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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

Ab Initio Enhanced calphad Modeling of Actinide-Rich Nuclear Fuels

Description: The process of fuel recycling is central to the Advanced Fuel Cycle Initiative (AFCI), where plutonium and the minor actinides (MA) Am, Np, and Cm are extracted from spent fuel and fabricated into new fuel for a fast reactor. Metallic alloys of U-Pu-Zr-MA are leading candidates for fast reactor fuels and are the current basis for fast spectrum metal fuels in a fully recycled closed fuel cycle. Safe and optimal use of these fuels will require knowledge of their multicomponent phase stability and thermodynamics (Gibbs free energies). In additional to their use as nuclear fuels, U-Pu-Zr-MA contain elements and alloy phases that pose fundamental questions about electronic structure and energetics at the forefront of modern many-body electron theory. This project will validate state-of-the-art electronic structure approaches for these alloys and use the resulting energetics to model U-Pu-Zr-MA phase stability. In order to keep the work scope practical, researchers will focus on only U-Pu-Zr-{Np,Am}, leaving Cm for later study. The overall objectives of this project are to: Provide a thermodynamic model for U-Pu-Zr-MA for improving and controlling reactor fuels; and, Develop and validate an ab initio approach for predicting actinide alloy energetics for thermodynamic modeling.
Date: October 28, 2013
Creator: Morgan, Dane & Yang, Yong Austin
Partner: UNT Libraries Government Documents Department

Power Reactor Fuel Reprocessing: Mechanical Phase

Description: The major events in·the.mechanical phase of the Power Reactor fuels reprocessing program during May. were: 1. Detailed design of the equipment necessary for the SRE fuel element reprocessing continues with some items released for fabrication. 2. Decision was reached to plan to start installation of equipment in the segmenting facility July 1. 3. Most SRE fuel element reprocessing equipment will be installed directly in the segmenting facility without prior testing in Building. 4505.
Date: June 1, 1959
Creator: Klima, B. B.
Partner: UNT Libraries Government Documents Department

Power Reactor Fuel Reprocessing: Mechanical Phase

Description: The major events in the mechanical phase of the Power Reactor fuels reprocessing program during June were: 1. Feasibility of shearing of fuel elements without disassembly has been demonstrated in tests using porcelain-loaded prototype fuel elements. 2. Further work with the Manco shear was not deemed tb be advisable since permission has been granted to use another shear for cutting UO{sub 2}-loaded fuel elements. 3. Necessity to strip the windows in Building 3048, to sandblast, and repaint them has seriously disrupted occupancy of the cell by July 1. Start of installation probably will not be before August 1. 4. A cold SRE element should be received during July which will permit a direct look a t the problems associated with processing of these irradiated fuel elements. 5. Concurrence with AEC, Atomics International, and ORNL people on the fabrication of a poisoned carrier was obtained and all criteria for the carrier were released and the design was completed. 6. A decision was made to install and use a 24-inch Ty-Sa-Man saw which is on hand and was originally purchased for use in the Segmenting Facility for the SRE reprocessing. This will be used instead of the multipurpose saw to allow more time to refine the design of that saw. The multipurpose saw will be installed for use in subsequent reprocessing programs. This report will chronicle the changes in status which occurred during the calendar month of June. A complete description of each item is not included and may be found in the parent report. The dates indicated on the schedule have slipped since the last report primarily due to increase in scope of the work and postponement on all phases of the work except for the SRE preparations. Twenty-four new items have been added to the schedule. The status of procurement ...
Date: July 1, 1959
Creator: Klima, B. B.
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

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

Reactivity Initiated Accident Test Series RIA Scoping Test Experiment Predictions

Description: The Reactivity 'Initiated Accident (RIA) test series to be conducted in the Power Burst Facility (PBF) has been designed.to determine fuel failure thresholds, modes, and consequences as a function of energy deposition, irradiation history, and fuel design. The RIA Scoping Test will be comprised of five single unirradiated rod sub-tests. The first rod will be subjected to a series of transient power bursts of increasing energy release to determine the energy deposition at cladding failure. The second and third rods will be subjected to energy depositions near that which caused failure of the first rod, to further define the failure threshold. Rods four and five will be subjected to large radially averaged energy depositions, 1990 and 2510 J/g respectively, to investigate facility safety concerns. Several analyses were performed to predict test fuel rod and system behavior during the five RIA Scoping Test phases. A reactor physics analysis was performed to obtain the relationship between test fuel rod and reactor core energy during a power transient. The calculations were made with the RAFFLE computer code. The thermal-hydraulic behavior of the test rod coolant was investigated for pellet surface energy depositions of 900, 1125, and 1350 J/g for the first three phases of the Scoping Test. The RELAP4 computer code was used for these thermal-hydraulic analyses. The results of the RELAP4 calculations provided input to the FRAP-T4 computer code for three fuel rod behavior analyses at pellet surface energy depositions of 815, 1020, and 1225 J/g. A cladding embrittlement analysis, using the results of the FRAP-T4 calculations as input, was made to investigate the cladding oxidation mode of rod failure for the lower energy phases. BUILD5 was the analytical tool used in this investigation. Finally, the pressure pulses generated as a result of failure of the test fuel rods in the final ...
Date: June 1978
Creator: Semken, R. S.; Eaton, A. M.; Smith, R. H. & Resch, S. C.
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

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

Review of Design Approaches Applicable to Dewatering Uranium Mill Tailings Disposal Pits

Description: This report is a review of design approaches in the literature that may be applicable to uranium mill tailings drainage. Tailings dewatering is required in the deep mined-out pits used for wet tailings disposal. Agricultural drainage theory is reviewed because it is seen as the most applicable technology. It is concluded that the standard drain-pipe envelope design criteria should be easily adapted. The differences in dewatering objectives and physical characteristics between agricultural and tailings drainage systems prevent direct technology transfer with respect to drain spacing calculations. Recommendations for further research are based on the drainage features unique to uranium mill tailings. It is recommended that transient solutions be applied to describe liquid movement through saturated and partially saturated tailings. Modeling should be used to evaluate the benefits of drainage design approaches after careful consideration of potential construction problems.
Date: March 1, 1982
Creator: Gutknecht, P. J. & Gates, T. E.
Partner: UNT Libraries Government Documents Department

Accident Generated Particulate Materials and Their Characteristics -- A Review of Background Information

Description: Safety assessments and environmental impact statements for nuclear fuel cycle facilities require an estimate of the amount of radioactive particulate material initially airborne (source term) during accidents. Pacific Northwest Laboratory (PNL) has surveyed the literature, gathering information on the amount and size of these particles that has been developed from limited experimental work, measurements made from operational accidents, and known aerosol behavior. Information useful for calculating both liquid and powder source terms is compiled in this report. Potential aerosol generating events discussed are spills, resuspension, aerodynamic entrainment, explosions and pressurized releases, comminution, and airborne chemical reactions. A discussion of liquid behavior in sprays, sparging, evaporation, and condensation as applied to accident situations is also included.
Date: May 1, 1982
Creator: Sutter, S. L.
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

Criticality Experiments with Subcritical Clusters of 2.35 Wt% and 4.31 Wt% {sup 235}U Enriched UO{sub 2} Rods in Water with Steel Reflecting Walls

Description: A series of criticality experiments with 2.35 wt% and 4.31 wt% {sup 235}U enriched UO{sub 2} rods in water were performed to provide well defined benchmark type data on the effects of thick steel reflecting walls. For each fuel enrichment. the critical separation between three subcritical fuel clusters was observed to increase as 178.5 mm thick reflecting walls of reactor grade steel was moved towards the fuel. This increase was observed for fuel clusters having an undermoderated water-to-fuel volume ratio of 1.6 and for fuel clusters having near optimum neutron moderation (2.92 for the 2.35 wt% {sup 235}U enriched fuel and 3,88 for the 4.31 wt% {sup 235}U enriched fuel). In all cases the critical separation between fuel clusters increased to a maximum as the steel walls were moved towards the fuel clusters. This maximum effect was observed with about 10 mm of water between the fuel clusters and the steel reflecting walls. As this water gap was decreased, the critical separation between the fuel clusters also decreased slightly. Measurement data were also obtained for each enrichment with neutron absorber plates between the fuel clusters having the l .6 water-to-fuel volume ratio. During these measurements, the steel reflecting walls were at the near optimum distance from the fuel clusters. The fixed neutron absorbers for which data were obtained include 304-L steel, borated 304-L steel, copper, copper containing 1 wt% cadmium, cadmium, and two trade name materials containing boron (Boral and Boroflex), A comparison between these data and data from previous experiments indicates a slight reduction in the effectiveness of the absorber plates when the steel reflecting walls are present.
Date: April 1, 1981
Creator: Bierman, S. R. & Clayton, E. D.
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

Technology, Safety and Costs of Decommissioning a Reference Uranium Hexafluoride Conversion Plant

Description: Safety and cost information is developed for the conceptual decommissioning of a commercial uranium hexafluoride conversion (UF{sub 6}) plant. Two basic decommissioning alternatives are studied to obtain comparisons between cost and safety impacts: DECON, and passive SAFSTOR. A third alternative, DECON of the plant and equipment with stabilization and long-term care of lagoon wastes. is also examined. DECON includes the immediate removal (following plant shutdown) of all radioactivity in excess of unrestricted release levels, with subsequent release of the site for public use. Passive SAFSTOR requires decontamination, preparation, maintenance, and surveillance for a period of time after shutdown, followed by deferred decontamination and unrestricted release. DECON with stabilization and long-term care of lagoon wastes (process wastes generated at the reference plant and stored onsite during plant operation} is also considered as a decommissioning method, although its acceptability has not yet been determined by the NRC. The decommissioning methods assumed for use in each decommissioning alternative are based on state-of-the-art technology. The elapsed time following plant shutdown required to perform the decommissioning work in each alternative is estimated to be: for DECON, 8 months; for passive SAFSTOR, 3 months to prepare the plant for safe storage and 8 months to accomplish deferred decontamination. Planning and preparation for decommissioning prior to plant shutdown is estimated to require about 6 months for either DECON or passive SAFSTOR. Planning and preparation prior to starting deferred decontamination is estimated to require an additional 6 months. OECON with lagoon waste stabilization is estimated to take 6 months for planning and about 8 months to perform the decommissioning work. Decommissioning cost, in 1981 dollars, is estimated to be $5.91 million for OECON. For passive SAFSTOR, preparing the facility for safe storage is estimated to cost $0.88 million, the annual maintenance and surveillance cost is estimated to be ...
Date: October 1, 1981
Creator: Elder, H. K.
Partner: UNT Libraries Government Documents Department

Technical Support for Improving the Licensing Regulatory Base for Selected Facilities Associated with the Front End of the Fuel Cycle

Description: Pacific Northwest Laboratory (PNL) was asked by the NRC Office of Nuclear Material Safety and Safeguards (NMSS) to determine the adequacy of its health, safety and environmental regulatory base as a guide to applicants for licenses to operate UF{sub 6} conversion facilities and fuel fabrication plants. The regulatory base was defined as the body of documented requirements and guidance to licensees, including laws passed by Congress, Federal Regulations developed by the NRC to implement the laws, license conditions added to each license to deal with special requirements for that specific license, and Regulatory Guides. The study concentrated on the renewal licensing accomplished in the last few years at five typical facilities, and included analyses of licensing documents and interviews with individuals involved with different aspects of the licensing process. Those interviewed included NMSS staff, Inspection and Enforcement (IE) officials, and selected licensees. From the results of the analyses and interviews, the PNL study team concludes that the regulatory base is adequate but should be codified for greater visibility. PNL recommends that NMSS clarify distinctions among legal requirements of the licensee, acceptance criteria employed by NMSS, and guidance used by all. In particular, a prelicensing conference among NMSS, IE and each licensee would be a practical means of setting license conditions acceptable to all parties.
Date: April 1, 1982
Creator: Clark, R. G.; Schreiber, R. E.; Jamison, J. D.; Davenport, L. C. & Brite, D. W.
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

Enterprise SRS: Leveraging Ongoing Operations To Advance Nuclear Fuel Cycles Research And Development Programs

Description: The Savannah River Site (SRS) is repurposing its vast array of assets to solve future national issues regarding environmental stewardship, national security, and clean energy. The vehicle for this transformation is Enterprise SRS which presents a new, radical view of SRS as a united endeavor for ''all things nuclear'' as opposed to a group of distinct and separate entities with individual missions and organizations. Key among the Enterprise SRS strategic initiatives is the integration of research into facilities in conjunction with on-going missions to provide researchers from other national laboratories, academic institutions, and commercial entities the opportunity to demonstrate their technologies in a relevant environment and scale prior to deployment. To manage that integration of research demonstrations into site facilities, The Department of Energy, Savannah River Operations Office, Savannah River Nuclear Solutions, the Savannah River National Laboratory (SRNL) have established a center for applied nuclear materials processing and engineering research (hereafter referred to as the Center). The key proposition of this initiative is to bridge the gap between promising transformational nuclear fuel cycle processing discoveries and large commercial-scale-technology deployment by leveraging SRS assets as facilities for those critical engineering-scale demonstrations necessary to assure the successful deployment of new technologies. The Center will coordinate the demonstration of R&D technologies and serve as the interface between the engineering-scale demonstration and the R&D programs, essentially providing cradle-to-grave support to the research team during the demonstration. While the initial focus of the Center will be on the effective use of SRS assets for these demonstrations, the Center also will work with research teams to identify opportunities to perform research demonstrations at other facilities. Unique to this approach is the fact that these SRS assets will continue to accomplish DOE's critical nuclear material missions (e.g., processing in H-Canyon and plutonium storage in K-Area). Thus, the ...
Date: July 3, 2013
Creator: Murray, Alice M.; Marra, John E.; Wilmarth, William R.; Mcguire, Patrick W. & Wheeler, Vickie B.
Partner: UNT Libraries Government Documents Department


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


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

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