUORIDE TREATMENT OF USED NUCLEAR FUEL TO ENHANCE SEPARATIONS

Description: Reactive Gas Recycling (RGR) technology development has been initiated at Savannah River National Laboratory (SRNL), with a stretch-goal to develop a fully dry recycling technology for Used Nuclear Fuel (UNF). This approach is attractive due to the potential of targeted gas-phase treatment steps to reduce footprint and secondary waste volumes associated with separations relying primarily on traditional technologies, so long as the fluorinators employed in the reaction are recycled for use in the reactors or are optimized for conversion of fluorinator reactant. The developed fluorination via SF{sub 6}, similar to the case for other fluorinators such as NF{sub 3}, can be used to address multiple fuel forms and downstream cycles including continued processing for LWR via fluorination or incorporation into a aqueous process (e.g. modified FLUOREX) or for subsequent pyro treatment to be used in advanced gas reactor designs such metal- or gas-cooled reactors. This report details the most recent experimental results on the reaction of SF{sub 6} with various fission product surrogate materials in the form of oxides and metals, including uranium oxides using a high-temperature DTA apparatus capable of temperatures in excess of 1000{deg}C . The experimental results indicate that the majority of the fission products form stable solid fluorides and sulfides, while a subset of the fission products form volatile fluorides such as molybdenum fluoride and niobium fluoride, as predicted thermodynamically. Additional kinetic analysis has been performed on additional fission products. A key result is the verification that SF{sub 6} requires high temperatures for direct fluorination and subsequent volatilization of uranium oxides to UF{sub 6}, and thus is well positioned as a head-end treatment for other separations technologies, such as the volatilization of uranium oxide by NF{sub 3} as reported by colleagues at PNNL, advanced pyrochemical separations or traditional full recycle approaches. Based on current results of …
Date: September 25, 2012
Creator: Gray, J.; Torres, R.; Korinko, P.; Martinez-Rodriguez, M.; Becnel, J.; Garcia-Diaz, B. et al.
Partner: UNT Libraries Government Documents Department
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