PREPARATION OF URANIUM HEXAFLUORIDE

DOEpatents

Lawroski, S.; Jonke, A.A.; Steunenberg, R.K.

1959-10-01

A process is described for preparing uranium hexafluoride from carbonate- leach uranium ore concentrate. The briquetted, crushed, and screened concentrate is reacted with hydrogen fluoride in a fluidized bed, and the uranium tetrafluoride formed is mixed with a solid diluent, such as calcium fluoride. This mixture is fluorinated with fluorine and an inert diluent gas, also in a fluidized bed, and the uranium hexafluoride obtained is finally purified by fractional distillation.

49 CFR 173.477 - Approval of packagings containing greater than 0.1 kg of non-fissile or fissile-excepted uranium...

Code of Federal Regulations, 2014 CFR

2014-10-01

... kg of non-fissile or fissile-excepted uranium hexafluoride. 173.477 Section 173.477 Transportation... non-fissile or fissile-excepted uranium hexafluoride. (a) Each offeror of a package containing more than 0.1 kg of uranium hexafluoride must maintain on file for at least two years after the offeror's...

49 CFR 173.477 - Approval of packagings containing greater than 0.1 kg of non-fissile or fissile-excepted uranium...

Code of Federal Regulations, 2011 CFR

2011-10-01

... kg of non-fissile or fissile-excepted uranium hexafluoride. 173.477 Section 173.477 Transportation... non-fissile or fissile-excepted uranium hexafluoride. (a) Each offeror of a package containing more than 0.1 kg of uranium hexafluoride must maintain on file for at least one year after the latest...

49 CFR 173.477 - Approval of packagings containing greater than 0.1 kg of non-fissile or fissile-excepted uranium...

Code of Federal Regulations, 2012 CFR

2012-10-01

... kg of non-fissile or fissile-excepted uranium hexafluoride. 173.477 Section 173.477 Transportation... non-fissile or fissile-excepted uranium hexafluoride. (a) Each offeror of a package containing more than 0.1 kg of uranium hexafluoride must maintain on file for at least one year after the latest...

49 CFR 173.477 - Approval of packagings containing greater than 0.1 kg of non-fissile or fissile-excepted uranium...

Code of Federal Regulations, 2013 CFR

2013-10-01

... kg of non-fissile or fissile-excepted uranium hexafluoride. 173.477 Section 173.477 Transportation... non-fissile or fissile-excepted uranium hexafluoride. (a) Each offeror of a package containing more than 0.1 kg of uranium hexafluoride must maintain on file for at least one year after the latest...

49 CFR 173.477 - Approval of packagings containing greater than 0.1 kg of non-fissile or fissile-excepted uranium...

Code of Federal Regulations, 2010 CFR

2010-10-01

... kg of non-fissile or fissile-excepted uranium hexafluoride. 173.477 Section 173.477 Transportation... non-fissile or fissile-excepted uranium hexafluoride. (a) Each offeror of a package containing more than 0.1 kg of uranium hexafluoride must maintain on file for at least one year after the latest...

Thermodynamic properties of UF sub 6 measured with a ballistic piston compressor

NASA Technical Reports Server (NTRS)

Sterritt, D. E.; Lalos, G. T.; Schneider, R. T.

1973-01-01

From experiments performed with a ballistic piston compressor, certain thermodynamic properties of uranium hexafluoride were investigated. Difficulties presented by the nonideal processes encountered in ballistic compressors are discussed and a computer code BCCC (Ballistic Compressor Computer Code) is developed to analyze the experimental data. The BCCC unfolds the thermodynamic properties of uranium hexafluoride from the helium-uranium hexafluoride mixture used as the test gas in the ballistic compressor. The thermodynamic properties deduced include the specific heat at constant volume, the ratio of specific heats for UF6, and the viscous coupling constant of helium-uranium hexafluoride mixtures.

MOLDED SEALING ELEMENT

DOEpatents

Bradford, B.W.; Skinner, W.J.

1959-03-24

Molded sealing elements suitable for use under conditions involving exposure to uranium hexafluoride vapor are described. Such sealing elements are made by subjecting graphitic carbons to a preliminary treatment with uranium hexafluoride vapor, and then incorporating polytetrafluorethylene in them. The resulting composition has good wear resistant and frictional properties and is resistant to disintegration by uranium hexafluoride over long periods of exposure.

ELECTROLYTIC PRODUCTION OF URANIUM TETRAFLUORIDE

DOEpatents

Lofthouse, E.

1954-08-31

This patent relates to electrolytic methods for the production of uranium tetrafluoride. According to the present invention a process for the production of uranium tetrafluoride comprises submitting to electrolysis an aqueous solution of uranyl fluoride containing free hydrofluoric acid. Advantageously the aqueous solution of uranyl fluoride is obtained by dissolving uranium hexafluoride in water. On electrolysis, the uranyl ions are reduced to uranous tons at the cathode and immediately combine with the fluoride ions in solution to form the insoluble uranium tetrafluoride which is precipitated.

10 CFR 40.65 - Effluent monitoring reporting requirements.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 1 2010-01-01 2010-01-01 false Effluent monitoring reporting requirements. 40.65 Section 40.65 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF SOURCE MATERIAL Records, Reports... possess and use source material in uranium milling, in production of uranium hexafluoride, or in a uranium...

THE RECOVERY OF URANIUM FROM GAS MIXTURE

DOEpatents

Jury, S.H.

1964-03-17

A method of separating uranium from a mixture of uranium hexafluoride and other gases is described that comprises bringing the mixture into contact with anhydrous calcium sulfate to preferentially absorb the uranium hexafluoride on the sulfate. The calcium sulfate is then leached with a selective solvent for the adsorbed uranium. (AEC)

75 FR 17170 - Notice of Opportunity To Request a Hearing for the License Application From International...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-04-05

... Fluoride Extraction and Uranium Deconversion Facility in Lea County NM and Order Imposing Procedures for... application from International Isotopes Fluorine Products, Inc. (IIFP), for a proposed fluoride extraction and... applicant to process depleted uranium hexafluoride (DUF 6 ) into commercially resalable fluoride products...

METHOD FOR THE RECOVERY AND PURIFICATION OF GASEOUS UF$sub 6$ FROM GASEOUS MIXTURES AND UF$sub 7$NO AND UF$sub 7$NO$sub 2$ PRODUCTS PRODUCED THEREBY

DOEpatents

Ogle, P.R. Jr.

1962-06-16

A method is given for recovering uranium hexafluoride from a gaseous mixture containing said uranium hexafluoride and extraneous gaseous impurities. The method comprises reacting said mixture with a nitrogen oxyfluoride at a temperature in the range - 100 to 50 deg C to thereby form a solid compound having the empirical formula UF/sub 7/N(O)/sub x/ where x is a number from 1 to 2. (AEC)

Pattern of explosive reaction between uranium hexafluoride and hydrocarbon oils. Revision 1

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rapp, K.E.

Examination of uranium hexafluoride release incidents occurring over the past three decades of ORGDP experience has identified only four which apparently involved an explosion of a container resulting from reaction between uranium hexafluoride and an impurity. These four incidents exhibit a certain degree of commonality. Each has involved: (1) condensed phase uranium hexafluoride, (2) a moderately elevated temperature, (3) a sufficient quantity of uranium hexafluoride for a significant partial pressure to be maintained independently above that which can be consumed by chemical reaction, and (4) an organic liquid (probably hydrocarbon oil) accidentally present in the container as a contaminant. Themore » purpose of this investigative search was to establish some conditional pattern for these four incidents to which their violent consequences could be attributed. Fortunately, the number of such incidents is relatively small, which emphasizes even more pointedly the unfortunate fact that documentation ranges from thorough to very limited. Documented sources of information are given in the bibliography. Copies of those which are not readily available are contained in six appendices. 8 refs.« less

METHOD FOR DETERMINING THE STABILITY OF FLUOROCARBON IOLS

DOEpatents

Sheldon, Z.D.; Haendler, H.M.

1959-07-21

A method of determining the stability of a fluorocarbon oil to uranium hexafluoride is presented. The method comprises reacting a weighed sample of the oil with condensed uranium hexafluoride in a reaction zone and titrating the amount of uranium tetrafluoride produced with potassium dichromate.

Galvanic cell for processing of used nuclear fuel

DOEpatents

Garcia-Diaz, Brenda L.; Martinez-Rodriguez, Michael J.; Gray, Joshua R.; Olson, Luke C.

2017-02-07

A galvanic cell and methods of using the galvanic cell is described for the recovery of uranium from used nuclear fuel according to an electrofluorination process. The galvanic cell requires no input energy and can utilize relatively benign gaseous fluorinating agents. Uranium can be recovered from used nuclear fuel in the form of gaseous uranium compound such as uranium hexafluoride, which can then be converted to metallic uranium or UO.sub.2 and processed according to known methodology to form a useful product, e.g., fuel pellets for use in a commercial energy production system.

Electrochemical fluorination for processing of used nuclear fuel

DOEpatents

Garcia-Diaz, Brenda L.; Martinez-Rodriguez, Michael J.; Gray, Joshua R.; Olson, Luke C.

2016-07-05

A galvanic cell and methods of using the galvanic cell is described for the recovery of uranium from used nuclear fuel according to an electrofluorination process. The galvanic cell requires no input energy and can utilize relatively benign gaseous fluorinating agents. Uranium can be recovered from used nuclear fuel in the form of gaseous uranium compound such as uranium hexafluoride, which can then be converted to metallic uranium or UO.sub.2 and processed according to known methodology to form a useful product, e.g., fuel pellets for use in a commercial energy production system.

Vacuum ultraviolet spectra of uranium hexafluoride/argon mixtures

NASA Technical Reports Server (NTRS)

Krascella, N. L.

1976-01-01

The transmission properties of room temperature helium at pressures up to 20 atmospheres were determined in the wavelength range from 80 to 300 nm. Similarly, the transmission properties of uranium hexafluoride at 393 K (pressures less than 1.0 mm) were determined in the wavelength range from 80 to about 120 nm. The results show that high pressure helium is sufficiently transparent in the vacuum ultraviolet region (provided trace contaminants are removed) to be utilized as a transparent purge gas in future fissioning gaseous uranium plasma reactor experiments. Absorption cross sections for uranium hexafluoride were calculated from the data between 80 and 120 nm and were of the order of 10 to the -17 power sq cm.

Plasma core reactor simulations using RF uranium seeded argon discharges

NASA Technical Reports Server (NTRS)

Roman, W. C.

1976-01-01

Experimental results are described in which pure uranium hexafluoride was injected into an argon-confined, steady-state, RF-heated plasma to investigate characteristics of plasma core nuclear reactors. The 80 kW (13.56 MHz) and 1.2 MW (5.51 MHz) rf induction heater facilities were used to determine a test chamber flow scheme which offered best uranium confinement with minimum wall coating. The cylindrical fused-silica test chamber walls were 5.7-cm-ID by 10-cm-long. Test conditions included RF powers of 2-85 kW, chamber pressures of 1-12 atm, and uranium hexafluoride mass-flow rates of 0.005-0.13 g/s. Successful techniques were developed for fluid-mechanical confinement of RF-heated plasmas with pure uranium hexafluoride injection.

PRODUCTION OF URANIUM HEXAFLUORIDE

DOEpatents

Fowler, R.D.

1957-08-27

A process for the production of uranium hexafluoride from the oxides of uranium is reported. In accordance with the method, the higher oxides of uranium may be reduced to uranium dioxide (UO/sub 2/), the latter converted into uranium tetrafluoride by reaction with hydrogen fluoride, and the UF/sub 4/ converted to UF/sub 6/ by reaction with a fluorinating agent, such as CoF/sub 3/. The UO/sub 3/ or U/sub 3/O/sub 8/ is placed in a reac tion chamber in a copper boat or tray enclosed in a copper oven, and heated to 500 to 650 deg C while hydrogen gas is passed through the oven. After nitrogen gas is used to sweep out the hydrogen and the water vapor formed, and while continuing to inaintain the temperature between 400 deg C and 600 deg C, anhydrous hydrogen fluoride is passed through. After completion of the conversion of UO/sub 2/ to UF/sub 4/ the temperature of the reaction chamber is lowered to about 400 deg C or less, the UF/sub 4/ is mixed with the requisite quantity of CoF/sub 3/, and after evacuating the chamber, the mixture is heated to 300 to 400 deg C, and the resulting UF/sub 6/ is led off and delivered to a condenser.

PROCESS FOR PRODUCING URANIUM HEXAFLUORIDE

DOEpatents

Fowler, R.D.

1957-10-22

A process for the production of uranium hexafluoride from the oxides of uranium is reported. In accordance with the method the higher oxides of uranium may be reduced to uranium dioxide (UO/sub 2/), the latter converted into uranium tetrafluoride by reaction with hydrogen fluoride, and the UF/sub 4/ convented to UF/sub 6/ by reaction with a fluorinating agent. The UO/sub 3/ or U/sub 3/O/sub 8/ is placed in a reaction chamber in a copper boat or tray enclosed in a copper oven, and heated to 500 to 650 deg C while hydrogen gas is passed through the oven. The oven is then swept clean of hydrogen and the water vapor formed by means of nitrogen and then while continuing to maintain the temperature between 400 and 600 deg C, anhydrous hydrogen fluoride is passed through. After completion of the conversion to uranium tetrafluoride, the temperature of the reaction chamber is lowered to ahout 400 deg C, and elemental fluorine is used as the fluorinating agent for the conversion of UF/sub 4/ into UF/sub 6/. The fluorine gas is passed into the chamber, and the UF/sub 6/ formed passes out and is delivered to a condenser.

Geological conditions of safe long-term storage and disposal of depleted uranium hexafluoride

NASA Astrophysics Data System (ADS)

Laverov, N. P.; Velichkin, V. I.; Omel'Yanenko, B. I.; Yudintsev, S. V.; Tagirov, B. R.

2010-08-01

The production of enriched uranium used in nuclear weapons and fuel for atomic power plants is accompanied by the formation of depleted uranium (DU), the amount of which annually increases by 35-40 kt. To date, more than 1.6 Mt DU has accumulated in the world. The main DU mass is stored as environ-mentally hazardous uranium hexafluoride (UF6), which is highly volatile and soluble in water with the formation of hydrofluoric acid. To ensure safe UF6 storage, it is necessary to convert this compound in chemically stable phases. The industrial reprocessing of UF6 into U3O8 and HF implemented in France is highly expensive. We substantiate the expediency of long-term storage of depleted uranium hexafluoride in underground repositories localized in limestone. On the basis of geochemical data and thermodynamic calculations, we show that interaction in the steel container-UF6-limestone-groundwater system gives rise to the development of a slightly alkaline reductive medium favorable for chemical reaction with formation of uraninite (UO2) and fluorite (CaF2). The proposed engineering solution not only ensures safe DU storage but also makes it possible to produce uraninite, which can be utilized, if necessary, in fast-neutron reactors. In the course of further investigations aimed at safe maintenance of DU, it is necessary to study the kinetics of conversion of UF6 into stable phases, involving laboratory and field experiments.

Study of fluoride corrosion of nickel alloys

NASA Technical Reports Server (NTRS)

Gunther, W. H.; Steindler, M. J.

1969-01-01

Report contains the results of an investigation of the corrosion resistance of nickel and nickel alloys exposed to fluorine, uranium hexafluoride, and volatile fission product fluorides at high temperatures. Survey of the unclassified literature on the subject is included.

Study of the Dry Processing of Uranium Ores; ETUDE DES TRAITEMENTS DE MINERAIS D'URANIUM PAR VOIE SECHE

DOE Office of Scientific and Technical Information (OSTI.GOV)

Guillet, H.

1959-02-01

A description is given of direct fluorination of preconcentrated uranium ores in order to obtain the hexafluoride. After normal sulfuric acid treatment of the ore to eliminate silica, the uranium is precipitated by lime to obtain either impure calcium uranate of medium grade, or containing around 10% of uranium. This concentrate is dried in an inert atmosphere and then treated with a current of elementary fluorine. The uranium hexafluoride formed is condensed at the outlet of the reaction vessel and may be used either for reduction to tetrafluoride and the subsequent manufacture of uranium metal or as the initial productmore » in a diffusion plant. (auth)« less

REACTOR NOZZLE ASSEMBLY

DOEpatents

Capuder, F.C.; Dearwater, J.R.

1959-02-10

An improved nozzle assembly useful in a process for the direct reduction of uranium hexafluoride to uranium tetrafluoride by means of dissociated ammonia in a heated reaction vessel is descrlbed. The nozzle design provides for intimate mixing of the two reactants and at the same time furnishes a layer of dissociated ammonia adjacent to the interior wall of the reaction vessel, thus preventing build-up of the reaction product on the vessel wall.

Development of solid materials for UF 6 sampling: FY16 Annual Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Smith, Nicholas; Savina, Joseph; Hebden, Andrew

2016-10-31

A handheld implementation of the ABACC-developed Cristallini method, which captures uranium hexafluoride samples as an inert salt, was organized in FY17 and succeeded in demonstrating the handheld sampler concept with reactive hexafluoride gases. The Cristallini method relies on the use of a hydrated substrate to react the incoming hexafluoride resulting in the formation of a stable uranyl fluoride salt. The Cristallini method has been demonstrated as a facility modification installed near the sampling tap of a gas centrifuge enrichment plant. While very successful in reducing the hazards of uranium hexafluoride sample, the method still takes a considerable amount of timemore » and can only be used in facilities where the apparatus has been installed; this arrangement generally prohibits the sampling of filled cylinders that have already exited the facility and have been deposited in the on-site tank storage yard. The handheld unit under development will allow the use of the Cristallini method at facilities that have not been converted as well as tanks in the storage yard. The handheld system utilizes an active vacuum system, rather than a passive vacuum system in the facility setup, to drive the uranium hexafluoride onto the adsorbing media. The handheld unit will be battery operated for fully autonomous operation and will include onboard pressure sensing and flushing capability. To date, the system concept of operations was demonstrated with tungsten hexafluoride that showed the active vacuum pump with multiple cartridges of adsorbing media was viable. Concurrently, the hardened prototype system was developed and tested; removable sample cartridges were developed (the only non-COTS component to date); and preparations were made for uranium tests and a domestic field test.« less

Depleted uranium hexafluoride: The source material for advanced shielding systems

DOE Office of Scientific and Technical Information (OSTI.GOV)

Quapp, W.J.; Lessing, P.A.; Cooley, C.R.

1997-02-01

The U.S. Department of Energy (DOE) has a management challenge and financial liability problem in the form of 50,000 cylinders containing 555,000 metric tons of depleted uranium hexafluoride (UF{sub 6}) that are stored at the gaseous diffusion plants. DOE is evaluating several options for the disposition of this UF{sub 6}, including continued storage, disposal, and recycle into a product. Based on studies conducted to date, the most feasible recycle option for the depleted uranium is shielding in low-level waste, spent nuclear fuel, or vitrified high-level waste containers. Estimates for the cost of disposal, using existing technologies, range between $3.8 andmore » $11.3 billion depending on factors such as the disposal site and the applicability of the Resource Conservation and Recovery Act (RCRA). Advanced technologies can reduce these costs, but UF{sub 6} disposal still represents large future costs. This paper describes an application for depleted uranium in which depleted uranium hexafluoride is converted into an oxide and then into a heavy aggregate. The heavy uranium aggregate is combined with conventional concrete materials to form an ultra high density concrete, DUCRETE, weighing more than 400 lb/ft{sup 3}. DUCRETE can be used as shielding in spent nuclear fuel/high-level waste casks at a cost comparable to the lower of the disposal cost estimates. Consequently, the case can be made that DUCRETE shielded casks are an alternative to disposal. In this case, a beneficial long term solution is attained for much less than the combined cost of independently providing shielded casks and disposing of the depleted uranium. Furthermore, if disposal is avoided, the political problems associated with selection of a disposal location are also avoided. Other studies have also shown cost benefits for low level waste shielded disposal containers.« less

FLUORIDE VOLATILITY PROCESS FOR THE RECOVERY OF URANIUM

DOEpatents

Katz, J.J.; Hyman, H.H.; Sheft, I.

1958-04-15

The separation and recovery of uraniunn from contaminants introduced by neutron irradiation by a halogenation and volatilization method are described. The irradiated uranium is dissolved in bromine trifluoride in the liquid phase. The uranium is converted to the BrF/sub 3/ soluble urmium hexafluoride compound whereas the fluorides of certain contaminating elements are insoluble in liquid BrF/sub 3/, and the reaction rate of the BrF/sub 3/ with certain other solid uranium contamirnnts is sufficiently slower than the reaction rate with uranium that substantial portions of these contaminating elements will remain as solids. These solids are then separated from the solution by a distillation, filtration, or centrifugation step. The uranium hexafluoride is then separated from the balance of the impurities and solvent by one or more distillations.

In-line assay monitor for uranium hexafluoride

DOEpatents

Wallace, S.A.

1980-03-21

An in-line assay monitor for determining the content of uranium-235 in a uranium hexafluoride gas isotopic separation system is provided which removes the necessity of complete access to the operating parameters of the system for determining the uranium-235 content. The method and monitor for carrying out the method involve cooling of a radiation pervious chamber connected in fluid communication with the selected point in the system to withdraw a specimen and solidify the specimen in the chamber. The specimen is irradiated by means of an ionizing radiation source of energy different from that of the 185 keV gamma emissions from uranium-235. The uranium-235 content of the specimen is determined from comparison of the accumulated 185 keV energy counts and reference energy counts. The latter is used to measure the total uranium isotopic content of the specimen.

Gas core reactors for actinide transmutation. [uranium hexafluoride

NASA Technical Reports Server (NTRS)

Clement, J. D.; Rust, J. H.; Wan, P. T.; Chow, S.

1979-01-01

The preliminary design of a uranium hexafluoride actinide transmutation reactor to convert long-lived actinide wastes to shorter-lived fission product wastes was analyzed. It is shown that externally moderated gas core reactors are ideal radiators. They provide an abundant supply of thermal neutrons and are insensitive to composition changes in the blanket. For the present reactor, an initial load of 6 metric tons of actinides is loaded. This is equivalent to the quantity produced by 300 LWR-years of operation. At the beginning, the core produces 2000 MWt while the blanket generates only 239 MWt. After four years of irradiation, the actinide mass is reduced to 3.9 metric tonnes. During this time, the blanket is becoming more fissile and its power rapidly approaches 1600 MWt. At the end of four years, continuous refueling of actinides is carried out and the actinide mass is held constant. Equilibrium is essentially achieved at the end of eight years. At equilibrium, the core is producing 1400 MWt and the blanket 1600 MWt. At this power level, the actinide destruction rate is equal to the production rate from 32 LWRs.

DISSOLUTION OF ZIRCONIUM-CONTAINING FUEL ELEMENTS

DOEpatents

Horn, F.L.

1961-12-12

Uranium is recovered from spent uranium fuel elements containing or clad with zirconium. These fuel elements are placed in an anhydrous solution of hydrogen fluoride and nitrogen dioxide. Within this system uranium forms a soluble complex and zirconium forms an insoluble complex. The uranium can then be separated, treated, and removed from solution as uranium hexafluoride. (AEC)

10 CFR 70.59 - Effluent monitoring reporting requirements.

Code of Federal Regulations, 2011 CFR

2011-01-01

... fabrication, scrap recovery, conversion of uranium hexafluoride, or in a uranium enrichment facility shall... this specifically. On the basis of these reports and any additional information the Commission may...

Partially Ionized Plasmas, Including the Third Symposium on Uranium Plasmas

NASA Technical Reports Server (NTRS)

Krishnan, M.

1976-01-01

Fundamentals of both electrically and fission generated plasmas are discussed. Research in gaseous fuel reactors using uranium hexafluoride is described and other partially ionized plasma applications are discussed.

DISSOLUTION OF URANIUM FUELS BY MONOOR DIFLUOROPHOSPHORIC ACID

DOEpatents

Johnson, R.; Horn, F.L.; Strickland, G.

1963-05-01

A method of dissolving and separating uranium from a uranium matrix fuel element by dissolving the uraniumcontaining matrix in monofluorophosphoric acid and/or difluorophosphoric acid at temperatures ranging from 150 to 275 un. Concent 85% C, thereafter neutralizing the solution to precipitate uranium solids, and converting the solids to uranium hexafluoride by treatment with a halogen trifluoride is presented. (AEC)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zoller, J.N.; Rosen, R.S.; Holliday, M.A.

With the publication of a Request for Recommendations and Advance Notice of Intent in the November 10, 1994 Federal Register, the Department of Energy initiated a program to assess alternative strategies for the long-term management or use of depleted uranium hexafluoride. This Request was made to help ensure that, by seeking as many recommendations as possible, Department management considers reasonable options in the long-range management strategy. The Depleted Uranium Hexafluoride Management Program consists of three major program elements: Engineering Analysis, Cost Analysis, and an Environmental Impact Statement. This Technology Assessment Report is the first part of the Engineering Analysis Project,more » and assesses recommendations from interested persons, industry, and Government agencies for potential uses for the depleted uranium hexafluoride stored at the gaseous diffusion plants in Paducah, Kentucky, and Portsmouth, Ohio, and at the Oak Ridge Reservation in Tennessee. Technologies that could facilitate the long-term management of this material are also assessed. The purpose of the Technology Assessment Report is to present the results of the evaluation of these recommendations. Department management will decide which recommendations will receive further study and evaluation.« less

Uniform deposition of uranium hexafluoride (UF6): Standardized mass deposits and controlled isotopic ratios using a thermal fluorination method

DOE Office of Scientific and Technical Information (OSTI.GOV)

McNamara, Bruce K.; O’Hara, Matthew J.; Casella, Andrew M.

2016-07-01

Abstract: We report a convenient method for the generation of volatile uranium hexafluoride (UF6) from solid uranium oxides and other uranium compounds, followed by uniform deposition of low levels of UF6 onto sampling coupons. Under laminar flow conditions, UF6 is shown to interact with surfaces within the chamber to a highly predictable degree. We demonstrate the preparation of uranium deposits that range between ~0.01 and 470±34 ng∙cm-2. The data suggest the method can be extended to creating depositions at the sub-picogram∙cm-2 level. Additionally, the isotopic composition of the deposits can be customized by selection of the uranium source materials. Wemore » demonstrate a layering technique whereby two uranium solids, each with a different isotopic composition, are employed to form successive layers of UF6 on a surface. The result is an ultra-thin deposit of UF6 that bears an isotopic signature that is a composite of the two uranium sources. The reported deposition method has direct application to the development of unique analytical standards for nuclear safeguards and forensics.« less

News Media Exits for Depleted Uranium and Depleted UF6 Articles

Science.gov Websites

hexafluoride, uranium privatization, Paducah, and Portsmouth. The New York Times The Washington Post The Chicago Tribune The Kentucky Post Dayton Daily News USA Today The Courier Journal The Wall Street Journal

METHOD OF SEPARATING PLUTONIUM

DOEpatents

Brown, H.S.; Hill, O.F.

1958-02-01

Plutonium hexafluoride is a satisfactory fluorinating agent and may be reacted with various materials capable of forming fluorides, such as copper, iron, zinc, etc., with consequent formation of the metal fluoride and reduction of the plutonium to the form of a lower fluoride. In accordance with the present invention, it has been found that the reactivity of plutonium hexafluoride with other fluoridizable materials is so great that the process may be used as a method of separating plutonium from mixures containing plutonium hexafluoride and other vaporized fluorides even though the plutonium is present in but minute quantities. This process may be carried out by treating a mixture of fluoride vapors comprising plutonium hexafluoride and fluoride of uranium to selectively reduce the plutonium hexafluoride and convert it to a less volatile fluoride, and then recovering said less volatile fluoride from the vapor by condensation.

Frequently Asked Questions (FAQs)

Science.gov Websites

DUF6 Guide DU Uses DUF6 Management and Uses DUF6 Conversion EIS Documents News FAQs Internet Resources , uranium hexafluoride (UF6), UF6 storage and depleted UF6 management. Below is a list of frequently asked management of depleted uranium. Click a question below to see the answer. Uranium and Its Properties What is

Ballistic piston fissioning plasma experiment.

NASA Technical Reports Server (NTRS)

Miller, B. E.; Schneider, R. T.; Thom, K.; Lalos, G. T.

1971-01-01

The production of fissioning uranium plasma samples such that the fission fragment stopping distance is less than the dimensions of the plasma is approached by using a ballistic piston device for the compression of uranium hexafluoride. The experimental apparatus is described. At room temperature the gun can be loaded up to 100 torr UF6 partial pressure, but at compression a thousand fold increase of pressure can be obtained at a particle density on the order of 10 to the 19th power per cu cm. Limited spectral studies of UF6 were performed while obtaining the pressure-volume data. The results obtained and their implications are discussed.

Uranium hexafluoride: Safe handling, processing, and transporting: Conference proceedings

DOE Office of Scientific and Technical Information (OSTI.GOV)

Strunk, W.D.; Thornton, S.G.

This conference seeks to provide a forum for the exchange of information and ideas of the safety aspects and technical issue related to the handling of uranium hexafluoride. By allowing operators, engineers, scientists, managers, educators, and others to meet and share experiences of mutual concern, the conference is also intended to provide the participants with a more complete knowledge of technical and operational issues. The topics for the papers in the proceedings are widely varied and include the results of chemical, metallurgical, mechanical, thermal, and analytical investigations, as well as the developed philosophies of operational, managerial, and regulatory guidelines. Papersmore » have been entered individually into EDB and ERA. (LTN)« less

Reduction of uranium hexafluoride to tetrafluoride by using the hydrogen atoms

NASA Astrophysics Data System (ADS)

Aleksandrov, B. P.; Gordon, E. B.; Ivanov, A. V.; Kotov, A. A.; Smirnov, V. E.

2016-09-01

We consider the reduction of UF6 to UF4 by chemical reaction with hydrogen atoms originated in the powerful chemical generator. The principal design of such a chemical convertor is described. The results of the mathematical modeling of the thermodynamics and kinetics of the UF6 to UF4 reduction process are analyzed. The few options for the hydrogen atom generator design are proposed. A layout of the experimental setup with the chemical reactor is presented. The high efficiency together with the ability of the process scaling without loss of its efficiency makes this approach to the uranium hexafluoride depletion into tetrafluoride promising for its application in the industry.

In-line assay monitor for uranium hexafluoride

DOEpatents

Wallace, Steven A.

1981-01-01

An in-line assay monitor for determining the content of uranium-235 in a uranium hexafluoride gas isotopic separation system is provided which removes the necessity of complete access to the operating parameters of the system for determining the uranium-235 content. The monitor is intended for uses such as safeguard applications to assure that weapons grade uranium is not being produced in an enrichment cascade. The method and monitor for carrying out the method involve cooling of a radiation pervious chamber connected in fluid communication with the selected point in the system to withdraw a specimen and solidify the specimen in the chamber. The specimen is irradiated by means of an ionizing radiation source of energy different from that of the 185 keV gamma emissions from the uranium-235 present in the specimen. Simultaneously, the gamma emissions from the uranium-235 of the specimen and the source emissions transmitted through the sample are counted and stored in a multiple channel analyzer. The uranium-235 content of the specimen is determined from the comparison of the accumulated 185 keV energy counts and the reference energy counts. The latter is used to measure the total uranium isotopic content of the specimen. The process eliminates the necessity of knowing the system operating conditions and yet obtains the necessary data without need for large scintillation crystals and sophisticated mechanical designs.

Parametric analyses of planned flowing uranium hexafluoride critical experiments

NASA Technical Reports Server (NTRS)

Rodgers, R. J.; Latham, T. S.

1976-01-01

Analytical investigations were conducted to determine preliminary design and operating characteristics of flowing uranium hexafluoride (UF6) gaseous nuclear reactor experiments in which a hybrid core configuration comprised of UF6 gas and a region of solid fuel will be employed. The investigations are part of a planned program to perform a series of experiments of increasing performance, culminating in an approximately 5 MW fissioning uranium plasma experiment. A preliminary design is described for an argon buffer gas confined, UF6 flow loop system for future use in flowing critical experiments. Initial calculations to estimate the operating characteristics of the gaseous fissioning UF6 in a confined flow test at a pressure of 4 atm, indicate temperature increases of approximately 100 and 1000 K in the UF6 may be obtained for total test power levels of 100 kW and 1 MW for test times of 320 and 32 sec, respectively.

49 CFR 173.417 - Authorized fissile materials packages.

Code of Federal Regulations, 2010 CFR

2010-10-01

... for export and import shipments. (2) A residual “heel” of enriched solid uranium hexafluoride may be... “Heel” in a Specification 7A Cylinder) Maximum cylinder diameter Centimeters Inches Cylinder volume Liters Cubic feet Maximum Uranium 235-enrichment (weight)percent Maximum “Heel” weight per cylinder UF6...

49 CFR 173.417 - Authorized fissile materials packages.

Code of Federal Regulations, 2011 CFR

2011-10-01

... for export and import shipments. (2) A residual “heel” of enriched solid uranium hexafluoride may be... “Heel” in a Specification 7A Cylinder) Maximum cylinder diameter Centimeters Inches Cylinder volume Liters Cubic feet Maximum Uranium 235-enrichment (weight)percent Maximum “Heel” weight per cylinder UF6...

METHOD OF DEHYDRATING URANIUM TETRAFLUORIDE

DOEpatents

Davis, J.O.; Fogel, C.C.; Palmer, W.E.

1962-12-18

Drying and dehydration of aqueous-precipitated uranium tetrafluoride are described. The UF/sub 4/ which normally contains 3 to 4% water, is dispersed into the reaction zone of an operating reactor wherein uranium hexafluoride is being reduced to UF/sub 4/ with hydrogen. The water-containing UF/sub 4/ is dried and blended with the UF/sub 4/ produced in the reactor without interfering with the reduction reaction. (AEC)

CONTINUOUS PROCESS FOR PREPARING URANIUM HEXAFLUORIDE FROM URANIUM TETRAFLUORIDE AND OXYGEN

DOEpatents

Adams, J.B.; Bresee, J.C.; Ferris, L.M.

1961-11-21

A process for preparing UF/sub 6/ by reacting UF/sub 4/ and oxygen is described. The UF/sub 4/ and oxygen are continuously introduced into a fluidized bed of UO/sub 2/F/sub 2/ at a temperature of 600 to 900 deg C. The concentration of UF/sub 4/ in the bed is maintained below 25 weight per cent in order to avoid sintering and intermediate compound formation. By-product U0/sub 2/F/sub 2/ is continuously removed from the top of the bed recycled. In an alternative embodiment heat is supplied to the reaction bed by burning carbon monoxide in the bed. The product UF/sub 6/ is filtered to remove entrained particles and is recovered in cold traps and chemical traps. (AEC)

Refurbishment of uranium hexafluoride cylinder storage yards C-745-K, L, M, N, and P and construction of a new uranium hexafluoride cylinder storage yard (C-745-T) at the Paducah Gaseous Diffusion Plant, Paducah, Kentucky

DOE Office of Scientific and Technical Information (OSTI.GOV)

NONE

1996-07-01

The Paducah Gaseous Diffusion Plant (PGDP) is a uranium enrichment facility owned by the US Department of Energy (DOE). A residual of the uranium enrichment process is depleted uranium hexafluoride (UF6). Depleted UF6, a solid at ambient temperature, is stored in 32,200 steel cylinders that hold a maximum of 14 tons each. Storage conditions are suboptimal and have resulted in accelerated corrosion of cylinders, increasing the potential for a release of hazardous substances. Consequently, the DOE is proposing refurbishment of certain existing yards and construction of a new storage yard. This environmental assessment (EA) evaluates the impacts of the proposedmore » action and no action and considers alternate sites for the proposed new storage yard. The proposed action includes (1) renovating five existing cylinder yards; (2) constructing a new UF6 storage yard; handling and onsite transport of cylinders among existing yards to accommodate construction; and (4) after refurbishment and construction, restacking of cylinders to meet spacing and inspection requirements. Based on the results of the analysis reported in the EA, DOE has determined that the proposed action is not a major Federal action that would significantly affect the quality of the human environment within the context of the National Environmental Policy Act of 1969. Therefore, DOE is issuing a Finding of No Significant Impact. Additionally, it is reported in this EA that the loss of less than one acre of wetlands at the proposed project site would not be a significant adverse impact.« less

Process for producing enriched uranium having a .sup.235 U content of at least 4 wt. % via combination of a gaseous diffusion process and an atomic vapor laser isotope separation process to eliminate uranium hexafluoride tails storage

DOEpatents

Horton, James A.; Hayden, Jr., Howard W.

1995-01-01

An uranium enrichment process capable of producing an enriched uranium, having a .sup.235 U content greater than about 4 wt. %, is disclosed which will consume less energy and produce metallic uranium tails having a lower .sup.235 U content than the tails normally produced in a gaseous diffusion separation process and, therefore, eliminate UF.sub.6 tails storage and sharply reduce fluorine use. The uranium enrichment process comprises feeding metallic uranium into an atomic vapor laser isotope separation process to produce an enriched metallic uranium isotopic mixture having a .sup.235 U content of at least about 2 wt. % and a metallic uranium residue containing from about 0.1 wt. % to about 0.2 wt. % .sup.235 U; fluorinating this enriched metallic uranium isotopic mixture to form UF.sub.6 ; processing the resultant isotopic mixture of UF.sub.6 in a gaseous diffusion process to produce a final enriched uranium product having a .sup.235 U content of at least 4 wt. %, and up to 93.5 wt. % or higher, of the total uranium content of the product, and a low .sup.235 U content UF.sub.6 having a .sup.235 U content of about 0.71 wt. % of the total uranium content of the low .sup.235 U content UF.sub.6 ; and converting this low .sup.235 U content UF.sub.6 to metallic uranium for recycle to the atomic vapor laser isotope separation process.

Process for producing enriched uranium having a {sup 235}U content of at least 4 wt. % via combination of a gaseous diffusion process and an atomic vapor laser isotope separation process to eliminate uranium hexafluoride tails storage

DOEpatents

Horton, J.A.; Hayden, H.W. Jr.

1995-05-30

An uranium enrichment process capable of producing an enriched uranium, having a {sup 235}U content greater than about 4 wt. %, is disclosed which will consume less energy and produce metallic uranium tails having a lower {sup 235}U content than the tails normally produced in a gaseous diffusion separation process and, therefore, eliminate UF{sub 6} tails storage and sharply reduce fluorine use. The uranium enrichment process comprises feeding metallic uranium into an atomic vapor laser isotope separation process to produce an enriched metallic uranium isotopic mixture having a {sup 235} U content of at least about 2 wt. % and a metallic uranium residue containing from about 0.1 wt. % to about 0.2 wt. % {sup 235} U; fluorinating this enriched metallic uranium isotopic mixture to form UF{sub 6}; processing the resultant isotopic mixture of UF{sub 6} in a gaseous diffusion process to produce a final enriched uranium product having a {sup 235}U content of at least 4 wt. %, and up to 93.5 wt. % or higher, of the total uranium content of the product, and a low {sup 235}U content UF{sub 6} having a {sup 235}U content of about 0.71 wt. % of the total uranium content of the low {sup 235}U content UF{sub 6}; and converting this low {sup 235}U content UF{sub 6} to metallic uranium for recycle to the atomic vapor laser isotope separation process. 4 figs.

Trimolecular reactions of uranium hexafluoride with water.

PubMed

Lind, Maria C; Garrison, Stephen L; Becnel, James M

2010-04-08

The hydrolysis reaction of uranium hexafluoride (UF(6)) is a key step in the synthesis of uranium dioxide (UO(2)) powder for nuclear fuels. Mechanisms for the hydrolysis reactions are studied here with density functional theory and the Stuttgart small-core scalar relativistic pseudopotential and associated basis set for uranium. The reaction of a single UF(6) molecule with a water molecule in the gas phase has been previously predicted to proceed over a relatively sizable barrier of 78.2 kJ x mol(-1), indicating this reaction is only feasible at elevated temperatures. Given the observed formation of a second morphology for the UO(2) product coupled with the observations of rapid, spontaneous hydrolysis at ambient conditions, an alternate reaction pathway must exist. In the present work, two trimolecular hydrolysis mechanisms are studied with density functional theory: (1) the reaction between two UF(6) molecules and one water molecule, and (2) the reaction of two water molecules with a single UF(6) molecule. The predicted reaction of two UF(6) molecules with one water molecule displays an interesting "fluorine-shuttle" mechanism, a significant energy barrier of 69.0 kJ x mol(-1) to the formation of UF(5)OH, and an enthalpy of reaction (DeltaH(298)) of +17.9 kJ x mol(-1). The reaction of a single UF(6) molecule with two water molecules displays a "proton-shuttle" mechanism, and is more favorable, having a slightly lower computed energy barrier of 58.9 kJ x mol(-1) and an exothermic enthalpy of reaction (DeltaH(298)) of -13.9 kJ x mol(-1). The exothermic nature of the overall UF(6) + 2H(2)O trimolecular reaction and the lowering of the barrier height with respect to the bimolecular reaction are encouraging.

Fissioning uranium plasmas and nuclear-pumped lasers

NASA Technical Reports Server (NTRS)

Schneider, R. T.; Thom, K.

1975-01-01

Current research into uranium plasmas, gaseous-core (cavity) reactors, and nuclear-pumped lasers is discussed. Basic properties of fissioning uranium plasmas are summarized together with potential space and terrestrial applications of gaseous-core reactors and nuclear-pumped lasers. Conditions for criticality of a uranium plasma are outlined, and it is shown that the nonequilibrium state and the optical thinness of a fissioning plasma can be exploited for the direct conversion of fission fragment energy into coherent light (i.e., for nuclear-pumped lasers). Successful demonstrations of nuclear-pumped lasers are described together with gaseous-fuel reactor experiments using uranium hexafluoride.

Optimization of air plasma reconversion of UF6 to UO2 based on thermodynamic calculations

NASA Astrophysics Data System (ADS)

Tundeshev, Nikolay; Karengin, Alexander; Shamanin, Igor

2018-03-01

The possibility of plasma-chemical conversion of depleted uranium-235 hexafluoride (DUHF) in air plasma in the form of gas-air mixtures with hydrogen is considered in the paper. Calculation of burning parameters of gas-air mixtures is carried out and the compositions of mixtures obtained via energy-efficient conversion of DUHF in air plasma are determined. With the help of plasma-chemical conversion, thermodynamic modeling optimal composition of UF6-H2-Air mixtures and its burning parameters, the modes for production of uranium dioxide in the condensed phase are determined. The results of the conducted researches can be used for creation of technology for plasma-chemical conversion of DUHF in the form of air-gas mixtures with hydrogen.

Synthesis of pentafluorides

DOEpatents

Asprey, L.B.; Paine, R.T. Jr.

1975-12-30

The reactions of uranium, molybdenum, rhenium, osmium and iridium hexafluorides with hydrogen gas in the presence of ultraviolet radiation or with silicon powder in an anhydrous HF slurry provide especially useful, high yield syntheses of pure pentafluorides.

Investigation of applications for high-power, self-critical fissioning uranium plasma reactors

NASA Technical Reports Server (NTRS)

Rodgers, R. J.; Latham, T. S.; Krascella, N. L.

1976-01-01

Analytical studies were conducted to investigate potentially attractive applications for gaseous nuclear cavity reactors fueled by uranium hexafluoride and its decomposition products at temperatures of 2000 to 6000 K and total pressures of a few hundred atmospheres. Approximate operating conditions and performance levels for a class of nuclear reactors in which fission energy removal is accomplished principally by radiant heat transfer from the high temperature gaseous nuclear fuel to surrounding absorbing media were determined. The results show the radiant energy deposited in the absorbing media may be efficiently utilized in energy conversion system applications which include (1) a primary energy source for high thrust, high specific impulse space propulsion, (2) an energy source for highly efficient generation of electricity, and (3) a source of high intensity photon flux for heating working fluid gases for hydrogen production or MHD power extraction.

Physics and potentials of fissioning plasmas for space power and propulsion

NASA Technical Reports Server (NTRS)

Thom, K.; Schwenk, F. C.; Schneider, R. T.

1976-01-01

Fissioning uranium plasmas are the nuclear fuel in conceptual high-temperature gaseous-core reactors for advanced rocket propulsion in space. A gaseous-core nuclear rocket would be a thermal reactor in which an enriched uranium plasma at about 10,000 K is confined in a reflector-moderator cavity where it is nuclear critical and transfers its fission power to a confining propellant flow for the production of thrust at a specific impulse up to 5000 sec. With a thrust-to-engine weight ratio approaching unity, the gaseous-core nuclear rocket could provide for propulsion capabilities needed for manned missions to the nearby planets and for economical cislunar ferry services. Fueled with enriched uranium hexafluoride and operated at temperatures lower than needed for propulsion, the gaseous-core reactor scheme also offers significant benefits in applications for space and terrestrial power. They include high-efficiency power generation at low specific mass, the burnup of certain fission products and actinides, the breeding of U-233 from thorium with short doubling times, and improved convenience of fuel handling and processing in the gaseous phase.

Spectral properties of gaseous uranium hexafluoride at high temperature

NASA Technical Reports Server (NTRS)

Krascella, N. L.

1980-01-01

A study to determine relative spectral emission and spectral absorption data for UF6-argon mixtures at elevated temperatures is discussed. These spectral data are required to assist in the theoretical analysis of radiation transport in the nuclear fuel-buffer gas region of a plasma core reactor. Relative emission measurements were made for UF6-argon mixtures over a range of temperatures from 650 to 1900 K and in the wavelength range from 600 to 5000 nanometers. All emission results were determined for a total pressure of 1.0 atm. Uranium hexafluoride partial pressures varied from about 3.5 to 12.7 mm Hg. Absorption measurements were attempted at 600, 625, 650 and 675 nanometers for a temperature of 1000 K. The uranium partial pressure for these determinations was 25 mm Hg. The results exhibit appreciable emission for hot UF6-argon mixtures at wavelengths between 600 and 1800 nanometers and no measurable absorption. The equipment used to evaluate the spectral properties of the UF6-argon mixtures included a plasma torch-optical plenum assembly, the monochromator, and the UF6 transfer system. Each is described.

SEPARATION OF URANIUM HEXAFLUORIDE FROM ORGANIC FLUORO COMPOUNDS

DOEpatents

Libby, W.F.

1958-10-01

A method is presented for removing perfiuoroorganic compounds such as C/ sub 7/F/sub 16/ from UF/sub 6/. The physical and chemical properties of the perfluoro compounds are such as to render their removal from UF/sub 6/ difficulty by conventional techniques. The mixture containing UF/sub 6/ and the perfluoro compounds is pyrolyzed in an inert container at high temperature and pressure. The properties of the products obtained by pyrolysis differ from the properties of UF/sub 6/ to a sufficient degree to render their separation possible by ordinary methods.

FLUORINE PROCESS FOR SEPARATION OF MATERIALS

DOEpatents

Seaborg, G.T.; Brown, H.S.

1958-05-01

A process is described for separating plutoniunn from neutron-irradiated uranium, which consists of reacting the irradiated uranium mass with HF to form the tetrafluorides of U, Pu, and Np, and then reacting this mixture of tetrafluorides with fiuorine at temperature between 140 and 315 d C. This causes volatile hexafluorides of U and Np to form while at the temperature employed the Pu tetrafluoride is unaffected and remains as a residue.

CONTINUOUS PRECIPITATION METHOD FOR CONVERSION OF URANYL NITRATE TO URANIUM HEXAFLUORIDE

DOEpatents

Reinhart, G.M.; Collopy, T.J.

1962-11-13

A continuous precipitation process is given for converting a uranyl nitrate solution to uranium tetrafluoride. A stream of the uranyl nitrate solution and a stream of an aqueous ammonium hydroxide solution are continuously introduced into an agitated reaction zone maintained at a pH of 5.0 to 6.5. Flow rates are adjusted to provide a mean residence time of the resulting slurry in the reaction zone of at least 30 minutes. After a startup period of two hours the precipitate is recovered from the effluent stream by filtration and is converted to uranium tetrafluoride by reduction to uranium dioxide with hydrogen and reaction of the uranium dioxide with anhydrous hydrogen fluoride. (AEC)

Properties of radio-frequency heated argon confined uranium plasmas

NASA Technical Reports Server (NTRS)

1976-01-01

Pure uranium hexafluoride (UF6) was injected into an argon confined, steady state, rf-heated plasma within a fused silica peripheral wall test chamber. Exploratory tests conducted using an 80 kW rf facility and different test chamber flow configurations permitted selection of the configuration demonstrating the best confinement characteristics and minimum uranium compound wall coating. The overall test results demonstrated applicable flow schemes and associated diagnostic techniques were developed for the fluid mechanical confinement and characterization of uranium within an rf plasma discharge when pure UF6 is injected for long test times into an argon-confined, high-temperature, high-pressure, rf-heated plasma.

Method for fluorination of uranium oxide

DOEpatents

Petit, George S.

1987-01-01

Highly pure uranium hexafluoride is made from uranium oxide and fluorine. The uranium oxide, which includes UO.sub.3, UO.sub.2, U.sub.3 O.sub.8 and mixtures thereof, is introduced together with a small amount of a fluorine-reactive substance, selected from alkali chlorides, silicon dioxide, silicic acid, ferric oxide, and bromine, into a constant volume reaction zone. Sufficient fluorine is charged into the zone at a temperature below approximately 0.degree. C. to provide an initial pressure of at least approximately 600 lbs/sq. in. at the ambient atmospheric temperature. The temperature is then allowed to rise in the reaction zone until reaction occurs.

PLATINUM HEXAFLUORIDE AND METHOD OF FLUORINATING PLUTONIUM CONTAINING MIXTURES THERE-WITH

DOEpatents

Malm, J.G.; Weinstock, B.; Claassen, H.H.

1959-07-01

The preparation of platinum hexafluoride and its use as a fluorinating agent in a process for separating plutonium from fission products is presented. According to the invention, platinum is reacted with fluorine gas at from 900 to 1100 deg C to form platinum hexafluoride. The platinum hexafluoride is then contacted with the plutonium containing mixture at room temperature to form plutonium hexafluoride which is more volatile than the fission products fluorides and therefore can be isolated by distillation.

Containment and storage of uranium hexafluoride at US Department of Energy uranium enrichment plants

DOE Office of Scientific and Technical Information (OSTI.GOV)

Barlow, C.R.; Alderson, J.H.; Blue, S.C.

Isotopically depleted UF{sub 6} (uranium hexafluoride) accumulates at a rate five to ten times greater than the enriched product and is stored in steel vessels at the enrichment plant sites. There are approximately 55,000 large cylinders now in storage at Paducah, Kentucky; Portsmouth, Ohio; and Oak Ridge, Tennessee. Most of them contain a nominal 14 tons of depleted UF{sub 6}. Some of these cylinders have been in the unprotected outdoor storage environment for periods approaching 40 years. Storage experience, supplemented by limited corrosion data, suggests a service life of about 70 years under optimum conditions for the 48-in. diameter, 5/16-in.-wallmore » pressure vessels (100 psi working pressure), using a conservative industry-established 1/4-in.-wall thickness as the service limit. In the past few years, however, factors other than atmospheric corrosion have become apparent that adversely affect the serviceability of small numbers of the storage containers and that indicate the need for a managed program to ensure maintenance ofcontainment integrity for all the cylinders in storage. The program includes periodic visual inspections of cylinders and storage yards with documentation for comparison with other inspections, a group of corrosion test programs to permit cylinder life forecasts, and identification of (and scheduling for remedial action) situations in which defects, due to handling damage or accelerated corrosion, can seriously shorten the storage life or compromise the containment integrity of individual cylinders. The program also includes rupture testing to assess the effects of certain classes of damage on overall cylinder strength, aswell as ongoing reviews of specifications, procedures, practices, and inspection results to effect improvements in handling safety, containment integrity, and storage life.« less

The spectral properties of uranium hexafluoride and its thermal decomposition products

NASA Technical Reports Server (NTRS)

Krascella, N. L.

1976-01-01

This investigation was initiated to provide basic spectral data for gases of interest to the plasma core reactor concept. The attenuation of vacuum ultraviolet (VUV) radiation by helium at pressures up to 20 atm over path lengths of about 61 cm and in the approximate wavelength range between 80 and 300 nm was studied. Measurements were also conducted to provide basic VUV data with respect to UF6 and UF6/argon mixtures in the wavelength range between 80 and 120 nm. Finally, an investigation was initiated to provide basic spectral emission and absorption data for UF6 and possible thermal decomposition products of UF6 at elevated temperatures.

Conversion of depleted uranium hexafluoride to a solid uranium compound

DOEpatents

Rothman, Alan B.; Graczyk, Donald G.; Essling, Alice M.; Horwitz, E. Philip

2001-01-01

A process for converting UF.sub.6 to a solid uranium compound such as UO.sub.2 and CaF. The UF.sub.6 vapor form is contacted with an aqueous solution of NH.sub.4 OH at a pH greater than 7 to precipitate at least some solid uranium values as a solid leaving an aqueous solution containing NH.sub.4 OH and NH.sub.4 F and remaining uranium values. The solid uranium values are separated from the aqueous solution of NH.sub.4 OH and NH.sub.4 F and remaining uranium values which is then diluted with additional water precipitating more uranium values as a solid leaving trace quantities of uranium in a dilute aqueous solution. The dilute aqueous solution is contacted with an ion-exchange resin to remove substantially all the uranium values from the dilute aqueous solution. The dilute solution being contacted with Ca(OH).sub.2 to precipitate CaF.sub.2 leaving dilute NH.sub.4 OH.

Bonding in uranium(V) hexafluoride based on the experimental electron density distribution measured at 20 K

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gianopoulos, Christopher G.; Zhurov, Vladimir V.; Minasian, Stefan G.

The electron density distribution of [PPh 4][UF 6] was obtained from high-resolution X-ray diffraction data measured at 20 K. The electron density was modeled with an augmented Hansen–Coppens multipolar formalism. Topological analysis reveals that the U–F bond is of incipient covalent nature. Theoretical calculations add further support to the bonding description gleaned from the experimental model. The impact of the uranium anomalous dispersion terms on the refinement is also discussed.

Bonding in uranium(V) hexafluoride based on the experimental electron density distribution measured at 20 K

DOE PAGES

Gianopoulos, Christopher G.; Zhurov, Vladimir V.; Minasian, Stefan G.; ...

2017-02-06

The electron density distribution of [PPh 4][UF 6] was obtained from high-resolution X-ray diffraction data measured at 20 K. The electron density was modeled with an augmented Hansen–Coppens multipolar formalism. Topological analysis reveals that the U–F bond is of incipient covalent nature. Theoretical calculations add further support to the bonding description gleaned from the experimental model. The impact of the uranium anomalous dispersion terms on the refinement is also discussed.

Method of absorbing UF.sub.6 from gaseous mixtures in alkamine absorbents

DOEpatents

Lafferty, Robert H.; Smiley, Seymour H.; Radimer, Kenneth J.

1976-04-06

A method of recovering uranium hexafluoride from gaseous mixtures employing as an absorbent a liquid composition at least one of the components of which is chosen from the group consisting of ethanolamine, diethanolamine, and 3-methyl-3-amino-propane-diol-1,2.

High temperature UF6 RF plasma experiments applicable to uranium plasma core reactors

NASA Technical Reports Server (NTRS)

Roman, W. C.

1979-01-01

An investigation was conducted using a 1.2 MW RF induction heater facility to aid in developing the technology necessary for designing a self critical fissioning uranium plasma core reactor. Pure, high temperature uranium hexafluoride (UF6) was injected into an argon fluid mechanically confined, steady state, RF heated plasma while employing different exhaust systems and diagnostic techniques to simulate and investigate some potential characteristics of uranium plasma core nuclear reactors. The development of techniques and equipment for fluid mechanical confinement of RF heated uranium plasmas with a high density of uranium vapor within the plasma, while simultaneously minimizing deposition of uranium and uranium compounds on the test chamber peripheral wall, endwall surfaces, and primary exhaust ducts, is discussed. The material tests and handling techniques suitable for use with high temperature, high pressure, gaseous UF6 are described and the development of complementary diagnostic instrumentation and measurement techniques to characterize the uranium plasma, effluent exhaust gases, and residue deposited on the test chamber and exhaust system components is reported.

Method for monitoring stack gases for uranium activity

DOEpatents

Beverly, C.R.; Ernstberger, E.G.

1985-07-03

A method for monitoring the stack gases of a purge cascade of gaseous diffusion plant for uranium activity. A sample stream is taken from the stack gases and contacted with a volume of moisture-laden air for converting trace levels of uranium hexafluoride, if any, in the stack gases into particulate uranyl fluoride. A continuous strip of filter paper from a supply roll is passed through this sampling stream to intercept and gather any uranyl fluoride in the sampling stream. This filter paper is then passed by an alpha scintillation counting device where any radioactivity on the filter paper is sensed so as to provide a continuous monitoring of the gas stream for activity indicative of the uranium content in the stack gases. 1 fig.

Method for monitoring stack gases for uranium activity

DOEpatents

Beverly, Claude R.; Ernstberger, Harold G.

1988-01-01

A method for monitoring the stack gases of a purge cascade of a gaseous diffusion plant for uranium activity. A sample stream is taken from the stack gases and contacted with a volume of moisture-laden air for converting trace levels of uranium hexafluoride, if any, in the stack gases into particulate uranyl fluoride. A continuous strip of filter paper from a supply roll is passed through this sampling stream to intercept and gather any uranyl fluoride in the sampling stream. This filter paper is then passed by an alpha scintillation counting device where any radioactivity on the filter paper is sensed so as to provide a continuous monitoring of the gas stream for activity indicative of the uranium content in the stack gases.

Field test of short-notice random inspections for inventory-change verification at a low-enriched-uranium fuel-fabrication plant: Preliminary summary

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fishbone, L.G.; Moussalli, G.; Naegele, G.

1994-04-01

An approach of short-notice random inspections (SNRIs) for inventory-change verification can enhance the effectiveness and efficiency of international safeguards at natural or low-enriched uranium (LEU) fuel fabrication plants. According to this approach, the plant operator declares the contents of nuclear material items before knowing if an inspection will occur to verify them. Additionally, items about which declarations are newly made should remain available for verification for an agreed time. This report details a six-month field test of the feasibility of such SNRIs which took place at the Westinghouse Electric Corporation Commercial Nuclear Fuel Division. Westinghouse personnel made daily declarations aboutmore » both feed and product items, uranium hexafluoride cylinders and finished fuel assemblies, using a custom-designed computer ``mailbox``. Safeguards inspectors from the IAEA conducted eight SNRIs to verify these declarations. Items from both strata were verified during the SNRIs by means of nondestructive assay equipment. The field test demonstrated the feasibility and practicality of key elements of the SNRI approach for a large LEU fuel fabrication plant.« less

Argon/UF6 plasma experiments: UF6 regeneration and product analysis

NASA Technical Reports Server (NTRS)

Roman, W. C.

1980-01-01

An experimental and analytical investigation was conducted to aid in developing some of the technology necessary for designing a self-critical fissioning uranium plasma core reactors (PCR). This technology is applicable to gaseous uranium hexafluoride nuclear-pumped laser systems. The principal equipment used included 1.2 MW RF induction heater, a d.c. plasma torch, a uranium tetrafluoride feeder system, and batch-type fluorine/UF6 regeneration systems. Overall objectives were to continue to develop and test materials and handling techniques suitable for use with high-temperature, high-pressure, gaseous UF6; and to continue development of complementary diagnostic instrumentation and measurement techniques to characterize the effluent exhaust gases and residue deposited on the test chamber and exhaust system components. Specific objectives include: a development of a batch-type UF6 regeneration system employing pure high-temperature fluorine; development of a ruggedized time-of-flight mass spectrometer and associated data acquisition system capable of making on-line concentration measurements of the volatile effluent exhaust gas species in a high RF environment and corrosive environment of UF6 and related halide compounds.

Uniform deposition of uranium hexafluoride (UF6): Standardized mass deposits and controlled isotopic ratios using a thermal fluorination method.

PubMed

McNamara, Bruce K; O'Hara, Matthew J; Casella, Andrew M; Carter, Jennifer C; Addleman, R Shane; MacFarlan, Paul J

2016-07-01

We report a convenient method for the generation of volatile uranium hexafluoride (UF6) from solid uranium oxides and other U compounds, followed by uniform deposition of low levels of UF6 onto sampling coupons. Under laminar flow conditions, UF6 is shown to interact with surfaces within a fixed reactor geometry to a highly predictable degree. We demonstrate the preparation of U deposits that range between approximately 0.01 and 500ngcm(-2). The data suggest the method can be extended to creating depositions at the sub-picogramcm(-2) level. The isotopic composition of the deposits can be customized by selection of the U source materials and we demonstrate a layering technique whereby two U solids, each with a different isotopic composition, are employed to form successive layers of UF6 on a surface. The result is an ultra-thin deposit that bears an isotopic signature that is a composite of the two U sources. The reported deposition method has direct application to the development of unique analytical standards for nuclear safeguards and forensics. Further, the method allows access to very low atomic or molecular coverages of surfaces. Copyright © 2016 Elsevier B.V. All rights reserved.

Gaseous fuel reactors for power systems

NASA Technical Reports Server (NTRS)

Kendall, J. S.; Rodgers, R. J.

1977-01-01

Gaseous-fuel nuclear reactors have significant advantages as energy sources for closed-cycle power systems. The advantages arise from the removal of temperature limits associated with conventional reactor fuel elements, the wide variety of methods of extracting energy from fissioning gases, and inherent low fissile and fission product in-core inventory due to continuous fuel reprocessing. Example power cycles and their general performance characteristics are discussed. Efficiencies of gaseous fuel reactor systems are shown to be high with resulting minimal environmental effects. A technical overview of the NASA-funded research program in gaseous fuel reactors is described and results of recent tests of uranium hexafluoride (UF6)-fueled critical assemblies are presented.

Analysis of solid uranium samples using a small mass spectrometer

NASA Astrophysics Data System (ADS)

Kahr, Michael S.; Abney, Kent D.; Olivares, José A.

2001-07-01

A mass spectrometer for isotopic analysis of solid uranium samples has been constructed and evaluated. This system employs the fluorinating agent chlorine trifluoride (ClF 3) to convert solid uranium samples into their volatile uranium hexafluorides (UF 6). The majority of unwanted gaseous byproducts and remaining ClF 3 are removed from the sample vessel by condensing the UF 6 and then pumping away the unwanted gases. The UF 6 gas is then introduced into a quadrupole mass spectrometer and ionized by electron impact ionization. The doubly charged bare metal uranium ion (U 2+) is used to determine the U 235/U 238 isotopic ratio. Precision and accuracy for several isotopic standards were found to be better than 12%, without further calibration of the system. The analysis can be completed in 25 min from sample loading, to UF 6 reaction, to mass spectral analysis. The method is amenable to uranium solid matrices, and other actinides.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fincke, J.R.; Swank, W.D.; Haggard, D.C.

This paper describes the experimental demonstration of a process for the direct plasma reduction of depleted uranium hexafluoride to uranium metal. The process exploits the large departures from equilibrium that can be achieved in the rapid supersonic expansion of a totally dissociated and partially ionized mixture of UF{sub 6}, Ar, He, and H{sub 2}. The process is based on the rapid condensation of subcooled uranium vapor and the relatively slow rate of back reaction between metallic uranium and HF to F{sub 2} to reform stable fluorides. The high translational velocities and rapid cooling result in an overpopulation of atomic hydrogenmore » which persists throughout the expansion process. Atomic hydrogen shifts the equilibrium composition by inhibiting the reformation of uranium-fluorine compounds. This process has the potential to reduce the cost of reducing UF{sub 6} to uranium metal with the added benefit of being a virtually waste free process. The dry HF produced is a commodity which has industrial value.« less

Public Scoping Meeting Materials

Science.gov Websites

project and web site. your e-mail address Sign Me Up Search: OK Button DUF6 Guide DU Uses DUF6 Management , presentations, and other information from the Conversion EIS Public Scoping Meetings. The following materials ) Management Program 5.97 MB details DUF6 Fact Sheets PDF Icon Overview of Depleted Uranium Hexafluoride

MHD compressor---expander conversion system integrated with GCR inside a deployable reflector

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tuninetti, G.; Botta, E.; Criscuolo, C.

1989-04-20

This work originates from the proposal MHD Compressor-Expander Conversion System Integrated with a GCR Inside a Deployable Reflector''. The proposal concerned an innovative concept of nuclear, closed-cycle MHD converter for power generation on space-based systems in the multi-megawatt range. The basic element of this converter is the Power Conversion Unit (PCU) consisting of a gas core reactor directly coupled to an MHD expansion channel. Integrated with the PCU, a deployable reflector provides reactivity control. The working fluid could be either uranium hexafluoride or a mixture of uranium hexafluoride and helium, added to enhance the heat transfer properties. The original Statementmore » of Work, which concerned the whole conversion system, was subsequently redirected and focused on the basic mechanisms of neutronics, reactivity control, ionization and electrical conductivity in the PCU. Furthermore, the study was required to be inherently generic such that the study was required to be inherently generic such that the analysis an results can be applied to various nuclear reactor and/or MHD channel designs''.« less

Reactions of technetium hexafluoride with nitric acid, nitrosyl fluoride, and nitryl fluoride

NASA Technical Reports Server (NTRS)

Holloway, J. H.; Selig, H.

1970-01-01

Stoichiometry of technetium hexafluoride reactions is studied. Magnetic properties and infrared spectra of reaction products are studied and compared with those of analogous complexes of the hexafluorides of tungsten, rhenium, and osmium.

Laboratory-scale uranium RF plasma confinement experiments

NASA Technical Reports Server (NTRS)

Roman, W. C.

1976-01-01

An experimental investigation was conducted using 80 kW and 1.2 MW RF induction heater facilities to aid in developing the technology necessary for designing a self-critical fissioning uranium plasma core reactor. Pure uranium hexafluoride (UF6) was injected into argon-confined, steady-state, RF-heated plasmas in different uranium plasma confinement tests to investigate the characteristics of plamas core nuclear reactors. The objectives were: (1) to confine as high a density of uranium vapor as possible within the plasma while simultaneously minimizing the uranium compound wall deposition; (2) to develop and test materials and handling techniques suitable for use with high-temperature, high-pressure gaseous UF6; and (3) to develop complementary diagnostic instrumentation and measurement techniques to characterize the uranium plasma and residue deposited on the test chamber components. In all tests, the plasma was a fluid-mechanically-confined vortex-type contained within a fused-silica cylindrical test chamber. The test chamber peripheral wall was 5.7 cm ID by 10 cm long.

Plasma core reactor simulations using RF uranium seeded argon discharges

NASA Technical Reports Server (NTRS)

Roman, W. C.

1975-01-01

An experimental investigation was conducted using the United Technologies Research Center (UTRC) 80 kW and 1.2 MW RF induction heater systems to aid in developing the technology necessary for designing a self-critical fissioning uranium plasma core reactor (PCR). A nonfissioning, steady-state RF-heated argon plasma seeded with pure uranium hexafluoride (UF6) was used. An overall objective was to achieve maximum confinement of uranium vapor within the plasma while simultaneously minimizing the uranium compound wall deposition. Exploratory tests were conducted using the 80 kW RF induction heater with the test chamber at approximately atmospheric pressure and discharge power levels on the order of 10 kW. Four different test chamber flow configurations were tested to permit selection of the configuration offering the best confinement characteristics for subsequent tests at higher pressure and power in the 1.2 MW RF induction heater facility.

Production of sintered porous metal fluoride pellets

DOEpatents

Anderson, L.W.; Stephenson, M.J.

1973-12-25

Porous pellets characterized by a moderately reactive crust and a softer core of higher reactivity are produced by forming agglomerates containing a metal fluoride powder and a selected amount ofwater. The metal fluoride is selected to be sinterable and essentially non-reactive with gaseous fluorinating agents. The agglomerates are contacted with a gaseous fluorinating agent under controlled conditions whereby the heat generated by localized reaction of the agent and water is limited to values effccting bonding by localized sintering. Porous pellets composed of cryolite (Na/sub 3/AlF/sub 6/) can be used to selectively remove trace quantities of niobium pentafluoride from a feed gas consisting predominantly of uranium hexafluoride. (Official Gazette)

Occupational radiation exposure experience: Paducah Gaseous Diffusion Plant

DOE Office of Scientific and Technical Information (OSTI.GOV)

Baker, R.C.

1975-01-01

The potential for significant uranium exposure in gaseous diffusion plants is very low. The potential for significant radiation exposure in uranium hexafluoride manufacturing is very real. Exposures can be controlled to low levels only through the cooperation and commitment of facility management and operating personnel. Exposure control can be adequately monitored by a combination of air analyses, urinalyses, and measurements of internal deposition as obtained by the IVRML. A program based on control of air-borne uranium exposure has maintained the internal dose of the Paducah Gaseous Diffusion Plant workman to less than one-half the RPG dose to the lung (15more » rem/year) and probably to less than one-fourth that dose. (auth)« less

Fluorination process using catalyst

DOEpatents

Hochel, Robert C.; Saturday, Kathy A.

1985-01-01

A process for converting an actinide compound selected from the group consisting of uranium oxides, plutonium oxides, uranium tetrafluorides, plutonium tetrafluorides and mixtures of said oxides and tetrafluorides, to the corresponding volatile actinide hexafluoride by fluorination with a stoichiometric excess of fluorine gas. The improvement involves conducting the fluorination of the plutonium compounds in the presence of a fluoride catalyst selected from the group consisting of CoF.sub.3, AgF.sub.2 and NiF.sub.2, whereby the fluorination is significantly enhanced. The improvement also involves conducting the fluorination of one of the uranium compounds in the presence of a fluoride catalyst selected from the group consisting of CoF.sub.3 and AgF.sub.2, whereby the fluorination is significantly enhanced.

Fluorination process using catalysts

DOEpatents

Hochel, R.C.; Saturday, K.A.

1983-08-25

A process is given for converting an actinide compound selected from the group consisting of uranium oxides, plutonium oxides, uranium tetrafluorides, plutonium tetrafluorides and mixtures of said oxides and tetrafluorides, to the corresponding volatile actinide hexafluoride by fluorination with a stoichiometric excess of fluorine gas. The improvement involves conducting the fluorination of the plutonium compounds in the presence of a fluoride catalyst selected from the group consisting of CoF/sub 3/, AgF/sub 2/ and NiF/sub 2/, whereby the fluorination is significantly enhanced. The improvement also involves conducting the fluorination of one of the uranium compounds in the presence of a fluoride catalyst selected from the group consisting of CoF/sub 3/ and AgF/sub 2/, whereby the fluorination is significantly enhanced.

Gas centrifuge enrichment plants inspection frequency and remote monitoring issues for advanced safeguards implementation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Boyer, Brian David; Erpenbeck, Heather H; Miller, Karen A

2010-09-13

Current safeguards approaches used by the IAEA at gas centrifuge enrichment plants (GCEPs) need enhancement in order to verify declared low enriched uranium (LEU) production, detect undeclared LEU production and detect high enriched uranium (BEU) production with adequate probability using non destructive assay (NDA) techniques. At present inspectors use attended systems, systems needing the presence of an inspector for operation, during inspections to verify the mass and {sup 235}U enrichment of declared cylinders of uranium hexafluoride that are used in the process of enrichment at GCEPs. This paper contains an analysis of how possible improvements in unattended and attended NDAmore » systems including process monitoring and possible on-site destructive analysis (DA) of samples could reduce the uncertainty of the inspector's measurements providing more effective and efficient IAEA GCEPs safeguards. We have also studied a few advanced safeguards systems that could be assembled for unattended operation and the level of performance needed from these systems to provide more effective safeguards. The analysis also considers how short notice random inspections, unannounced inspections (UIs), and the concept of information-driven inspections can affect probability of detection of the diversion of nuclear material when coupled to new GCEPs safeguards regimes augmented with unattended systems. We also explore the effects of system failures and operator tampering on meeting safeguards goals for quantity and timeliness and the measures needed to recover from such failures and anomalies.« less

US Transuranium and Uranium Registries case study on accidental exposure to uranium hexafluoride.

PubMed

Avtandilashvili, Maia; Puncher, Matthew; McComish, Stacey L; Tolmachev, Sergei Y

2015-03-01

The United States Transuranium and Uranium Registries' (USTUR) whole-body donor (Case 1031) was exposed to an acute inhalation of uranium hexafluoride (UF6) produced from an explosion at a uranium processing plant 65 years prior to his death. The USTUR measurements of tissue samples collected at the autopsy indicated long-term retention of inhaled slightly enriched uranium material (0.85% (235)U) in the deep lungs and thoracic lymph nodes. In the present study, the authors combined the tissue measurement results with historical bioassay data, and analysed them with International Commission on Radiological Protection (ICRP) respiratory tract models and the ICRP Publication 69 systemic model for uranium using maximum likelihood and Bayesian statistical methods. The purpose of the analysis was to estimate intakes and model parameter values that best describe the data, and evaluate their effect on dose assessment. The maximum likelihood analysis, which used the ICRP Publication 66 human respiratory tract model, resulted in a point estimate of 79 mg of uranium for the occupational intake composed of 86% soluble, type F material and 14% insoluble, type S material. For the Bayesian approach, the authors applied the Markov Chain Monte Carlo method, but this time used the revised human respiratory tract model, which is currently being used by ICRP to calculate new dose coefficients for workers. The Bayesian analysis estimated that the mean uranium intake was 160 mg, and calculated the case-specific lung dissolution parameters with their associated uncertainties. The parameters were consistent with the inhaled uranium material being predominantly soluble with a small but significant insoluble component. The 95% posterior range of the rapid dissolution fraction (the fraction of deposited material that is absorbed to blood rapidly) was 0.12 to 0.91 with a median of 0.37. The remaining fraction was absorbed slowly, with a 95% range of 0.000 22 d(-1) to 0.000 36 d(-1) and a median of 0.000 31 d(-1). The effective dose per unit intake calculated using the dissolution parameters derived from the maximum likelihood and the Bayesian analyses was higher than the current ICRP dose coefficient for type F uranium by a factor of 2 or 7, respectively; the higher value of the latter was due to use of the revised respiratory tract model. The dissolution parameter values obtained here may be more appropriate to use for radiation protection purposes when individuals are exposed to a UF6 mixture that contains an insoluble uranium component.

Feasibility of a Fieldable Mass Spectrometer FY 2015 Year-end Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Barinaga, Charles J.; Hager, George J.; Hoegg, Edward D.

Currently, the International Atomic Energy Agency (IAEA) monitors the production of enriched uranium hexafluoride (UF 6) at declared facilities by collecting a few grams of product in sample tubes that are then sent to central laboratories for processing and isotope ratio analysis by thermal ionization mass spectrometry. Analysis of results may not be available for some time after collection. In addition, new shipping regulations will make it more difficult to transport this amount of UF 6 to a laboratory. The IAEA is interested in an isotope ratio technique for uranium in UF 6 that can be moved to and operatedmore » at the enrichment facility itself. This report covers the tasks and activities of the Feasibility of a Fieldable Mass Spectrometer Project for FY 2015, which investigates the feasibility of an in-field isotope ratio technique— the forward deployment of a technique to the non-laboratory situation of a protected room with power and heat at the facility of interest. A variety of nontraditional elemental ionization techniques were considered. It was determined that only two of these should be moved forward for testing with the candidate in-field mass spectrometer and with the adsorbed UF 6 sample types.« less

Vapor core propulsion reactors

NASA Technical Reports Server (NTRS)

Diaz, Nils J.

1991-01-01

Many research issues were addressed. For example, it became obvious that uranium tetrafluoride (UF4) is a most preferred fuel over uranium hexafluoride (UF6). UF4 has a very attractive vaporization point (1 atm at 1800 K). Materials compatible with UF4 were looked at, like tungsten, molybdenum, rhenium, carbon. It was found that in the molten state, UF4 and uranium attacked most everything, but in the vapor state they are not that bad. Compatible materials were identified for both the liquid and vapor states. A series of analyses were established to determine how the cavity should be designed. A series of experiments were performed to determine the properties of the fluid, including enhancement of the electrical conductivity of the system. CFD's and experimental programs are available that deal with most of the major issues.

Feasibility Study on the Use of On-line Multivariate Statistical Process Control for Safeguards Applications in Natural Uranium Conversion Plants

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ladd-Lively, Jennifer L

2014-01-01

The objective of this work was to determine the feasibility of using on-line multivariate statistical process control (MSPC) for safeguards applications in natural uranium conversion plants. Multivariate statistical process control is commonly used throughout industry for the detection of faults. For safeguards applications in uranium conversion plants, faults could include the diversion of intermediate products such as uranium dioxide, uranium tetrafluoride, and uranium hexafluoride. This study was limited to a 100 metric ton of uranium (MTU) per year natural uranium conversion plant (NUCP) using the wet solvent extraction method for the purification of uranium ore concentrate. A key component inmore » the multivariate statistical methodology is the Principal Component Analysis (PCA) approach for the analysis of data, development of the base case model, and evaluation of future operations. The PCA approach was implemented through the use of singular value decomposition of the data matrix where the data matrix represents normal operation of the plant. Component mole balances were used to model each of the process units in the NUCP. However, this approach could be applied to any data set. The monitoring framework developed in this research could be used to determine whether or not a diversion of material has occurred at an NUCP as part of an International Atomic Energy Agency (IAEA) safeguards system. This approach can be used to identify the key monitoring locations, as well as locations where monitoring is unimportant. Detection limits at the key monitoring locations can also be established using this technique. Several faulty scenarios were developed to test the monitoring framework after the base case or normal operating conditions of the PCA model were established. In all of the scenarios, the monitoring framework was able to detect the fault. Overall this study was successful at meeting the stated objective.« less

On the Use of Thermal NF3 as the Fluorination and Oxidation Agent in Treatment of Used Nuclear Fuels

DOE Office of Scientific and Technical Information (OSTI.GOV)

Scheele, Randall D.; McNamara, Bruce K.; Casella, Andrew M.

2012-05-01

This paper presents results of our investigation on the use of nitrogen trifluoride as the fluorination or fluorination/oxidation agent for use in a process for separating valuable constituents from used nuclear fuels by employing the volatility of many transition metal and actinide fluorides. Nitrogen trifluoride is less chemically and reactively hazardous than the hazardous and aggressive fluorinating agents used to prepare uranium hexafluoride and considered for fluoride volatility based nuclear fuels reprocessing. In addition, nitrogen trifluoride’s less aggressive character may be used to separate the volatile fluorides from used fuel and from themselves based on the fluorination reaction’s temperature sensitivitymore » (thermal tunability) rather than relying on differences in sublimation/boiling temperature and sorbents. Our thermodynamic calculations found that nitrogen trifluoride has the potential to produce volatile fission product and actinide fluorides from candidate oxides and metals. Our simultaneous thermogravimetric and differential thermal analyses found that the oxides of lanthanum, cerium, rhodium, and plutonium fluorinated but did not form volatile fluorides and that depending on temperature volatile fluorides formed from the oxides of niobium, molybdenum, ruthenium, tellurium, uranium, and neptunium. We also demonstrated near-quantitative removal of uranium from plutonium in a mixed oxide.« less

Benchmark gas core critical experiment.

NASA Technical Reports Server (NTRS)

Kunze, J. F.; Lofthouse, J. H.; Cooper, C. G.; Hyland, R. E.

1972-01-01

A critical experiment with spherical symmetry has been conducted on the gas core nuclear reactor concept. The nonspherical perturbations in the experiment were evaluated experimentally and produce corrections to the observed eigenvalue of approximately 1% delta k. The reactor consisted of a low density, central uranium hexafluoride gaseous core, surrounded by an annulus of void or low density hydrocarbon, which in turn was surrounded with a 97-cm-thick heavy water reflector.

The A. Q. Khan Network: Causes and Implications

DTIC Science & Technology

2005-12-01

relations battle to win popular esteem (and discredit the opposing laboratory) and a competition over responsibilities and resources between the two...technology was, we bought it with money,” she said.194 Bhutto is savvy and self -serving politician, whose statements should be taken with a grain...Pakistan, which then converted it and transferred it Libya.226 B.S.A. Tahir told Malaysian investigators that uranium hexafluoride was shipped

Effect of tank geometry on its average performance

NASA Astrophysics Data System (ADS)

Orlov, Aleksey A.; Tsimbalyuk, Alexandr F.; Malyugin, Roman V.; Leontieva, Daria A.; Kotelnikova, Alexandra A.

2018-03-01

The mathematical model of non-stationary filling of vertical submerged tanks with gaseous uranium hexafluoride is presented in the paper. There are calculations of the average productivity, heat exchange area, and filling time of various volumes tanks with smooth inner walls depending on their "height : radius" ratio as well as the average productivity, degree, and filling time of horizontal ribbing tank with volume 6.10-2 m3 with change central hole diameter of the ribs. It has been shown that the growth of "height / radius" ratio in tanks with smooth inner walls up to the limiting values allows significantly increasing tank average productivity and reducing its filling time. Growth of H/R ratio of tank with volume 1.0 m3 to the limiting values (in comparison with the standard tank having H/R equal 3.49) augments tank productivity by 23.5 % and the heat exchange area by 20%. Besides, we have demonstrated that maximum average productivity and a minimum filling time are reached for the tank with volume 6.10-2 m3 having central hole diameter of horizontal ribs 6.4.10-2 m.

The UO2 ex-ADU powder preparation and pellet sintering for optimum efficiency: experimental and modeling studies

NASA Astrophysics Data System (ADS)

Hung, Nguyen Trong; Thuan, Le Ba; Van Tung, Nguyen; Thuy, Nguyen Thanh; Lee, Jin-Young; Jyothi, Rajesh Kumar

2017-12-01

The UO2 nuclear fuel pellet process for light water reactors (LWR) includes the conversion of uranium hexafluoride (UF6) into UO2 powder and the fabrication of UO2 pellets from such UO2 powder. In the paper, studies on UO2 pellet process from ammonium diuranate-derived uranium dioxide powder (UO2 ex-ADU powder) were reported. The UO2 ex-ADU powders were converted from ADU at various temperatures of 973 K, 1023 K and 1073 K and then UO2 pellets prepared from the powders were sintered at temperatures of 1923 K, 1973 K and 2023 K for times of 4 h, 6 h and 8 h. Response surface methodology (RSM) based on quadratic central composite design (CCD) type of face centered (CCF) improved by Box and Hunter was used to model the UO2 pellet process, using MODDE 5.0 software as an assessing tool. On the base of the proposed model, the relationship between the technological parameters and density of the UO2 pellet product was suggested to control the UO2 ex-ADU pellet process as desired levels.

Tc-99 Decontamination From Heat Treated Gaseous Diffusion Membrane -Phase I

DOE Office of Scientific and Technical Information (OSTI.GOV)

Oji, L.; Wilmarth, B.; Restivo, M.

2017-03-13

Uranium gaseous diffusion cascades represent a significant environmental challenge to dismantle, containerize and dispose as low-level radioactive waste. Baseline technologies rely on manual manipulations involving direct access to technetium-contaminated piping and materials. There is a potential to utilize novel thermal decontamination technologies to remove the technetium and allow for on-site disposal of the very large uranium converters. Technetium entered these gaseous diffusion cascades as a hexafluoride complex in the same fashion as uranium. Technetium, as the isotope Tc-99, is an impurity that follows uranium in the first cycle of the Plutonium and Uranium Extraction (PUREX) process. The technetium speciation ormore » exact form in the gas diffusion cascades is not well defined. Several forms of Tc-99 compounds, mostly the fluorinated technetium compounds with varying degrees of volatility have been speculated by the scientific community to be present in these cascades. Therefore, there may be a possibility of using thermal desorption, which is independent of the technetium oxidation states, to perform an in situ removal of the technetium as a volatile species and trap the radionuclide on sorbent traps which could be disposed as low-level waste.« less

Recent measurements concerning uranium hexafluoride-electron collision processes

NASA Technical Reports Server (NTRS)

Trajmar, S.; Chutjian, A.; Srivastava, S.; Williams, W.; Cartwright, D. C.

1976-01-01

Scattering of electrons by UF6 molecules was studied at impact energies ranging from 5 to 100 eV and momentum transfer, elastic and inelastic scattering cross sections were determined. The measurements also yielded spectroscopic information which made possible to extend the optical absorption cross sections from 2000 angstroms to 435 angstroms. It was found that UF6 is a very strong absorber in the vacuum UV region. No transitions were found to lie below the onset of the optically detected 3.0 eV feature.

Building dismantlement and site remediation at the Apollo Fuel Plant: When is technology the answer?

DOE Office of Scientific and Technical Information (OSTI.GOV)

Walton, L.

1995-01-01

The Apollo fuel plant was located in Pennsylvania on a site known to have been used continuously for stell production from before the Civil War until after World War II. Then the site became a nuclear fuel chemical processing plants. Finally it was used to convert uranium hexafluoride to various oxide fuel forms. After the fuel manufacturing operations were teminated, the processing equipment was partially decontaminated, removed, packaged and shipped to a licensed low-level radioactive waste burial site. The work was completed in 1984. In 1990 a detailed site characterization was initiated to establishe the extent of contamination and tomore » plan the building dismantlement and soil remediation efforts. This article discusses the site characterization and remedial action at the site in the following subsections: characterization; criticality control; mobile containment; soil washing; in-process measurements; and the final outcome of the project.« less

PROCESS FOR TREATING VOLATILE METAL FLUORIDES

DOEpatents

Rudge, A.J.; Lowe, A.J.

1957-10-01

This patent relates to the purification of uranium hexafluoride, made by reacting the metal or its tetrafluoride with fluorine, from the frequently contained traces of hydrofluoric acid. According to the present process, UF/sub 6/ containing as an impurity a small amount of hydrofluoric acid, is treated to remove such impurity by contact with an anhydrous alkali metal fluoride such as sodium fluoride. In this way a non-volatile complex containing hydrofluoric acid and the alkali metal fluoride is formed, and the volatile UF /sub 6/ may then be removed by distillation.

Uranium hexafluoride public risk

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fisher, D.R.; Hui, T.E.; Yurconic, M.

1994-08-01

The limiting value for uranium toxicity in a human being should be based on the concentration of uranium (U) in the kidneys. The threshold for nephrotoxicity appears to lie very near 3 {mu}g U per gram kidney tissue. There does not appear to be strong scientific support for any other improved estimate, either higher or lower than this, of the threshold for uranium nephrotoxicity in a human being. The value 3 {mu}g U per gram kidney is the concentration that results from a single intake of about 30 mg soluble uranium by inhalation (assuming the metabolism of a standard person).more » The concentration of uranium continues to increase in the kidneys after long-term, continuous (or chronic) exposure. After chronic intakes of soluble uranium by workers at the rate of 10 mg U per week, the concentration of uranium in the kidneys approaches and may even exceed the nephrotoxic limit of 3 {mu}g U per gram kidney tissue. Precise values of the kidney concentration depend on the biokinetic model and model parameters assumed for such a calculation. Since it is possible for the concentration of uranium in the kidneys to exceed 3 {mu}g per gram tissue at an intake rate of 10 mg U per week over long periods of time, we believe that the kidneys are protected from injury when intakes of soluble uranium at the rate of 10 mg U per week do not continue for more than two consecutive weeks. For long-term, continuous occupational exposure to low-level, soluble uranium, we recommend a reduced weekly intake limit of 5 mg uranium to prevent nephrotoxicity in workers. Our analysis shows that the nephrotoxic limit of 3 {mu}g U per gram kidney tissues is not exceeded after long-term, continuous uranium intake at the intake rate of 5 mg soluble uranium per week.« less

Developments of solid materials for UF 6 sampling

DOE Office of Scientific and Technical Information (OSTI.GOV)

Smith, Nicholas; Hebden, Andrew; Savina, Joseph

2017-11-15

This project demonstrated that a device using majority Commercial Off the Shelf (COTS) components could be used to collect uranium hexafluoride samples safely from gaseous or solid sources. The device was based on the successful Cristallini method developed by ABACC over the past 10 years. The system was designed to capture and store the UF 6 as an inert fluoride salt to ease transportation regulations. In addition, the method was considerably faster than traditional cryogenic methods, collected enough material to perform analyses without undue waste, and could be used either inside a facility or in the storage yard.

Status Report on the Passive Neutron Enrichment Meter (PNEM) for UF6 Cylinder Assay

DOE Office of Scientific and Technical Information (OSTI.GOV)

Miller, Karen A.; Swinhoe, Martyn T.; Menlove, Howard O.

2012-05-02

The Passive Neutron Enrichment Meter (PNEM) is a nondestructive assay (NDA) system being developed at Los Alamos National Laboratory (LANL). It was designed to determine {sup 235}U mass and enrichment of uranium hexafluoride (UF{sub 6}) in product, feed, and tails cylinders (i.e., 30B and 48Y cylinders). These cylinders are found in the nuclear fuel cycle at uranium conversion, enrichment, and fuel fabrication facilities. The PNEM is a {sup 3}He-based neutron detection system that consists of two briefcase-sized detector pods. A photograph of the system during characterization at LANL is shown in Fig. 1. Several signatures are currently being studied tomore » determine the most effective measurement and data reduction technique for unfolding {sup 235}U mass and enrichment. The system collects total neutron and coincidence data for both bare and cadmium-covered detector pods. The measurement concept grew out of the success of the Uranium Cylinder Assay System (UCAS), which is an operator system at Rokkasho Enrichment Plant (REP) that uses total neutron counting to determine {sup 235}U mass in UF{sub 6} cylinders. The PNEM system was designed with higher efficiency than the UCAS in order to add coincidence counting functionality for the enrichment determination. A photograph of the UCAS with a 48Y cylinder at REP is shown in Fig. 2, and the calibration measurement data for 30B product and 48Y feed and tails cylinders is shown in Fig. 3. The data was collected in a low-background environment, meaning there is very little scatter in the data. The PNEM measurement concept was first presented at the 2010 Institute of Nuclear Materials Management (INMM) Annual Meeting. The physics design and uncertainty analysis were presented at the 2010 International Atomic Energy Agency (IAEA) Safeguards Symposium, and the mechanical and electrical designs and characterization measurements were published in the ESARDA Bulletin in 2011.« less

COLD TRAPS

DOEpatents

Thompson, W.I.

1958-09-30

A cold trap is presented for removing a condensable component from a gas mixture by cooling. It consists of a shell, the exterior surface of which is chilled by a refrigerant, and conductive fins welded inside the shell to condense the gas, and distribute the condensate evenly throughout the length of the trap, so that the trap may function until it becomes completely filled with the condensed solid. The contents may then be removed as either a gas or as a liquid by heating the trap. This device has particuinr use as a means for removing uranium hexafluoride from the gaseous diffusion separation process during equipment breakdown and repair periods.

Application of gaseous core reactors for transmutation of nuclear waste

NASA Technical Reports Server (NTRS)

Schnitzler, B. G.; Paternoster, R. R.; Schneider, R. T.

1976-01-01

An acceptable management scheme for high-level radioactive waste is vital to the nuclear industry. The hazard potential of the trans-uranic actinides and of key fission products is high due to their nuclear activity and/or chemical toxicity. Of particular concern are the very long-lived nuclides whose hazard potential remains high for hundreds of thousands of years. Neutron induced transmutation offers a promising technique for the treatment of problem wastes. Transmutation is unique as a waste management scheme in that it offers the potential for "destruction" of the hazardous nuclides by conversion to non-hazardous or more manageable nuclides. The transmutation potential of a thermal spectrum uranium hexafluoride fueled cavity reactor was examined. Initial studies focused on a heavy water moderated cavity reactor fueled with 5% enriched U-235-F6 and operating with an average thermal flux of 6 times 10 to the 14th power neutrons/sq cm-sec. The isotopes considered for transmutation were I-129, Am-241, Am-242m, Am-243, Cm-243, Cm-244, Cm-245, and Cm-246.

Study of Chemical Changes in Uranium Oxyfluoride Particles Progress Report March - October 2009

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kips, R; Kristo, M; Hutcheon, I

2009-11-22

Nuclear forensics relies on the analysis of certain sample characteristics to determine the origin and history of a nuclear material. In the specific case of uranium enrichment facilities, it is the release of trace amounts of uranium hexafluoride (UF{sub 6}) gas - used for the enrichment of uranium - that leaves a process-characteristic fingerprint. When UF{sub 6} gas interacts with atmospheric moisture, uranium oxyfluoride particles or particle agglomerates are formed with sizes ranging from several microns down to a few tens of nanometers. These particles are routinely collected by safeguards organizations, such as the International Atomic Energy Agency (IAEA), allowingmore » them to verify whether a facility is compliant with its declarations. Spectrometric analysis of uranium particles from UF{sub 6} hydrolysis has revealed the presence of both particles that contain fluorine, and particles that do not. It is therefore assumed that uranium oxyfluoride is unstable, and decomposes to form uranium oxide. Understanding the rate of fluorine loss in uranium oxyfluoride particles, and the parameters that control it, may therefore contribute to placing boundaries on the particle's exposure time in the environment. Expressly for the purpose of this study, we prepared a set of uranium oxyfluoride particles at the Institute for Reference Materials and Measurements (EU-JRC-IRMM) from a static release of UF{sub 6} in a humid atmosphere. The majority of the samples was stored in controlled temperature, humidity and lighting conditions. Single particles were characterized by a suite of micro-analytical techniques, including NanoSIMS, micro-Raman spectrometry (MRS), scanning (SEM) and transmission (TEM) electron microscopy, energy-dispersive X-ray spectrometry (EDX) and focused ion beam (FIB). The small particle size was found to be the main analytical challenge. The relative amount of fluorine, as well as the particle chemical composition and morphology were determined at different stages in the ageing process, and immediately after preparation. This report summarizes our most recent findings for each of the analytical techniques listed above, and provides an outlook on what remains to be resolved. Additional spectroscopic and mass spectrometric measurements were carried out at Pacific Northwest National Laboratory, but are not included in this summary.« less

Bearing assembly and the like for use in corrosive and non-corrosive atmospheres

DOEpatents

Mashburn, Douglas N.; Woodall, Harold C.; Wright, Ralph R.

1979-01-01

This invention relates to a novel machine element characterized by mutually rubbing surfaces which are composed of dissimilar materials having high hardness, a low coefficient of friction, and resistance to corrosion by halogen-containing atmospheres. As exemplified by the preferred embodiment for use in gaseous UF.sub.6, the rubbing surfaces are chemically deposited nickel and anodized aluminum. These surfaces permit jam-free operation despite long-term exposure to UF.sub.6. Preferably, both surfaces have a hardness of at least about 500 HV.sub.100 on the Vickers hardness scale, and preferably the anodized-aluminum surface is of a type having comparatively little tendency to sorb uranium hexafluoride.

A Monte Carlo Analysis of Weight Data from UF 6 Cylinder Feed and Withdrawal Stations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Garner, James R; Whitaker, J Michael

2015-01-01

As the number of nuclear facilities handling uranium hexafluoride (UF 6) cylinders (e.g., UF 6 production, enrichment, and fuel fabrication) increase in number and throughput, more automated safeguards measures will likely be needed to enable the International Atomic Energy Agency (IAEA) to achieve its safeguards objectives in a fiscally constrained environment. Monitoring the process data from the load cells built into the cylinder feed and withdrawal (F/W) stations (i.e., cylinder weight data) can significantly increase the IAEA’s ability to efficiently achieve the fundamental safeguards task of confirming operations as declared (i.e., no undeclared activities). Researchers at the Oak Ridge Nationalmore » Laboratory, Los Alamos National Laboratory, the Joint Research Center (in Ispra, Italy), and University of Glasgow are investigating how this weight data can be used for IAEA safeguards purposes while fully protecting the operator’s proprietary and sensitive information related to operations. A key question that must be resolved is, what is the necessary frequency of recording data from the process F/W stations to achieve safeguards objectives? This paper summarizes Monte Carlo simulations of typical feed, product, and tails withdrawal cycles and evaluates longer sampling frequencies to determine the expected errors caused by low-frequency sampling and its impact on material balance calculations.« less

Compilation of Requirements for Safe Handling of Fluorine and Fluorine-Containing Products of Uranium Hexafluoride Conversion

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ferrada, J.J.

2000-04-03

Public Law (PL) 105-204 requires the U.S. Department of Energy to develop a plan for inclusion in the fiscal year 2000 budget for conversion of the Department's stockpile of depleted uranium hexafluoride (DUF{sub 6}) to a more stable form over an extended period. The conversion process into a more stable form will produce fluorine compounds (e.g., elemental fluorine or hydrofluoric acid) that need to be handled safely. This document compiles the requirements necessary to handle these materials within health and safety standards, which may apply in order to ensure protection of the environment and the safety and health of workersmore » and the public. Fluorine is a pale-yellow gas with a pungent, irritating odor. It is the most reactive nonmetal and will react vigorously with most oxidizable substances at room temperature, frequently with ignition. Fluorine is a severe irritant of the eyes, mucous membranes, skin, and lungs. In humans, the inhalation of high concentrations causes laryngeal spasm and broncospasms, followed by the delayed onset of pulmonary edema. At sublethal levels, severe local irritation and laryngeal spasm will preclude voluntary exposure to high concentrations, unless the individual is trapped or incapacitated. A blast of fluorine gas on the shaved skin of a rabbit causes a second degree burn. Lower concentrations cause severe burns of insidious onset, resulting in ulceration, similar to the effects produced by hydrogen fluoride. Hydrofluoric acid is a colorless, fuming liquid or gas with a pungent odor. It is soluble in water with release of heat. Ingestion of an estimated 1.5 grams produced sudden death without gross pathological damage. Repeated ingestion of small amounts resulted in moderately advanced hardening of the bones. Contact of skin with anhydrous liquid produces severe burns. Inhalation of AHA or aqueous hydrofluoric acid mist or vapors can cause severe respiratory tract irritation that may be fatal. Based on the extreme chemical properties of these chemicals as noted above, fluorine or fluorine compounds must be handled appropriately within the boundaries of many safety requirements for the protection of the environment and the public. This report analyzes the safety requirements that regulatory agencies have issued to handle fluorine or fluorine compounds and lists them in Table 1. Table 1 lists the source of the requirements, the specific section of the source document, and a brief description of the requirements.« less

Characterization of the Kinetics of NF3-Fluorination of NpO2

DOE Office of Scientific and Technical Information (OSTI.GOV)

Casella, Andrew M.; Scheele, Randall D.; McNamara, Bruce K.

2015-12-23

The exploitation of selected actinide and fission product fluoride volatilities has long been considered as a potentially attractive compact method for recycling used nuclear fuels to avoid generating the large volumes of radioactive waste arising from aqueous reprocessing [1-7]. The most developed process uses the aggressive and hazardous fluorinating agents hydrogen fluoride (HF) and/or molecular fluorine (F2) at high temperatures to volatilize the greatest fraction of the used nuclear fuel into a single gas stream. The volatilized fluorides are subsequently separated using a series of fractionation and condensation columns to recover the valuable fuel constituents and fission products. In pursuitmore » of a safer and less complicated approach, we investigated an alternative fluoride volatility-based process using the less hazardous fluorinating agent nitrogen trifluoride (NF3) and leveraging its less aggressive nature to selectively evolve fission product and actinide fluorides from the solid phase based on their reaction temperatures into a single recycle stream [8-15]. In this approach, successive isothermal treatments using NF3 will first evolve the more thermally susceptible used nuclear fuel constituents leaving the other constituents in the residual solids until subsequent isothermal temperature treatments cause these others to volatilize. During investigation of this process, individual neat used fuel components were treated with isothermal NF3 in an attempt to characterize the kinetics of each fluorination reaction to provide input into the design of a new volatile fluoride separations approach. In these directed investigations, complex behavior was observed between NF3 and certain solid reactants such as the actinide oxides of uranium, plutonium, and neptunium. Given the similar thermal reaction susceptibilities of neptunium oxide (NpO2) and uranium dioxide (UO2) and the importance of Np and U, we initially focused our efforts on determining the reaction kinetic parameters for NpO2. Characterizing the NF3 fluorination of NpO2 using established models for gas-solid reactions [16] proved unsuccessful so we developed a series of successive fundamental reaction mechanisms to characterize the observed successive fluorination reactions leading to production of the volatile neptunium hexafluoride (NpF6).« less

Kerr-McGee and the NRC: from Indian country to Silkwood to Gore.

PubMed

Baer, H

1990-01-01

By focusing upon the Nuclear Regulatory Commission's appraisal of the Kerr-McGee Corporation's safety record in the Four Corners area and at two facilities in Oklahoma, this article examines the political economy of nuclear regulation in American society. Particular attention is given to the agency's response to intervenor groups which protested various operations at Kerr-McGee facility in Gore, Oklahoma, both prior to and following the accidental rupture of a cylinder containing uranium hexafluoride. Despite a consistent record of violations and nuclear mishaps by Kerr-McGee, the Nuclear Regulatory Commission permitted the company to essentially monitor its own activities. Rather than protecting workers and the public from the hazards of the nuclear industry, state regulation attempts to legitimize and defuse public opposition to its endeavors.

Biological assessment of the effects of construction and operation of adepleted uranium hexafluoride conversion facility at the Portsmouth, Ohio,site.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Van Lonkhuyzen, R.

The U.S. Department of Energy (DOE) Depleted Uranium Hexafluoride (DUF{sub 6}) Management Program evaluated alternatives for managing its inventory of DUF{sub 6} and issued the ''Programmatic Environmental Impact Statement for Alternative Strategies for the Long-Term Management and Use of Depleted Uranium Hexafluoride'' (DUF{sub 6} PEIS) in April 1999 (DOE 1999). The DUF{sub 6} inventory is stored in cylinders at three DOE sites: Paducah, Kentucky; Portsmouth, Ohio; and East Tennessee Technology Park (ETTP), near Oak Ridge, Tennessee. In the Record of Decision for the DUF{sub 6} PEIS, DOE stated its decision to promptly convert the DUF{sub 6} inventory to a moremore » stable chemical form. Subsequently, the U.S. Congress passed, and the President signed, the ''2002 Supplemental Appropriations Act for Further Recovery from and Response to Terrorist Attacks on the United States'' (Public Law No. 107-206). This law stipulated in part that, within 30 days of enactment, DOE must award a contract for the design, construction, and operation of a DUF{sub 6} conversion plant at the Department's Paducah, Kentucky, and Portsmouth, Ohio, sites, and for the shipment of DUF{sub 6} cylinders stored at ETTP to the Portsmouth site for conversion. This biological assessment (BA) has been prepared by DOE, pursuant to the National Environmental Policy Act of 1969 and the Endangered Species Act of 1974, to evaluate potential impacts to federally listed species from the construction and operation of a conversion facility at the DOE Portsmouth site. The Indiana bat is known to occur in the area of the Portsmouth site and may potentially occur on the site during spring or summer. Evaluations of the Portsmouth site indicated that most of the site was found to have poor summer habitat for the Indiana bat because of the small size, isolation, and insufficient maturity of the few woodlands on the site. Potential summer habitat for the Indiana bat was identified outside the developed area bounded by Perimeter Road, within the corridors along Little Beaver Creek, the Northwest Tributary stream, and a wooded area east of the X-100 facility. However, no Indiana bats were collected during surveys of these areas in 1994 and 1996. Locations A, B, and C do not support suitable habitat for the Indiana bat and would be unlikely to be used by Indiana bats. Indiana bat habitat also does not occur at Proposed Areas 1 and 2. Although Locations A and C contain small wooded areas, the small size and lack of suitable maturity of these areas indicate that they would provide poor habitat for Indiana bats. Trees that may be removed during construction would not be expected to be used for summer roosting by Indiana bats. Disturbance of Indiana bats potentially roosting or foraging in the vicinity of the facility during operations would be very unlikely, and any disturbance would be expected to be negligible. On the basis of these considerations, DOE concludes that the proposed action is not likely to adversely affect the Indiana bat. No critical habitat exists for this species in the action area. Although the timber rattlesnake occurs in the vicinity of the Portsmouth site, it has not been observed on the site. In addition, habitat for the timber rattlesnake is not present on the Portsmouth site. Therefore, DOE concludes that the proposed action would not affect the timber rattlesnake.« less

Pulmonary effects of acute exposure to degradation products of sulphur hexafluoride during electrical cable repair work.

PubMed Central

Kraut, A; Lilis, R

1990-01-01

Six electrical workers accidentally exposed to degradation products of sulphur hexafluoride (SF6) during electrical repair work were followed up for one year. One degradation product, sulphur tetrafluoride (SF4), was identified from worksite measurements. Unprotected exposure in an underground enclosed space occurred for six hours over a 12 hour period. Initial symptoms included shortness of breath, chest tightness, productive cough, nose and eye irritation, headache, fatigue, nausea, and vomiting. Symptoms subsided when exposure was interrupted during attempts to identify the cause of the problem. Although exposure ended after several hours, four workers remained symptomatic for between one week and one month. Pulmonary radiographic abnormalities included several discrete areas of transitory platelike atelectasis in one worker, and a slight diffuse infiltrate in the left lower lobe of another. One worker showed transient obstructive changes in tests of pulmonary function. Examination at follow up after one year showed no persistent abnormalities. Preliminary data from this paper were presented at the VIIth international pneumoconioses conference. Pittsburgh, PA, August 1988. PMID:2271390

Protocol for Tier 2 Evaluation of Vapor Intrusion at Corrective Action Sites

DTIC Science & Technology

2012-07-01

622.92 600.12 437.08 433.44 411.1 Sulfur Hexafluoride (SF6) by NIOSH 6602 Modified Sulfur Hexafluoride 600 130 380 290 120 370 Notes: 1. VOC and SF6...6602 Modified Sulfur Hexafluoride 2400 2600 24 1500 260 14 18 1000 Notes: 1. VOC and SF6 samples were analyzed by Columbia Analytical Services, Inc. in...NIOSH 6602 Modified Sulfur Hexafluoride 3900 15 1800 1700 24 1600 Notes: 1. VOC and SF6 samples were analyzed by Columbia Analytical Services, Inc. in

An unattended verification station for UF6 cylinders: Field trial findings

NASA Astrophysics Data System (ADS)

Smith, L. E.; Miller, K. A.; McDonald, B. S.; Webster, J. B.; Zalavadia, M. A.; Garner, J. R.; Stewart, S. L.; Branney, S. J.; Todd, L. C.; Deshmukh, N. S.; Nordquist, H. A.; Kulisek, J. A.; Swinhoe, M. T.

2017-12-01

In recent years, the International Atomic Energy Agency (IAEA) has pursued innovative techniques and an integrated suite of safeguards measures to address the verification challenges posed by the front end of the nuclear fuel cycle. Among the unattended instruments currently being explored by the IAEA is an Unattended Cylinder Verification Station (UCVS), which could provide automated, independent verification of the declared relative enrichment, 235U mass, total uranium mass, and identification for all declared uranium hexafluoride cylinders in a facility (e.g., uranium enrichment plants and fuel fabrication plants). Under the auspices of the United States and European Commission Support Programs to the IAEA, a project was undertaken to assess the technical and practical viability of the UCVS concept. The first phase of the UCVS viability study was centered on a long-term field trial of a prototype UCVS system at a fuel fabrication facility. A key outcome of the study was a quantitative performance evaluation of two nondestructive assay (NDA) methods being considered for inclusion in a UCVS: Hybrid Enrichment Verification Array (HEVA), and Passive Neutron Enrichment Meter (PNEM). This paper provides a description of the UCVS prototype design and an overview of the long-term field trial. Analysis results and interpretation are presented with a focus on the performance of PNEM and HEVA for the assay of over 200 "typical" Type 30B cylinders, and the viability of an "NDA Fingerprint" concept as a high-fidelity means to periodically verify that material diversion has not occurred.

TC-99 Decontaminant from heat treated gaseous diffusion membrane -Phase I, Part B

DOE Office of Scientific and Technical Information (OSTI.GOV)

Oji, L.; Restivo, M.; Duignan, M.

2017-11-01

Uranium gaseous diffusion cascades represent a significant environmental challenge to dismantle, containerize and dispose as low-level radioactive waste. Baseline technologies rely on manual manipulations involving direct access to technetium-contaminated piping and materials. There is a potential to utilize novel decontamination technologies to remove the technetium and allow for on-site disposal of the very large uranium converters. Technetium entered these gaseous diffusion cascades as a hexafluoride complex in the same fashion as uranium. Technetium, as the isotope Tc-99, is an impurity that follows uranium in the first cycle of the Plutonium and Uranium Extraction (PUREX) process. The technetium speciation or exactmore » form in the gaseous diffusion cascades is not well defined. Several forms of Tc-99 compounds, mostly the fluorinated technetium compounds with varying degrees of volatility have been speculated by the scientific community to be present in these cascades. Therefore, there may be a possibility of using thermal or leaching desorption, which is independent of the technetium oxidation states, to perform an insitu removal of the technetium as a volatile species and trap the radionuclide on sorbent traps which could be disposed as low-level waste. Based on the positive results of the first part of this work1 the use of steam as a thermal decontamination agent was further explored with a second piece of used barrier material from a different location. This new series of tests included exposing more of the material surface to the flow of high temperature steam through the change in the reactor design, subjecting it to alternating periods of stream and vacuum, as well as determining if a lower temperature steam, i.e., 121°C (250°F) would be effective, too. Along with these methods, one other simpler method involving the leaching of the Tc-99 contaminated barrier material with a 1.0 M aqueous solution of ammonium carbonate, with and without sonication, was evaluated.« less

The emission characteristics of uranium hexafluoride at high temperatures

NASA Technical Reports Server (NTRS)

Krascella, N. L.

1976-01-01

Relative emission measurements were made for UF6/Argon mixtures heated in a plasma torch over a range of temperatures from 800 to about 3600 K over a wavelength range from 80 to 600 nm. Total pressures were varied from 1 to approximately 1.7 atm. Similarly absorption measurements were carried out in the visible region from 420 to 580 nm over a temperature range from about 1000 to 1800 K. Total pressure for these measurements was 1.0 atm. The emission results exhibited relatively no emission at wavelengths below 250 nm over the range of temperatures investigated. At temperatures in excess of 1800 K an additional emission band centered at 310 nm appears and becomes more well defined at higher temperatures. Essentially no pressure effect was observed with respect to emission at pressures up to 1.7 atm.

In Situ NDA Conformation Measurements Performed at Auxiliary Charcoal Bed and Other Main Charcoal Beds After Uranium Removal from Molten Salt Reactor Experiment ACB at Oak Ridge National Laboratory

DOE Office of Scientific and Technical Information (OSTI.GOV)

Haghighi, M. H.; Kring, C. T.; McGehee, J. T.

2002-02-26

The Molten Salt Reactor Experiment (MSRE) site is located in Tennessee, on the U.S. Department of Energy (DOE) Oak Ridge Reservation (ORR). The MSRE was run by Oak Ridge National Laboratory (ORNL) to demonstrate the desirable features of the molten-salt concept in a practical reactor that could be operated safely and reliably. It introduced the idea of a homogeneous reactor using fuel salt media and graphite moderation for power and breeder reactors. The MSRE reactor and associated components are located in cells beneath the floor in the high-bay area of Building 7503. The reactor was operated from June 1965 tomore » December 1969. When the reactor was shut down, fuel salt was drained from the reactor circuit to two drain tanks. A ''clean'' salt was then circulated through the reactor as a decontamination measure and drained to a third drain tank. When operations ceased, the fuel and flush salts were allowed to cool and solidify in the drain tanks. At shutdown, the MSRE facility complex was placed in a surveillance and maintenance program. Beginning in 1987, it was discovered that gaseous uranium (U-233/U-232) hexafluoride (UF6) had moved throughout the MSRE process systems. The UF6 had been generated when radiolysis in the fluorine salts caused the individual constituents to dissociate to their component atoms, including free fluorine. Some of the free fluorine combined with uranium fluorides (UF4) in the salt to produce UF6. UF6 is gaseous at slightly above ambient temperatures; thus, periodic heating of the fuel salts (which was intended to remedy the radiolysis problems) and simple diffusion had allowed the UF6 to move out of the salt and into the process systems of MSRE. One of the systems that UF6 migrated into due to this process was the offgas system which is vented to the MSRE main charcoal beds and MSRE auxiliary charcoal bed (ACB). Recently, the majority of the uranium laden-charcoal material residing within the ACB was safely and successfully removed using the uranium deposit removal system and equipment. After removal a series of NDA measurements was performed to determine the amount of uranium material remaining in the ACB, the amount of uranium material removed from the ACB, and the amount of uranium material remaining in the uranium removal equipment due to removal activities.« less

Shipboard Tests of Halon 1301 Test Gas Simulants

DTIC Science & Technology

1990-08-22

Halon 1301 Test Gas Simulants," Memo Report, Naval Research Laboratory, Washington, DC, 1989. 8. Air Products and Chemicals , Inc ., "SF6 , An... Products and Chemicals , Inc ., "Specialty Gas Material Safety Data Sheet, Sulfur Hexafluoride " Air Products and Chemicals, Specialty Gas Department...34 Washington, DC, 1978. 11. N.I. Sax, Dangerous Properties of Industrial Materials, Sixth Edition, Van Nostrand Reinhold Company, New York, NY, 1984. 12. Air

Circulation system for flowing uranium hexafluoride cavity reactor experiments

NASA Technical Reports Server (NTRS)

Jaminet, J. F.; Kendall, J. S.

1976-01-01

Research related to determining the feasibility of producing continuous power from fissile fuel in the gaseous state is presented. The development of three laboratory-scale flow systems for handling gaseous UF6 at temperatures up to 500 K, pressure up to approximately 40 atm, and continuous flow rates up to approximately 50g/s is presented. A UF6 handling system fabricated for static critical tests currently being conducted is described. The system was designed to supply UF6 to a double-walled aluminum core canister assembly at temperatures between 300 K and 400 K and pressure up to 4 atm. A second UF6 handling system designed to provide a circulating flow of up to 50g/s of gaseous UF6 in a closed-loop through a double-walled aluminum core canister with controlled temperature and pressure is described. Data from flow tests using UF6 and UF6/He mixtures with this system at flow rates up to approximately 12g/s and pressure up to 4 atm are presented. A third UF6 handling system fabricated to provide a continuous flow of UF6 at flow rates up to 5g/s and at pressures up to 40 atm for use in rf-heated, uranium plasma confinement experiments is described.

An unattended verification station for UF 6 cylinders: Field trial findings

DOE PAGES

Smith, L. E.; Miller, K. A.; McDonald, B. S.; ...

2017-08-26

In recent years, the International Atomic Energy Agency (IAEA) has pursued innovative techniques and an integrated suite of safeguards measures to address the verification challenges posed by the front end of the nuclear fuel cycle. Among the unattended instruments currently being explored by the IAEA is an Unattended Cylinder Verification Station (UCVS), which could provide automated, independent verification of the declared relative enrichment, 235U mass, total uranium mass, and identification for all declared uranium hexafluoride cylinders in a facility (e.g., uranium enrichment plants and fuel fabrication plants). Under the auspices of the United States and European Commission Support Programs tomore » the IAEA, a project was undertaken to assess the technical and practical viability of the UCVS concept. The first phase of the UCVS viability study was centered on a long-term field trial of a prototype UCVS system at a fuel fabrication facility. A key outcome of the study was a quantitative performance evaluation of two nondestructive assay (NDA) methods being considered for inclusion in a UCVS: Hybrid Enrichment Verification Array (HEVA), and Passive Neutron Enrichment Meter (PNEM). This paper provides a description of the UCVS prototype design and an overview of the long-term field trial. In conclusion, analysis results and interpretation are presented with a focus on the performance of PNEM and HEVA for the assay of over 200 “typical” Type 30B cylinders, and the viability of an “NDA Fingerprint” concept as a high-fidelity means to periodically verify that material diversion has not occurred.« less

An unattended verification station for UF 6 cylinders: Field trial findings

DOE Office of Scientific and Technical Information (OSTI.GOV)

Smith, L. E.; Miller, K. A.; McDonald, B. S.

In recent years, the International Atomic Energy Agency (IAEA) has pursued innovative techniques and an integrated suite of safeguards measures to address the verification challenges posed by the front end of the nuclear fuel cycle. Among the unattended instruments currently being explored by the IAEA is an Unattended Cylinder Verification Station (UCVS), which could provide automated, independent verification of the declared relative enrichment, 235U mass, total uranium mass, and identification for all declared uranium hexafluoride cylinders in a facility (e.g., uranium enrichment plants and fuel fabrication plants). Under the auspices of the United States and European Commission Support Programs tomore » the IAEA, a project was undertaken to assess the technical and practical viability of the UCVS concept. The first phase of the UCVS viability study was centered on a long-term field trial of a prototype UCVS system at a fuel fabrication facility. A key outcome of the study was a quantitative performance evaluation of two nondestructive assay (NDA) methods being considered for inclusion in a UCVS: Hybrid Enrichment Verification Array (HEVA), and Passive Neutron Enrichment Meter (PNEM). This paper provides a description of the UCVS prototype design and an overview of the long-term field trial. In conclusion, analysis results and interpretation are presented with a focus on the performance of PNEM and HEVA for the assay of over 200 “typical” Type 30B cylinders, and the viability of an “NDA Fingerprint” concept as a high-fidelity means to periodically verify that material diversion has not occurred.« less

RESULTS FROM A DEMONSTRATION OF RF-BASED UF6 CYLINDER ACCOUNTING AND TRACKING SYSTEM INSTALLED AT A USEC FACILITY

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pickett, Chris A; Kovacic, Donald N; Morgan, Jim

Approved industry-standard cylinders are used globally for storing and transporting uranium hexafluoride (UF{sub 6}) at uranium enrichment plants and processing facilities. To verify that no diversion or undeclared production of nuclear material involving UF{sub 6} cylinders at the facility has occurred, the International Atomic Energy Agency (IAEA) conducts periodic, labor-intensive physical inspections to validate facility records, cylinder identities, and cylinder weights. A reliable cylinder monitoring system that would improve overall inspector effectiveness would be a significant improvement to the current international safeguards inspection regime. Such a system could include real-time unattended monitoring of cylinder movements, situation-specific rules-based event detection algorithms,more » and the capability to integrate with other types of safeguards technologies. This type of system could provide timely detection of abnormal operational activities that may be used to ensure more appropriate and efficient responses by the IAEA. A system of this type can reduce the reliance on paper records and have the additional benefit of facilitating domestic safeguards at the facilities at which it is installed. A radio-frequency (RF)-based system designed to track uranium hexafluoride (UF{sub 6}) cylinders during processing operations was designed, assembled, and tested at the United States Enrichment Corporation (USEC) facility in Portsmouth, Ohio, to determine the operational feasibility and durability of RF technology. The overall objective of the effort was to validate the robustness of RF technology for potential use as a future international safeguards tool for tracking UF6 cylinders at uranium-processing facilities. The results to date indicate that RF tags represent a feasible technique for tracking UF{sub 6} cylinders in operating facilities. Additional work will be needed to improve the operational robustness of the tags for repeated autoclave processing and to add tamper-indicating and data authentication features to some of the pertinent system components. Future efforts will focus on these needs along with implementing protocols relevant to IAEA safeguards. The work detailed in this report demonstrates the feasibility of constructing RF devices that can survive the operational rigors associated with the transportation, storage, and processing of UF6 cylinders. The system software specially designed for this project is called Cylinder Accounting and Tracking System (CATS). This report details the elements of the CATS rules-based architecture and its use in safeguards-monitoring and asset-tracking applications. Information is also provided on improvements needed to make the technology ready, as well as options for improving the safeguards aspects of the technology. The report also includes feedback from personnel involved in the testing, as well as individuals who could utilize an RF-based system in supporting the performance of their work. The system software was set up to support a Mailbox declaration, where a declaration can be made either before or after cylinder movements take place. When the declaration is made before cylinders move, the operators must enter this information into CATS. If the IAEA then shows up unexpectedly at the facility, they can see how closely the operational condition matches the declaration. If the declaration is made after the cylinders move, this provides greater operational flexibility when schedules are interrupted or are changed, by allowing operators to declare what moves have been completed. The IAEA can then compare where cylinders are with where CATS or the system says they are located. The ability of CATS to automatically generate Mailbox declarations is seen by the authors as a desirable feature. The Mailbox approach is accepted by the IAEA but has not been widely implemented (and never in enrichment facilities). During the course of this project, we have incorporated alternative methods for implementation.« less

Initial conceptual design study of self-critical nuclear pumped laser systems

NASA Technical Reports Server (NTRS)

Rodgers, R. J.

1979-01-01

An analytical study of self-critical nuclear pumped laser system concepts was performed. Primary emphasis was placed on reactor concepts employing gaseous uranium hexafluoride (UF6) as the fissionable material. Relationships were developed between the key reactor design parameters including reactor power level, critical mass, neutron flux level, reactor size, operating pressure, and UF6 optical properties. The results were used to select a reference conceptual laser system configuration. In the reference configuration, the 3.2 m cubed lasing volume is surrounded by a graphite internal moderator and a region of heavy water. Results of neutronics calculations yield a critical mass of 4.9 U(235) in the form (235)UF6. The configuration appears capable of operating in a continuous steady-state mode. The average gas temperature in the core is 600 K and the UF6 partial pressure within the lasing volume is 0.34 atm.

New insight into UO 2F 2 particulate structure by micro-Raman spectroscopy

DOE PAGES

Stefaniak, Elzbieta A.; Darchuk, Larysa; Sapundjiev, Danislav; ...

2013-02-19

Uranyl fluoride particles produced via hydrolysis of uranium hexafluoride have been deposited on different substrates: polished graphite disks, silver foil, stainless steel and gold-coated silicon wafer, and measured with micro-Raman spectroscopy (MRS). All three metallic substrates enhanced the Raman signal delivered by UO 2F 2 in comparison to graphite. The fundamental stretching of the U–O band appeared at 867 cm –1 in case of the graphite substrate, while in case of the others it was shifted to lower frequencies (down to 839 cm –1). All applied metallic substrates showed the expected effect of Raman signal enhancement; however the gold layermore » appeared to be most effective. Lastly, application of new substrates provides more information on the molecular structure of uranyl fluoride precipitation, which is interesting for nuclear safeguards and nuclear environmental analysis.« less

Method and apparatus for measuring enrichment of UF6

DOEpatents

Hill, Thomas Roy [Santa Fe, NM; Ianakiev, Kiril Dimitrov [Los Alamos, NM

2011-06-07

A system and method are disclosed for determining the enrichment of .sup.235U in Uranium Hexafluoride (UF6) utilizing synthesized X-rays which are directed at a container test zone containing a sample of UF6. A detector placed behind the container test zone then detects and counts the X-rays which pass through the container and the UF6. In order to determine the portion of the attenuation due to the UF6 gas alone, this count rate may then be compared to a calibration count rate of X-rays passing through a calibration test zone which contains a vacuum, the test zone having experienced substantially similar environmental conditions as the actual test zone. Alternatively, X-rays of two differing energy levels may be alternately directed at the container, where either the container or the UF6 has a high sensitivity to the difference in the energy levels, and the other having a low sensitivity.

Benchmark Gamma Spectroscopy Measurements of Uranium Hexafluoride in Aluminmum Pipe with a Sodium Iodide Detector

DOE Office of Scientific and Technical Information (OSTI.GOV)

March-Leuba, Jose A; Uckan, Taner; Gunning, John E

2010-01-01

The expected increased demand in fuel for nuclear power plants, combined with the fact that a significant portion of the current supply from the blend down of weapons-source material will soon be coming to an end, has led to the need for new sources of enriched uranium for nuclear fuel. As a result, a number of countries have announced plans, or are currently building, gaseous centrifuge enrichment plants (GCEPs) to supply this material. GCEPs have the potential to produce uranium at enrichments above the level necessary for nuclear fuel purposes-enrichments that make the uranium potentially usable for nuclear weapons. Asmore » a result, there is a critical need to monitor these facilities to ensure that nuclear material is not inappropriately enriched or diverted for unintended use. Significant advances have been made in instrument capability since the current International Atomic Energy Agency (IAEA) monitoring methods were developed. In numerous cases, advances have been made in other fields that have the potential, with modest development, to be applied in safeguards applications at enrichment facilities. A particular example of one of these advances is the flow and enrichment monitor (FEMO). (See Gunning, J. E. et al., 'FEMO: A Flow and Enrichment Monitor for Verifying Compliance with International Safeguards Requirements at a Gas Centrifuge Enrichment Facility,' Proceedings of the 8th International Conference on Facility Operations - Safeguards Interface. Portland, Oregon, March 30-April 4th, 2008.) The FEMO is a conceptual instrument capable of continuously measuring, unattended, the enrichment and mass flow of {sup 235}U in pipes at a GCEP, and consequently increase the probability that the potential production of HEU and/or diversion of fissile material will be detected. The FEMO requires no piping penetrations and can be installed on pipes containing the flow of uranium hexafluoride (UF{sub 6}) at a GCEP. This FEMO consists of separate parts, a flow monitor (FM) and an enrichment monitor (EM). Development of the FM is primarily the responsibility of Oak Ridge National Laboratory, and development of the EM is primarily the responsibility of Los Alamos National Laboratory. The FM will measure {sup 235}U mass flow rate by combining information from measuring the UF{sub 6} volumetric flow rate and the {sup 235}U density. The UF{sub 6} flow rate will be measured using characteristics of the process pumps used in product and tail UF{sub 6} header process lines of many GCEPs, and the {sup 235}U density will be measured using commercially available sodium iodide (NaI) gamma ray scintillation detectors. This report describes the calibration of the portion of the FM that measures the {sup 235}U density. Research has been performed to define a methodology and collect data necessary to perform this calibration without the need for plant declarations. The {sup 235}U density detector is a commercially available system (GammaRad made by Amptek, www.amptek.com) that contains the NaI crystal, photomultiplier tube, signal conditioning electronics, and a multichannel analyzer (MCA). Measurements were made with the detector system installed near four {sup 235}U sources. Two of the sources were made of solid uranium, and the other two were in the form of UF{sub 6} gas in aluminum piping. One of the UF{sub 6} gas sources was located at ORNL and the other at LANL. The ORNL source consisted of two pipe sections (schedule 40 aluminum pipe of 4-inch and 8-inch outside diameter) with 5.36% {sup 235}U enrichment, and the LANL source was a 4-inch schedule 40 aluminum pipe with 3.3% {sup 235}U enrichment. The configurations of the detector on these test sources, as well as on long straight pipe configurations expected to exist at GCEPs, were modeled using the computer code MCNP. The results of the MCNP calculations were used to define geometric correction factors between the test source and the GCEP application. Using these geometric correction factors, the experimental 186 keV counts in the test geometry were extrapolated to the expected GCEP geometry, and calibration curves were developed. A unique method to analyze the measurement was also developed that separated the detector spectrum into the five detectable decay gamma rays emitted by {sup 235}U in the 120 to 200 keV energy range. This analysis facilitated the assignment of a consistent value for the detector counts originating from {sup 235}U decays at 186 keV. This value is also more accurate because it includes the counts from gamma energies other than 186 keV, which results in increased counting statistics for the same measurement time. The 186 keV counts expected as a function of pressure and enrichment are presented in the body of this report. The main result of this research is a calibration factor for 4-inch and 8-inch schedule 40 aluminum pipes. For 4-inch pipes, the {sup 235}U density is 0.62 {sup 235}U g/m{sup 3} per each measured 186 keV count.« less

Potential uranium supply from phosphoric acid: A U.S. analysis comparing solvent extraction and Ion exchange recovery

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kim, Haeyeon; G. Eggert, Roderick; W. Carlsen, Brett

Phosphate rock contains significant amounts of uranium, although in low concentrations. Recovery of uranium as a by-product from phosphoric acid, an intermediate product produced during the recovery of phosphorus from phosphate rock, is not unprecedented. Phosphoric acid plants ceased to produce uranium as a by-product in the early 1990s with the fall of uranium prices. In the last decade, this topic has regained attention due to higher uranium prices and expected increase in demand for uranium. Our study revisits the topic and estimates how much uranium might be recoverable from current phosphoric acid production in the United States and whatmore » the associated costs might be considering two different recovery processes: solvent extraction and ion exchange. Based on U.S. phosphoric acid production in 2014, 5.5 million pounds of U 3O 8 could have been recovered, more than domestic U.S. mine production of uranium in the same year. Annualized costs for a hypothetical uranium recovery plant are US$48-66 per pound U 3O 8 for solvent extraction, the process used historically in the United States to recover uranium from phosphoric acid. For ion exchange, not yet proven at a commercial scale for uranium recovery, the estimated costs are US$33-54 per pound U 3O 8. Our results suggest that it is technically possible for the United States to recover significant quantities of uranium from current phosphoric acid production. And for this type of uranium production to be economically attractive on a large scale, either recovery costs must fall or uranium prices rise.« less

Potential uranium supply from phosphoric acid: A U.S. analysis comparing solvent extraction and Ion exchange recovery

DOE PAGES

Kim, Haeyeon; G. Eggert, Roderick; W. Carlsen, Brett; ...

2016-06-16

Phosphate rock contains significant amounts of uranium, although in low concentrations. Recovery of uranium as a by-product from phosphoric acid, an intermediate product produced during the recovery of phosphorus from phosphate rock, is not unprecedented. Phosphoric acid plants ceased to produce uranium as a by-product in the early 1990s with the fall of uranium prices. In the last decade, this topic has regained attention due to higher uranium prices and expected increase in demand for uranium. Our study revisits the topic and estimates how much uranium might be recoverable from current phosphoric acid production in the United States and whatmore » the associated costs might be considering two different recovery processes: solvent extraction and ion exchange. Based on U.S. phosphoric acid production in 2014, 5.5 million pounds of U 3O 8 could have been recovered, more than domestic U.S. mine production of uranium in the same year. Annualized costs for a hypothetical uranium recovery plant are US$48-66 per pound U 3O 8 for solvent extraction, the process used historically in the United States to recover uranium from phosphoric acid. For ion exchange, not yet proven at a commercial scale for uranium recovery, the estimated costs are US$33-54 per pound U 3O 8. Our results suggest that it is technically possible for the United States to recover significant quantities of uranium from current phosphoric acid production. And for this type of uranium production to be economically attractive on a large scale, either recovery costs must fall or uranium prices rise.« less

SEPARATION OF URANIUM, PLUTONIUM AND FISSION PRODUCTS FROM NEUTRON- BOMBARDED URANIUM

DOEpatents

Martin, A.E.; Johnson, I.; Burris, L. Jr.; Winsch, I.O.; Feder, H.M.

1962-11-13

A process is given for removing plutonium and/or fission products from uranium fuel. The fuel is dissolved in molten zinc--magnesium (10 to 18% Mg) alloy, more magnesium is added to obtain eutectic composition whereby uranium precipitates, and the uranium are separated from the Plutoniumand fission-product- containing eutectic. (AEC)

PRODUCTION OF PURIFIED URANIUM

DOEpatents

Burris, L. Jr.; Knighton, J.B.; Feder, H.M.

1960-01-26

A pyrometallurgical method for processing nuclear reactor fuel elements containing uranium and fission products and for reducing uranium compound; to metallic uranium is reported. If the material proccssed is essentially metallic uranium, it is dissolved in zinc, the sulution is cooled to crystallize UZn/sub 9/ , and the UZn/sub 9/ is distilled to obtain uranium free of fission products. If the material processed is a uranium compound, the sollvent is an alloy of zinc and magnesium and the remaining steps are the same.

Analysis of an open cycle gas core nuclear propulsion system using MHD driven vortices for fuel containment

NASA Astrophysics Data System (ADS)

Sedwick, Raymond John

1998-12-01

A novel method for containing gaseous uranium vapor in an open cycle nuclear space propulsion system is developed. In an attempt to increase the operating temperature of the nuclear reactor beyond the melting point of solid fuel rods (thus increasing specific impulse), the fuel is instead suspended as a vapor in the propellant using the pressure forces developed in a confined vortex flow. The introduction of the fuel as uranium hexafluoride is found to be effective in maintaining its vapor phase in the feed passages from the tank, but not in the main vortex. A mechanism by which the resulting condensation of the uranium may be tolerated is identified, and the electro- optical properties of the resulting mixture are investigated. Containment is modeled using a 1D- axisymmetric geometry, and radiative heat transfer is found to restrict the maximum specific impulse of the system to 1500 seconds using pumping pressures of 500 atm. The specific impulse is related to this pressure as pm1/4, allowing only marginal increases in Isp at increased pressure levels. Additional 2D- axisymmetric issues, such as non-uniform current distribution and bypass flows through the boundary layers, are investigated, with possible methods of solution cited. A two-group, two-region reactor analysis is performed, estimating the mass of the reactor to be about 10 metric tonnes, and establishing the thrust to weight ratio achievable by the system at about 50. To reduce the mass of the power system, a scheme for using cross-flow heat exchange with the propellant flow to minimize (and possibly eliminate) the need for radiators to reject waste heat is presented. (Copies available exclusively from MIT Libraries, Rm. 14-0551, Cambridge, MA 02139-4307. Ph. 617-253-5668; Fax 617-253-1690.)

Uranium production in Romania

DOE Office of Scientific and Technical Information (OSTI.GOV)

NONE

1994-12-01

This article reviews uranium production in Romania. Geological aspects of the country are discussed, and known uranium deposits are noted. Uranium mining and milling activities are also covered. Utilization of Romania`s uranium production industry will primarily be to supply the country`s nuclear power program, and with the present adequate supplies and the operation of their recently revamped fuel production facility, Romania should be self-reliant in the front end of the nuclear fuel cycle.

Plasma chemical conversion of sulphur hexafluoride initiated by a pulsed electron beam

NASA Astrophysics Data System (ADS)

Kholodnaya, Galina; Sazonov, Roman; Ponomarev, Denis; Guzeeva, Tatiana

2017-01-01

This paper presents the results of the experimental investigation of plasma chemical conversion of sulphur hexafluoride initiated by a pulsed electron beam (TEA-500 pulsed electron accelerator) with the following characteristics: 400-450 keV electron energy, 60 ns pulse duration, up to 200 J pulse energy, and 5 cm beam diameter. Experiments were conducted on the effect of the pulsed electron beam on SF6 and on mixtures of SF6 with O2, Ar, or N2. For the mixture of SF6 and oxygen, the results indicated chemical reactions involving the formation of a number of products of which one is sulphur, confirming the Wray - Fluorescence Analysis. The plasma chemical conversion of SF6 initiated by the pulsed electron beam was not detected when SF6 was mixed with Ar or N2, suggesting a possible mechanism for the reaction of SF6 in the presence of O2.

Method for the recovery of uranium values from uranium tetrafluoride

DOEpatents

Kreuzmann, Alvin B.

1983-01-01

The invention is a novel method for the recovery of uranium from dry, particulate uranium tetrafluoride. In one aspect, the invention comprises reacting particulate uranium tetrafluoride and calcium oxide in the presence of gaseous oxygen to effect formation of the corresponding alkaline earth metal uranate and alkaline earth metal fluoride. The product uranate is highly soluble in various acidic solutions wherein the product fluoride is virtually insoluble therein. The product mixture of uranate and alkaline earth metal fluoride is contacted with a suitable acid to provide a uranium-containing solution, from which the uranium is recovered. The invention can achieve quantitative recovery of uranium in highly pure form.

Method for the recovery of uranium values from uranium tetrafluoride

DOEpatents

Kreuzmann, A.B.

1982-10-27

The invention is a novel method for the recovery of uranium from dry, particulate uranium tetrafluoride. In one aspect, the invention comprises reacting particulate uranium tetrafluoride and calcium oxide in the presence of gaseous oxygen to effect formation of the corresponding alkaline earth metal uranate and alkaline earth metal fluoride. The product uranate is highly soluble in various acidic solutions whereas the product fluoride is virtually insoluble therein. The product mixture of uranate and alkaline earth metal fluoride is contacted with a suitable acid to provide a uranium-containing solution, from which the uranium is recovered. The invention can achieve quantitative recovery of uranium in highly pure form.

AIR INFILTRATION MEASUREMENTS USING TRACER GASES: A LITERATURE REVIEW

EPA Science Inventory

The report gives results of a literature review of air filtration measurements using tracer gases, including sulfur hexafluoride, hydrogen, carbon monoxide, carbon dioxide, nitrous oxide, and radioactive argon and krypton. Sulfur hexafluoride is the commonest tracer gas of choice...

Molten salt extraction of transuranic and reactive fission products from used uranium oxide fuel

DOEpatents

Herrmann, Steven Douglas

2014-05-27

Used uranium oxide fuel is detoxified by extracting transuranic and reactive fission products into molten salt. By contacting declad and crushed used uranium oxide fuel with a molten halide salt containing a minor fraction of the respective uranium trihalide, transuranic and reactive fission products partition from the fuel to the molten salt phase, while uranium oxide and non-reactive, or noble metal, fission products remain in an insoluble solid phase. The salt is then separated from the fuel via draining and distillation. By this method, the bulk of the decay heat, fission poisoning capacity, and radiotoxicity are removed from the used fuel. The remaining radioactivity from the noble metal fission products in the detoxified fuel is primarily limited to soft beta emitters. The extracted transuranic and reactive fission products are amenable to existing technologies for group uranium/transuranic product recovery and fission product immobilization in engineered waste forms.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lombardi, Marcie L.

2012-03-01

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 themore » 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 evaluated. Previously, UF{sub 6} gas enrichment monitors have required empty pipe measurements to accurately determine the pipe attenuation (the pipe attenuation is typically much larger than the attenuation in the gas). This dissertation reports on a method for determining the thickness of a pipe in a GCEP when obtaining an empty pipe measurement may not be feasible. This dissertation studies each of the components that may add to the final error in the enrichment measurement, and the factors that were taken into account to mitigate these issues are also detailed and tested. The use of an x-ray generator as a transmission source and the attending stability issues are addressed. Both analytical calculations and experimental measurements have been used. For completeness, some real-world analysis results from the URENCO Capenhurst enrichment plant have been included, where the final enrichment error has remained well below 1% for approximately two months.« less

PROCESS FOR SEPARATING URANIUM FISSION PRODUCTS

DOEpatents

Spedding, F.H.; Butler, T.A.; Johns, I.B.

1959-03-10

The removal of fission products such as strontium, barium, cesium, rubidium, or iodine from neutronirradiated uranium is described. Uranium halide or elemental halogen is added to melted irradiated uranium to convert the fission products to either more volatile compositions which vaporize from the melt or to higher melting point compositions which separate as solids.

76 FR 66777 - Office of Hazardous Materials Safety; Notice of Application for Special Permits

Federal Register 2010, 2011, 2012, 2013, 2014

2011-10-27

... hexafluoride. (modes 1, 2, 3, 4.) 15461-N Kidde Products 49 CFR 171.23... To authorize the High Bentham, Yo..., 3--Cargo vessel, 4--Cargo aircraft only, 5--Passenger-carrying aircraft. DATES: Comments must be... impracticable or not available. (mode 4.) 15469-N B.J. Alan Company 49 CFR 173.62... To authorize the Youngstown...

PROCESS FOR SEGREGATING URANIUM FROM PLUTONIUM AND FISSION-PRODUCT CONTAMINATION

DOEpatents

Ellison, C.V.; Runion, T.C.

1961-06-27

An aqueous nitric acid solution containing uranium, plutonium, and fission product values is contacted with an organic extractant comprised of a trialkyl phosphate and an organic diluent. The relative amounts of trialkyl phosphate and uranium values are controlled to achieve a concentration of uranium values in the organic extractant of at least 0.35 moles uranium per mole of trialkyl phosphate, thereby preferentially extracting uranium values into the organic extractant.

DISSOCIATION OF SULFUR HEXAFLUORIDE TRACER GAS IN THE PRESENCE OF AN INDOOR COMBUSTION SOURCE

EPA Science Inventory

As an odorless, non-toxic, and inert compound, sulfur hexafluoride (SF6) is one of the most widely used tracer gases in indoor air quality studies in both controlled and uncontrolled environments. This compound may be subject to hydrolysis under elevated temperature to form acidi...

THE CARBON DIOXIDE LEAKAGE FROM CHAMBERS MEASURED USING SULFUR HEXAFLUORIDE

EPA Science Inventory

In plant chamber studies, if Co2 leaking from a chamber is not quantified, it can lead to an overestimate of assimilation rates and an underestimate of respiration rates: consequently, it is critical that Co2 leakage be determined. Sulfur Hexafluoride (SF6) was introduced into t...

Process for continuous production of metallic uranium and uranium alloys

DOEpatents

Hayden, H.W. Jr.; Horton, J.A.; Elliott, G.R.B.

1995-06-06

A method is described for forming metallic uranium, or a uranium alloy, from uranium oxide in a manner which substantially eliminates the formation of uranium-containing wastes. A source of uranium dioxide is first provided, for example, by reducing uranium trioxide (UO{sub 3}), or any other substantially stable uranium oxide, to form the uranium dioxide (UO{sub 2}). This uranium dioxide is then chlorinated to form uranium tetrachloride (UCl{sub 4}), and the uranium tetrachloride is then reduced to metallic uranium by reacting the uranium chloride with a metal which will form the chloride of the metal. This last step may be carried out in the presence of another metal capable of forming one or more alloys with metallic uranium to thereby lower the melting point of the reduced uranium product. The metal chloride formed during the uranium tetrachloride reduction step may then be reduced in an electrolysis cell to recover and recycle the metal back to the uranium tetrachloride reduction operation and the chlorine gas back to the uranium dioxide chlorination operation. 4 figs.

Process for continuous production of metallic uranium and uranium alloys

DOEpatents

Hayden, Jr., Howard W.; Horton, James A.; Elliott, Guy R. B.

1995-01-01

A method is described for forming metallic uranium, or a uranium alloy, from uranium oxide in a manner which substantially eliminates the formation of uranium-containing wastes. A source of uranium dioxide is first provided, for example, by reducing uranium trioxide (UO.sub.3), or any other substantially stable uranium oxide, to form the uranium dioxide (UO.sub.2). This uranium dioxide is then chlorinated to form uranium tetrachloride (UCl.sub.4), and the uranium tetrachloride is then reduced to metallic uranium by reacting the uranium chloride with a metal which will form the chloride of the metal. This last step may be carried out in the presence of another metal capable of forming one or more alloys with metallic uranium to thereby lower the melting point of the reduced uranium product. The metal chloride formed during the uranium tetrachloride reduction step may then be reduced in an electrolysis cell to recover and recycle the metal back to the uranium tetrachloride reduction operation and the chlorine gas back to the uranium dioxide chlorination operation.

DECONTAMINATION OF URANIUM

DOEpatents

Feder, H.M.; Chellew, N.R.

1958-02-01

This patent deals with the separation of rare earth and other fission products from neutron bombarded uranium. This is accomplished by melting the uranium in contact with either thorium oxide, maguesium oxide, alumnum oxide, beryllium oxide, or uranium dioxide. The melting is preferably carried out at from 1150 deg to 1400 deg C in an inert atmosphere, such as argon or helium. During this treatment a scale of uranium dioxide forms on the uranium whtch contains most of the fission products.

Investigation of molybdate melts as an alternative method of reprocessing used nuclear fuel

DOE PAGES

Hames, Amber L.; Tkac, Peter; Paulenova, Alena; ...

2017-01-17

Here, an investigation of molybdate melts containing sodium molybdate (Na 2MoO 4) and molybdenum trioxide (MoO 3) to achieve the separation of uranium from fission products by crystallization has been performed. The separation is based on the difference in solubility of the fission product metal oxides compared to the uranium oxide or molybdate in the molybdate melt. The molybdate melt dissolves uranium dioxide at high temperatures, and upon cooling, uranium precipitates as uranium dioxide or molybdate, whereas the fission product metals remain soluble in the melt. Small-scale experiments using gram quantities of uranium dioxide have been performed to investigate themore » feasibility of UO 2 purification from the fission products. The composition of the uranium precipitate as well as data for partitioning of several fission product surrogates between the uranium precipitate and molybdate melt for various melt compositions are presented and discussed. The fission products Cs, Sr, Ru and Rh all displayed very large distribution ratios. The fission products Zr, Pd, and the lanthanides also displayed good distribution ratios (D > 10). A melt consisting of 20 wt% MoO 3-50 wt% Na 2MoO 4-30 wt% UO 2 heated to 1313 K and cooled to 1123 K for the physical separation of the UO 2 product from the melt, and washed once with Na 2MoO 4 displays optimum conditions for separation of the UO 2 from the fission products.« less

Helium and Sulfur Hexafluoride in Musical Instruments

ERIC Educational Resources Information Center

Forinash, Kyle; Dixon, Cory L.

2014-01-01

The effects of inhaled helium on the human voice were investigated in a recent article in "The Physics Teacher." As mentioned in that article, demonstrations of the effect are a popular classroom activity. If the number of YouTube videos is any indication, the effects of sulfur hexafluoride on the human voice are equally popular.…

PRODUCTION OF URANIUM

DOEpatents

Spedding, F.H.; Wilhelm, H.A.; Keller, W.H.

1958-04-15

The production of uranium metal by the reduction of uranium tetrafluoride is described. Massive uranium metal of high purily is produced by reacting uranium tetrafluoride with 2 to 20% stoichiometric excess of magnesium at a temperature sufficient to promote the reaction and then mantaining the reaction mass in a sealed vessel at temperature in the range of 1150 to 2000 d C, under a superatomospheric pressure of magnesium for a period of time sufficient 10 allow separation of liquid uranium and liquid magnesium fluoride into separate layers.

Uranium-mediated electrocatalytic dihydrogen production from water.

PubMed

Halter, Dominik P; Heinemann, Frank W; Bachmann, Julien; Meyer, Karsten

2016-02-18

Depleted uranium is a mildly radioactive waste product that is stockpiled worldwide. The chemical reactivity of uranium complexes is well documented, including the stoichiometric activation of small molecules of biological and industrial interest such as H2O, CO2, CO, or N2 (refs 1 - 11), but catalytic transformations with actinides remain underexplored in comparison to transition-metal catalysis. For reduction of water to H2, complexes of low-valent uranium show the highest potential, but are known to react violently and uncontrollably forming stable bridging oxo or uranyl species. As a result, only a few oxidations of uranium with water have been reported so far; all stoichiometric. Catalytic H2 production, however, requires the reductive recovery of the catalyst via a challenging cleavage of the uranium-bound oxygen-containing ligand. Here we report the electrocatalytic water reduction observed with a trisaryloxide U(III) complex [(((Ad,Me)ArO)3mes)U] (refs 18 and 19)--the first homogeneous uranium catalyst for H2 production from H2O. The catalytic cycle involves rare terminal U(IV)-OH and U(V)=O complexes, which have been isolated, characterized, and proven to be integral parts of the catalytic mechanism. The recognition of uranium compounds as potentially useful catalysts suggests new applications for such light actinides. The development of uranium-based catalysts provides new perspectives on nuclear waste management strategies, by suggesting that mildly radioactive depleted uranium--an abundant waste product of the nuclear power industry--could be a valuable resource.

DOE Office of Scientific and Technical Information (OSTI.GOV)

B.R. Westphal; J.C. Price; R.D. Mariani

The pyroprocessing of used nuclear fuel via electrorefining requires the continued addition of uranium trichloride to sustain operations. Uranium trichloride is utilized as an oxidant in the system to allow separation of uranium metal from the minor actinides and fission products. The inventory of uranium trichloride had diminished to a point that production was necessary to continue electrorefiner operations. Following initial experimentation, cupric chloride was chosen as a reactant with uranium metal to synthesize uranium trichloride. Despite the variability in equipment and charge characteristics, uranium trichloride was produced in sufficient quantities to maintain operations in the electrorefiner. The results andmore » conclusions from several experiments are presented along with a set of optimized operating conditions for the synthesis of uranium trichloride.« less

Zirconium determination by cooling curve analysis during the pyroprocessing of used nuclear fuel

NASA Astrophysics Data System (ADS)

Westphal, B. R.; Price, J. C.; Bateman, K. J.; Marsden, K. C.

2015-02-01

An alternative method to sampling and chemical analyses has been developed to monitor the concentration of zirconium in real-time during the casting of uranium products from the pyroprocessing of used nuclear fuel. The method utilizes the solidification characteristics of the uranium products to determine zirconium levels based on standard cooling curve analyses and established binary phase diagram data. Numerous uranium products have been analyzed for their zirconium content and compared against measured zirconium data. From this data, the following equation was derived for the zirconium content of uranium products:

The Use of Heavy Gas for Increased Reynolds Numbers in Transonic Wind Tunnels

NASA Technical Reports Server (NTRS)

Anders, J. B.; Anderson, W. K.; Murthy, A. V.

1998-01-01

The use of a high molecular weight test gas to increase the Reynolds number range of transonic wind tunnels is explored. Modifications to a small transonic wind tunnel are described and the real gas properties of the example heavy gas (sulfur hexafluoride) are discussed. Sulfur hexafluoride is shown to increase the test Reynolds number by a factor of more than 2 over air at the same Mach number. Experimental and computational pressure distributions on an advanced supercritical airfoil configuration at Mach 0.7 in both sulfur hexafluoride and nitrogen are presented. Transonic similarity theory is shown to be partially successful in transforming the heavy gas results to equivalent nitrogen (air) results, provided the correct definition of gamma is used.

High-voltage electrical apparatus utilizing an insulating gas of sulfur hexafluoride and helium

DOEpatents

Wootton, Roy E.

1980-01-01

High-voltage electrical apparatus includes an outer housing at low potential, an inner electrode disposed within the outer housing at high potential with respect thereto, and support means for insulatably supporting the inner electrode within the outer housing. Conducting particles contaminate the interior of the outer housing, and an insulating gas electrically insulates the inner electrode from the outer housing even in the presence of the conducting particles. The insulating gas is comprised of sulfur hexafluoride at a partial pressure of from about 2.9 to about 3.4 atmospheres absolute, and helium at a partial pressure from about 1.1 to about 11.4 atmospheres absolute. The sulfur hexafluoride comprises between 20 and 65 volume percent of the insulating gas.

26 CFR 1.993-3 - Definition of export property.

Code of Federal Regulations, 2012 CFR

2012-04-01

... deduction for depletion under section 613 or 613A. (3) Primary product from oil, gas, coal, or uranium. A primary product from oil, gas, coal, or uranium is not export property. For purposes of this paragraph— (i... primary products from oil, gas, coal, or uranium described in subdivisions (i) through (iv) of this...

26 CFR 1.993-3 - Definition of export property.

Code of Federal Regulations, 2014 CFR

2014-04-01

... deduction for depletion under section 613 or 613A. (3) Primary product from oil, gas, coal, or uranium. A primary product from oil, gas, coal, or uranium is not export property. For purposes of this paragraph— (i... primary products from oil, gas, coal, or uranium described in subdivisions (i) through (iv) of this...

26 CFR 1.993-3 - Definition of export property.

Code of Federal Regulations, 2013 CFR

2013-04-01

... deduction for depletion under section 613 or 613A. (3) Primary product from oil, gas, coal, or uranium. A primary product from oil, gas, coal, or uranium is not export property. For purposes of this paragraph— (i... primary products from oil, gas, coal, or uranium described in subdivisions (i) through (iv) of this...

Fuel preparation for use in the production of medical isotopes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Policke, Timothy A.; Aase, Scott B.; Stagg, William R.

The present invention relates generally to the field of medical isotope production by fission of uranium-235 and the fuel utilized therein (e.g., the production of suitable Low Enriched Uranium (LEU is uranium having 20 weight percent or less uranium-235) fuel for medical isotope production) and, in particular to a method for producing LEU fuel and a LEU fuel product that is suitable for use in the production of medical isotopes. In one embodiment, the LEU fuel of the present invention is designed to be utilized in an Aqueous Homogeneous Reactor (AHR) for the production of various medical isotopes including, butmore » not limited to, molybdenum-99, cesium-137, iodine-131, strontium-89, xenon-133 and yttrium-90.« less

Recovery of uranium from an irradiated solid target after removal of molybdenum-99 produced from the irradiated target

DOE Office of Scientific and Technical Information (OSTI.GOV)

Reilly, Sean Douglas; May, Iain; Copping, Roy

A process for minimizing waste and maximizing utilization of uranium involves recovering uranium from an irradiated solid target after separating the medical isotope product, molybdenum-99, produced from the irradiated target. The process includes irradiating a solid target comprising uranium to produce fission products comprising molybdenum-99, and thereafter dissolving the target and conditioning the solution to prepare an aqueous nitric acid solution containing irradiated uranium. The acidic solution is then contacted with a solid sorbent whereby molybdenum-99 remains adsorbed to the sorbent for subsequent recovery. The uranium passes through the sorbent. The concentrations of acid and uranium are then adjusted tomore » concentrations suitable for crystallization of uranyl nitrate hydrates. After inducing the crystallization, the uranyl nitrate hydrates are separated from a supernatant. The process results in the purification of uranyl nitrate hydrates from fission products and other contaminants. The uranium is therefore available for reuse, storage, or disposal.« less

Separation of uranium from technetium in recovery of spent nuclear fuel

NASA Astrophysics Data System (ADS)

Friedman, H. A.

1984-06-01

A method for decontaminating uranium product from the Purex 5 process is described. Hydrazine is added to the product uranyl nitrate stream from the Purex process, which contains hexavalent (UO2(2+)) uranium and heptavalent technetius (TcO4-). Technetium in the product stream is reduced and then complexed by the addition of oxalic acid (H2O2O4), and the Tc-oxalate complex is readily separated from the 10 uranium by solvent extraction with 30 vol % tributyl phosphate in n-dodecane.

Safeguards on uranium ore concentrate? the impact of modern mining and milling process

DOE Office of Scientific and Technical Information (OSTI.GOV)

Francis, Stephen

2013-07-01

Increased purity in uranium ore concentrate not only raises the question as to whether Safeguards should be applied to the entirety of uranium conversion facilities, but also as to whether some degree of coverage should be moved back to uranium ore concentrate production at uranium mining and milling facilities. This paper looks at uranium ore concentrate production across the globe and explores the extent to which increased purity is evident and the underlying reasons. Potential issues this increase in purity raises for IAEA's strategy on the Starting Point of Safeguards are also discussed.

Plutonium recovery from spent reactor fuel by uranium displacement

DOEpatents

Ackerman, John P.

1992-01-01

A process for separating uranium values and transuranic values from fission products containing rare earth values when the values are contained together in a molten chloride salt electrolyte. A molten chloride salt electrolyte with a first ratio of plutonium chloride to uranium chloride is contacted with both a solid cathode and an anode having values of uranium and fission products including plutonium. A voltage is applied across the anode and cathode electrolytically to transfer uranium and plutonium from the anode to the electrolyte while uranium values in the electrolyte electrolytically deposit as uranium metal on the solid cathode in an amount equal to the uranium and plutonium transferred from the anode causing the electrolyte to have a second ratio of plutonium chloride to uranium chloride. Then the solid cathode with the uranium metal deposited thereon is removed and molten cadmium having uranium dissolved therein is brought into contact with the electrolyte resulting in chemical transfer of plutonium values from the electrolyte to the molten cadmium and transfer of uranium values from the molten cadmium to the electrolyte until the first ratio of plutonium chloride to uranium chloride is reestablished.

Uranium ores and depleted uranium in the environment, with a reference to uranium in the biosphere from the Erzgebirge/Sachsen, Germany.

PubMed

Meinrath, A; Schneider, P; Meinrath, G

2003-01-01

The Erzgebirge ('Ore Mountains') area in the eastern part of Germany was a major source of uranium for Soviet nuclear programs between 1945 and 1989. During this time, the former German Democratic Republic became the third largest uranium producer in the world. The high abundance of uranium in the geological formations of the Erzgebirge are mirrored in the discovery of uranium by M. Klaproth close to Freiberg City in 1789 and the description of the so-called 'Schneeberg' disease, lung cancer caused in miners by the accumulation of the uranium decay product, radon, in the subsurfaces of shafts. Since 1991, remediation and mitigation of uranium at production facilities, rock piles and mill tailings has taken place. In parallel, efforts were initiated to assess the likely adverse effects of uranium mining to humans. The costs of these activities amount to about 6.5 10(9) Euro. A comparison with concentrations of depleted uranium at certain sites is given.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Schmoyer, RLS

The United States Department of Energy (DOE) currently manages the UF{sub 6} Cylinder Project. The project was formed to maintain and safely manage the depleted uranium hexafluoride (UF{sub 6}) stored in approximately 50,000 carbon steel cylinders. The cylinders are located at three DOE sites: the ETTP site (K-25) at Oak Ridge, Tennessee; the Paducah Gaseous Diffusion Plant (PGDP) in Paducah, Kentucky, and the Portsmouth Gaseous Diffusion Plant (PORTS) in Portsmouth, Ohio. The System Requirements Document (SRD) (LMES 1997a) delineates the requirements of the project. The appropriate actions needed to fulfill these requirements are then specified within the System Engineering Managementmore » Plan (SEMP) (LMES 1997b). This report documents activities that in whole or in part satisfy specific requirements and actions stated in the UF{sub 6} Cylinder Project SRD and SEMP with respect to forecasting cylinder conditions. The results presented here supercede those presented previously (Lyon 1995, 1996, 1997, 1998, 2000). Many of the wall thickness projections made in this report are conservative, because they are based on the assumption that corrosion trends will continue, despite activities such as improved monitoring, relocations to better storage, and painting.« less

Separation of uranium from technetium in recovery of spent nuclear fuel

DOEpatents

Friedman, H.A.

1984-06-13

A method for decontaminating uranium product from the Purex 5 process comprises addition of hydrazine to the product uranyl nitrate stream from the Purex process, which contains hexavalent (UO/sub 2//sup 2 +/) uranium and heptavalent technetium (TcO/sub 4/-). Technetium in the product stream is reduced and then complexed by the addition of oxalic acid (H/sub 2/C/sub 2/O/sub 4/), and the Tc-oxalate complex is readily separated from the 10 uranium by solvent extraction with 30 vol % tributyl phosphate in n-dodecane.

Separation of uranium from technetium in recovery of spent nuclear fuel

DOEpatents

Friedman, Horace A.

1985-01-01

A method for decontaminating uranium product from the Purex process comprises addition of hydrazine to the product uranyl nitrate stream from the Purex process, which contains hexavalent (UO.sub.2.sup.2+) uranium and heptavalent technetium (TcO.sub.4 -). Technetium in the product stream is reduced and then complexed by the addition of oxalic acid (H.sub.2 C.sub.2 O.sub.4), and the Tc-oxalate complex is readily separated from the uranium by solvent extraction with 30 vol. % tributyl phosphate in n-dodecane.

On the Melting Curve of Sulfur Hexafluoride

NASA Astrophysics Data System (ADS)

Harvey, Allan H.

2017-12-01

A previous correlation for the melting curve of sulfur hexafluoride (SF6) is inconsistent with the thermodynamic slope at the triple point derived from the Clapeyron equation. It is shown that this is probably due to the previous authors combining an accurate measurement of the triple point with melting-curve data that were distorted by impurities. A new equation is proposed that is consistent with the Clapeyron slope.

Use of a portable, automated, open-circuit gas quantification system and the sulfur hexafluoride tracer technique for measuring enteric methane emissions in Holstein cows fed ad libitum or restricted

USDA-ARS?s Scientific Manuscript database

The sulfur hexafluoride tracer technique (SF**6) is a commonly used method for measuring CH**4 enteric emissions in ruminants. Studies using SF**6 have shown large variation in CH**4 emissions data, inconsistencies in CH**4 emissions across studies, and potential methodological errors. Therefore, th...

Silica-alumina trihydrate filled epoxy castings resistant to arced SF.sub.6

DOEpatents

Chenoweth, Terrence E.; Yeoman, Frederick A.

1978-01-01

A cured, insulating, casting composition, having a coefficient of linear thermal expansion of below about 38 .times. 10.sup.-6 in./in./.degree. C and being resistant to arced sulfur hexafluoride gas, in contact with a metal surface in a sulfur hexafluoride gas environment, is made from hydantoin epoxy resin, anhydride curing agent and a filler combination of fused silica and alumina trihydrate.

Target and method for the production of fission product molybdenum-99

DOEpatents

Vandegrift, George F.; Vissers, Donald R.; Marshall, Simon L.; Varma, Ravi

1989-01-01

A target for the reduction of fission product Mo-99 is prepared from uranium of low U-235 enrichment by coating a structural support member with a preparatory coating of a substantially oxide-free substrate metal. Uranium metal is electrodeposited from a molten halide electrolytic bath onto a substrate metal. The electrodeposition is performed at a predetermined direct current rate or by using pulsed plating techniques which permit relaxation of accumulated uranium ion concentrations within the melt. Layers of as much as to 600 mg/cm.sup.2 of uranium can be prepared to provide a sufficient density to produce acceptable concentrations of fission product Mo-99.

Preliminary investigations on the use of uranium silicide targets for fission Mo-99 production

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cols, H.; Cristini, P.; Marques, R.

1997-08-01

The National Atomic Energy Commission (CNEA) of Argentine Republic owns and operates an installation for production of molybdenum-99 from fission products since 1985, and, since 1991, covers the whole national demand of this nuclide, carrying out a program of weekly productions, achieving an average activity of 13 terabecquerel per week. At present they are finishing an enlargement of the production plant that will allow an increase in the volume of production to about one hundred of terabecquerel. Irradiation targets are uranium/aluminium alloy with 90% enriched uranium with aluminium cladding. In view of international trends held at present for replacing highmore » enrichment uranium (HEU) for enrichment values lower than 20 % (LEU), since 1990 the authors are in contact with the RERTR program, beginning with tests to adapt their separation process to new irradiation target conditions. Uranium silicide (U{sub 3}Si{sub 2}) was chosen as the testing material, because it has an uranium mass per volume unit, so that it allows to reduce enrichment to a value of 20%. CNEA has the technology for manufacturing miniplates of uranium silicide for their purposes. In this way, equivalent amounts of Molybdenum-99 could be obtained with no substantial changes in target parameters and irradiation conditions established for the current process with Al/U alloy. This paper shows results achieved on the use of this new target.« less

PROCESSING OF URANIUM-METAL-CONTAINING FUEL ELEMENTS

DOEpatents

Moore, R.H.

1962-10-01

A process is given for recovering uranium from neutronbombarded uranium- aluminum alloys. The alloy is dissolved in an aluminum halide--alkali metal halide mixture in which the halide is a mixture of chloride and bromide, the aluminum halide is present in about stoichiometric quantity as to uranium and fission products and the alkali metal halide in a predominant quantity; the uranium- and electropositive fission-products-containing salt phase is separated from the electronegative-containing metal phase; more aluminum halide is added to the salt phase to obtain equimolarity as to the alkali metal halide; adding an excess of aluminum metal whereby uranium metal is formed and alloyed with the excess aluminum; and separating the uranium-aluminum alloy from the fission- productscontaining salt phase. (AEC)

SELECTIVE SEPARATION OF URANIUM FROM THORIUM, PROTACTINIUM AND FISSION PRODUCTS BY PEROXIDE DISSOLUTION METHOD

DOEpatents

Seaborg, G.T.; Gofman, J.W.; Stoughton, R.W.

1959-08-18

A method is described for separating U/sup 233/ from thorium and fission products. The separation is effected by forming a thorium-nitric acid solution of about 3 pH, adding hydrogen peroxide to precipitate uranium and thorium peroxide, treating the peroxides with sodium hydroxide to selectively precipitate the uranium peroxide, and reacting the separated solution with nitric acid to re- precipitate the uranium peroxide.

ELECTROCHEMICAL DECONTAMINATION AND RECOVERY OF URANIUM VALUES

DOEpatents

McLaren, J.A.; Goode, J.H.

1958-05-13

An electrochemical process is described for separating uranium from fission products. The method comprises subjecting the mass of uranium to anodic dissolution in an electrolytic cell containing aqueous alkali bicarbonate solution as its electrolyte, thereby promoting a settling from the solution of a solid sludge from about the electrodes and separating the resulting electrolyte solution containing the anodically dissolved uranium from the sludge which contains the rare earth fission products.

Uranium-mediated electrocatalytic dihydrogen production from water

NASA Astrophysics Data System (ADS)

Halter, Dominik P.; Heinemann, Frank W.; Bachmann, Julien; Meyer, Karsten

2016-02-01

Depleted uranium is a mildly radioactive waste product that is stockpiled worldwide. The chemical reactivity of uranium complexes is well documented, including the stoichiometric activation of small molecules of biological and industrial interest such as H2O, CO2, CO, or N2 (refs 1, 2, 3, 4, 5, 6, 7, 8, 9, 10, 11), but catalytic transformations with actinides remain underexplored in comparison to transition-metal catalysis. For reduction of water to H2, complexes of low-valent uranium show the highest potential, but are known to react violently and uncontrollably forming stable bridging oxo or uranyl species. As a result, only a few oxidations of uranium with water have been reported so far; all stoichiometric. Catalytic H2 production, however, requires the reductive recovery of the catalyst via a challenging cleavage of the uranium-bound oxygen-containing ligand. Here we report the electrocatalytic water reduction observed with a trisaryloxide U(III) complex [((Ad,MeArO)3mes)U] (refs 18 and 19)—the first homogeneous uranium catalyst for H2 production from H2O. The catalytic cycle involves rare terminal U(IV)-OH and U(V)=O complexes, which have been isolated, characterized, and proven to be integral parts of the catalytic mechanism. The recognition of uranium compounds as potentially useful catalysts suggests new applications for such light actinides. The development of uranium-based catalysts provides new perspectives on nuclear waste management strategies, by suggesting that mildly radioactive depleted uranium—an abundant waste product of the nuclear power industry—could be a valuable resource.

SEPARATION OF URANIUM, PLUTONIUM AND FISSION PRODUCTS

DOEpatents

Nicholls, C.M.; Wells, I.; Spence, R.

1959-10-13

The separation of uranium and plutonium from neutronirradiated uranium is described. The neutron-irradiated uranium is dissolved in nitric acid to provide an aqueous solution 3N in nitric acid. The fission products of the solution are extruded by treating the solution with dibutyl carbitol substantially 1.8N in nitric acid. The organic solvent phase is separated and neutralized with ammonium hydroxide and the plutonium reduced with hydroxylamine base to the trivalent state. Treatment of the mixture with saturated ammonium nitrate extracts the reduced plutonium and leaves the uranium in the organic solvent.

Plutonium recovery from spent reactor fuel by uranium displacement

DOEpatents

Ackerman, J.P.

1992-03-17

A process is described for separating uranium values and transuranic values from fission products containing rare earth values when the values are contained together in a molten chloride salt electrolyte. A molten chloride salt electrolyte with a first ratio of plutonium chloride to uranium chloride is contacted with both a solid cathode and an anode having values of uranium and fission products including plutonium. A voltage is applied across the anode and cathode electrolytically to transfer uranium and plutonium from the anode to the electrolyte while uranium values in the electrolyte electrolytically deposit as uranium metal on the solid cathode in an amount equal to the uranium and plutonium transferred from the anode causing the electrolyte to have a second ratio of plutonium chloride to uranium chloride. Then the solid cathode with the uranium metal deposited thereon is removed and molten cadmium having uranium dissolved therein is brought into contact with the electrolyte resulting in chemical transfer of plutonium values from the electrolyte to the molten cadmium and transfer of uranium values from the molten cadmium to the electrolyte until the first ratio of plutonium chloride to uranium chloride is reestablished.

Spectroscopic studies of uranium species for environmental decontamination applications

NASA Astrophysics Data System (ADS)

Eng, Charlotte

After the Cold War, Department of Energy began to concentrate its efforts on cleanup of former nuclear material processing facilities, especially uranium-contaminated groundwater and soil. This research aims to study uranium association to both organic and inorganic compounds found in the contaminated environment in the hopes that the information gathered can be applied to the development and optimization of cost-effective remediation techniques. Spectroscopic and electrochemical methods will be employed to examine the behavior of uranium in given conditions to further our understanding of its impact on the environment. Uranium found in groundwater and soil bind with various ligands, especially organic ligands present in the environment due to natural sources (e.g. metabolic by-products or degradation of plants and animals) or man-made sources (e.g. chelating agents used in operating or cleanup of uranium processing facilities). We selected reasonable analogs of naturally occurring matter and studied their structure, chemical and electrochemical behavior and found that the structure of uranyl complexes depends heavily on the nature of the ligand and environmental factors such as pH. Association of uranium-organic complexes with anaerobic bacteria, Clostridium sp. was studied to establish if the bacteria can effectively bioreduce uranium while going through normal bacterial activity. It was found that the nature of the organic ligand affected the bioavailability and toxicity of the uranium on the bacteria. In addition, we have found that the type of iron corrosion products and uranyl species present on the surface of corroded steel depended on various environmental factors, which subsequently affected the removal rate of uranium by a citric acid/hydrogen peroxide/deionized water cleaning process. The method was found to remove uranium from only the topmost corrosion layers and residual uranium could be found (a) deeper in the corrosion layers where it is occluded by the steel corrosion products or (b) in areas where the dissolved uranium/iron species, the products generated by the dissolution power of citric acid, was not properly rinsed away.

Separation of uranium from technetium in recovery of spent nuclear fuel

DOEpatents

Pruett, D.J.; McTaggart, D.R.

1983-08-31

Uranium and technetium in the product stream of the Purex process for recovery of uranium in spent nuclear fuel are separated by (1) contacting the aqueous Purex product stream with hydrazine to reduce Tc/sup +7/ therein to a reduced species, and (2) contacting said aqueous stream with an organic phase containing tributyl phosphate and an organic diluent to extract uranium from said aqueous stream into said organic phase.

Separation of uranium from technetium in recovery of spent nuclear fuel

DOEpatents

Pruett, David J.; McTaggart, Donald R.

1984-01-01

Uranium and technetium in the product stream of the Purex process for recovery of uranium in spent nuclear fuel are separated by (1) contacting the aqueous Purex product stream with hydrazine to reduce Tc.sup.+7 therein to a reduced species, and (2) contacting said aqueous stream with an organic phase containing tributyl phosphate and an organic diluent to extract uranium from said aqueous stream into said organic phase.

Patterns and Features of Global Uranium Resources and Production

NASA Astrophysics Data System (ADS)

Wang, Feifei; Song, Zisheng; Cheng, Xianghu; Huanhuan, MA

2017-11-01

With the entry into force of the Paris Agreement, the development of clean and low-carbon energy has become the consensus of the world. Nuclear power is one energy that can be vigorously developed today and in the future. Its sustainable development depends on a sufficient supply of uranium resources. It is of great practical significance to understand the distribution pattern of uranium resources and production. Based on the latest international authoritative reports and data, this paper analysed the distribution of uranium resources, the distribution of resources and production in the world, and the developing tendency in future years. The results show that the distribution of uranium resources is uneven in the world, and the discrepancies between different type deposits is very large. Among them, sandstone-type uranium deposits will become the main type owing to their advantages of wide distribution, minor environmental damage, mature mining technology and high economic benefit.

Target and method for the production of fission product molybdenum-99

DOEpatents

Vandegrift, G.F.; Vissers, D.R.; Marshall, S.L.; Varma, R.

1987-10-26

A target for the reduction of fission product Mo-99 is prepared from uranium of low U-235 enrichment by coating a structural support member with a preparatory coating of a substantially oxide-free substrate metal. Uranium metal is electrodeposited from a molten halide electrolytic bath onto a substrate metal. The electrodeposition is performed at a predetermined direct current rate or by using pulsed plating techniques which permit relaxation of accumulated uranium ion concentrations within the melt. Layers of as much as to 600 mg/cm/sup 2/ of uranium can be prepared to provide a sufficient density to produce acceptable concentrations of fission product Mo-99. 2 figs.

Neutron-rich isotope production using a uranium carbide - carbon nanotubes SPES target prototype

NASA Astrophysics Data System (ADS)

Corradetti, S.; Biasetto, L.; Manzolaro, M.; Scarpa, D.; Carturan, S.; Andrighetto, A.; Prete, G.; Vasquez, J.; Zanonato, P.; Colombo, P.; Jost, C. U.; Stracener, D. W.

2013-05-01

The SPES (Selective Production of Exotic Species) project, under development at the Istituto Nazionale di Fisica Nucleare - Laboratori Nazionali di Legnaro (INFN-LNL), is a new-generation Isotope Separation On-Line (ISOL) facility for the production of radioactive ion beams by means of the proton-induced fission of uranium. In the framework of the research on the SPES target, seven uranium carbide discs, obtained by reacting uranium oxide with graphite and carbon nanotubes, were irradiated with protons at the Holifield Radioactive Ion Beam Facility (HRIBF) of Oak Ridge National Laboratory (ORNL). In the following, the yields of several fission products obtained during the experiment are presented and discussed. The experimental results are then compared to those obtained using a standard uranium carbide target. The reported data highlights the capability of the new type of SPES target to produce and release isotopes of interest for the nuclear physics community.

Uranium oxidation: Characterization of oxides formed by reaction with water by infrared and sorption analyses

NASA Astrophysics Data System (ADS)

Fuller, E. L.; Smyrl, N. R.; Condon, J. B.; Eager, M. H.

1984-04-01

Three different uranium oxide samples have been characterized with respect to the different preparation techniques. The results show that the water reaction with uranium metal occurs cyclically forming laminar layers of oxide which spall off due to the strain at the oxide/metal interface. Single laminae are released if liquid water is present due to the prizing penetration at the reaction zone. The rate of reaction of water with uranium is directly proportional to the amount of adsorbed water on the oxide product. Rapid transport is effected through the open hydrous oxide product. Dehydration of the hydrous oxide irreversibly forms a more inert oxide which cannot be rehydrated to the degree that prevails in the original hydrous product of uranium oxidation with water. Inert gas sorption analyses and diffuse reflectance infrared studies combined with electron microscopy prove valuable in defining the chemistry and morphology of the oxidic products and hydrated intermediates.

Chemical state of fission products in irradiated uranium carbide fuel

NASA Astrophysics Data System (ADS)

Arai, Yasuo; Iwai, Takashi; Ohmichi, Toshihiko

1987-12-01

The chemical state of fission products in irradiated uranium carbide fuel has been estimated by equilibrium calculation using the SOLGASMIX-PV program. Solid state fission products are distributed to the fuel matrix, ternary compounds, carbides of fission products and intermetallic compounds among the condensed phases appearing in the irradiated uranium carbide fuel. The chemical forms are influenced by burnup as well as stoichiometry of the fuel. The results of the present study almost agree with the experimental ones reported for burnup simulated carbides.

Uranium deposits of Brazil

DOE Office of Scientific and Technical Information (OSTI.GOV)

NONE

Brazil is a country of vast natural resources, including numerous uranium deposits. In support of the country`s nuclear power program, Brazil has developed the most active uranium industry in South America. Brazil has one operating reactor (Angra 1, a 626-MWe PWR), and two under construction. The country`s economic challenges have slowed the progress of its nuclear program. At present, the Pocos de Caldas district is the only active uranium production. In 1990, the Cercado open-pit mine produced approximately 45 metric tons (MT) U{sub 3}O{sub 8} (100 thousand pounds). Brazil`s state-owned uranium production and processing company, Uranio do Brasil, announced itmore » has decided to begin shifting its production from the high-cost and nearly depleted deposits at Pocos de Caldas, to lower-cost reserves at Lagoa Real. Production at Lagoa Real is schedules to begin by 1993. In addition to these two districts, Brazil has many other known uranium deposits, and as a whole, it is estimated that Brazil has over 275,000 MT U{sub 3}O{sub 8} (600 million pounds U{sub 3}O{sub 8}) in reserves.« less

PRODUCTION OF URANIUM METAL BY CARBON REDUCTION

DOEpatents

Holden, R.B.; Powers, R.M.; Blaber, O.J.

1959-09-22

The preparation of uranium metal by the carbon reduction of an oxide of uranium is described. In a preferred embodiment of the invention a charge composed of carbon and uranium oxide is heated to a solid mass after which it is further heated under vacuum to a temperature of about 2000 deg C to produce a fused uranium metal. Slowly ccoling the fused mass produces a dendritic structure of uranium carbide in uranium metal. Reacting the solidified charge with deionized water hydrolyzes the uranium carbide to finely divide uranium dioxide which can be separated from the coarser uranium metal by ordinary filtration methods.

Chemistry of the 5g Elements: Relativistic Calculations on Hexafluorides.

PubMed

Dognon, Jean-Pierre; Pyykkö, Pekka

2017-08-14

A Periodic System was proposed for the elements 1-172 by Pyykkö on the basis of atomic and ionic calculations. In it, the elements 121-138 were nominally assigned to a 5g row. We now perform molecular, relativistic four-component DFT calculations and find that the hexafluorides of the elements 125-129 indeed enjoy occupied 5g states. © 2017 Wiley-VCH Verlag GmbH & Co. KGaA, Weinheim.

Research on technology of online gas chromatograph for SF6 decomposition products

NASA Astrophysics Data System (ADS)

Li, L.; Fan, X. P.; Zhou, Y. Y.; Tang, N.; Zou, Z. L.; Liu, M. Z.; Huang, G. J.

2017-12-01

Sulfur hexafluoride (SF6) decomposition products were qualitatively and quantitatively analyzed by several gas chromatographs in the laboratory. Test conditions and methods were selected and optimized to minimize and eliminate the SF6’ influences on detection of other trace components. The effective separation and detection of selected characteristic gases were achieved. And by comparison among different types of gas chromatograph, it was found that GPTR-S101 can effectively separate and detect SF6 decomposition products and has best the best detection limit and sensitivity. On the basis of GPTR-S101, online gas chromatograph for SF6decomposition products (GPTR-S201) was developed. It lays the foundation for further online monitoring and diagnosis of SF6.

Gas Sensing Analysis of Ag-Decorated Graphene for Sulfur Hexafluoride Decomposition Products Based on the Density Functional Theory

PubMed Central

Zhang, Xiaoxing; Huang, Rong; Gui, Yingang; Zeng, Hong

2016-01-01

Detection of decomposition products of sulfur hexafluoride (SF6) is one of the best ways to diagnose early latent insulation faults in gas-insulated equipment, and the occurrence of sudden accidents can be avoided effectively by finding early latent faults. Recently, functionalized graphene, a kind of gas sensing material, has been reported to show good application prospects in the gas sensor field. Therefore, calculations were performed to analyze the gas sensing properties of intrinsic graphene (Int-graphene) and functionalized graphene-based material, Ag-decorated graphene (Ag-graphene), for decomposition products of SF6, including SO2F2, SOF2, and SO2, based on density functional theory (DFT). We thoroughly investigated a series of parameters presenting gas-sensing properties of adsorbing process about gas molecule (SO2F2, SOF2, SO2) and double gas molecules (2SO2F2, 2SOF2, 2SO2) on Ag-graphene, including adsorption energy, net charge transfer, electronic state density, and the highest and lowest unoccupied molecular orbital. The results showed that the Ag atom significantly enhances the electrochemical reactivity of graphene, reflected in the change of conductivity during the adsorption process. SO2F2 and SO2 gas molecules on Ag-graphene presented chemisorption, and the adsorption strength was SO2F2 > SO2, while SOF2 absorption on Ag-graphene was physical adsorption. Thus, we concluded that Ag-graphene showed good selectivity and high sensitivity to SO2F2. The results can provide a helpful guide in exploring Ag-graphene material in experiments for monitoring the insulation status of SF6-insulated equipment based on detecting decomposition products of SF6. PMID:27809269

A top-down assessment of energy, water and land use in uranium mining, milling, and refining

DOE Office of Scientific and Technical Information (OSTI.GOV)

E. Schneider; B. Carlsen; E. Tavrides

2013-11-01

Land, water and energy use are key measures of the sustainability of uranium production into the future. As the most attractive, accessible deposits are mined out, future discoveries may prove to be significantly, perhaps unsustainably, more intensive consumers of environmental resources. A number of previous attempts have been made to provide empirical relationships connecting these environmental impact metrics to process variables such as stripping ratio and ore grade. These earlier attempts were often constrained by a lack of real world data and perform poorly when compared against data from modern operations. This paper conditions new empirical models of energy, watermore » and land use in uranium mining, milling, and refining on contemporary data reported by operating mines. It shows that, at present, direct energy use from uranium production represents less than 1% of the electrical energy produced by the once-through fuel cycle. Projections of future energy intensity from uranium production are also possible by coupling the empirical models with estimates of uranium crustal abundance, characteristics of new discoveries, and demand. The projections show that even for the most pessimistic of scenarios considered, by 2100, the direct energy use from uranium production represents less than 3% of the electrical energy produced by the contemporary once-through fuel cycle.« less

Method for converting uranium oxides to uranium metal

DOEpatents

Duerksen, Walter K.

1988-01-01

A process is described for converting scrap and waste uranium oxide to uranium metal. The uranium oxide is sequentially reduced with a suitable reducing agent to a mixture of uranium metal and oxide products. The uranium metal is then converted to uranium hydride and the uranium hydride-containing mixture is then cooled to a temperature less than -100.degree. C. in an inert liquid which renders the uranium hydride ferromagnetic. The uranium hydride is then magnetically separated from the cooled mixture. The separated uranium hydride is readily converted to uranium metal by heating in an inert atmosphere. This process is environmentally acceptable and eliminates the use of hydrogen fluoride as well as the explosive conditions encountered in the previously employed bomb-reduction processes utilized for converting uranium oxides to uranium metal.

12. VIEW OF DEPLETED URANIUM INGOT AND MOLDS. DEPLETED URANIUM ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

12. VIEW OF DEPLETED URANIUM INGOT AND MOLDS. DEPLETED URANIUM CASTING OPERATIONS CEASED IN 1988. (11/14/57) - Rocky Flats Plant, Non-Nuclear Production Facility, South of Cottonwood Avenue, west of Seventh Avenue & east of Building 460, Golden, Jefferson County, CO

URANIUM RECOVERY PROCESS

DOEpatents

Stevenson, J.W.; Werkema, R.G.

1959-07-28

The recovery of uranium from magnesium fluoride slag obtained as a by- product in the production of uranium metal by the bomb reduction prccess is presented. Generally the recovery is accomplished by finely grinding the slag, roasting ihe ground slag air, and leaching the roasted slag with a hot, aqueous solution containing an excess of the sodium bicarbonate stoichiometrically required to form soluble uranium carbonate complex. The roasting is preferably carried out at between 425 and 485 deg C for about three hours. The leaching is preferably done at 70 to 90 deg C and under pressure. After leaching and filtration the uranium may be recovered from the clear leach liquor by any desired method.

PRODUCTION OF URANIUM MONOCARBIDE

DOEpatents

Powers, R.M.

1962-07-24

A method of making essentially stoichiometric uranium monocarbide by pelletizing a mixture of uranium tetrafluoride, silicon, and carbon and reacting the mixture at a temperature of approximately 1500 to 1700 deg C until the reaction goes to completion, forming uranium monocarbide powder and volatile silicon tetrafluoride, is described. The powder is then melted to produce uranium monocarbide in massive form. (AEC)

Recovering and recycling uranium used for production of molybdenum-99

DOE Office of Scientific and Technical Information (OSTI.GOV)

Reilly, Sean Douglas; May, Iain; Copping, Roy

A processes for recycling uranium that has been used for the production of molybdenum-99 involves irradiating a solution of uranium suitable for forming fission products including molybdenum-99, conditioning the irradiated solution to one suitable for inducing the formation of crystals of uranyl nitrate hydrates, then forming the crystals and a supernatant and then separating the crystals from the supernatant, thus using the crystals as a source of uranium for recycle. Molybdenum-99 is recovered from the supernatant using an adsorbent such as alumina. Another process involves irradiation of a solid target comprising uranium, forming an acidic solution from the irradiated targetmore » suitable for inducing the formation of crystals of uranyl nitrate hydrates, then forming the crystals and a supernatant and then separating the crystals from the supernatant, thus using the crystals as a source of uranium for recycle. Molybdenum-99 is recovered from the supernatant using an adsorbent such as alumina.« less

METHOD OF SEPARATING URANIUM VALUES, PLUTONIUM VALUES AND FISSION PRODUCTS BY CHLORINATION

DOEpatents

Brown, H.S.; Seaborg, G.T.

1959-02-24

The separation of plutonium and uranium from each other and from other substances is described. In general, the method comprises the steps of contacting the uranium with chlorine in the presence of a holdback material selected from the group consisting of lanthanum oxide and thorium oxide to form a uranium chloride higher than uranium tetrachloride, and thereafter heating the uranium chloride thus formed to a temperature at which the uranium chloride is volatilized off but below the volatilizalion temperature of plutonium chloride.

10 CFR 765.3 - Definitions.

Code of Federal Regulations, 2013 CFR

2013-01-01

... ENERGY REIMBURSEMENT FOR COSTS OF REMEDIAL ACTION AT ACTIVE URANIUM AND THORIUM PROCESSING SITES General... uranium or thorium processing site or active processing site means: (1) Any uranium or thorium processing... an Agreement State, for the production at a site of any uranium or thorium derived from ore— (i) Was...

10 CFR 765.3 - Definitions.

Code of Federal Regulations, 2012 CFR

2012-01-01

... ENERGY REIMBURSEMENT FOR COSTS OF REMEDIAL ACTION AT ACTIVE URANIUM AND THORIUM PROCESSING SITES General... uranium or thorium processing site or active processing site means: (1) Any uranium or thorium processing... an Agreement State, for the production at a site of any uranium or thorium derived from ore— (i) Was...

10 CFR 765.3 - Definitions.

Code of Federal Regulations, 2014 CFR

2014-01-01

... ENERGY REIMBURSEMENT FOR COSTS OF REMEDIAL ACTION AT ACTIVE URANIUM AND THORIUM PROCESSING SITES General... uranium or thorium processing site or active processing site means: (1) Any uranium or thorium processing... an Agreement State, for the production at a site of any uranium or thorium derived from ore— (i) Was...

DOE Office of Scientific and Technical Information (OSTI.GOV)

Curtis, Michael M.

As a result of NSG restrictions, India cannot import the natural uranium required to fuel its Pressurized Heavy Water Reactors (PHWRs); consequently, it is forced to rely on the expediency of domestic uranium production. However, domestic production from mines and byproduct sources has not kept pace with demand from commercial reactors. This shortage has been officially confirmed by the Indian Planning Commission’s Mid-Term Appraisal of the country’s current Five Year Plan. The report stresses that as a result of the uranium shortage, Indian PHWR load factors have been continually decreasing. The Uranium Corporation of India Ltd (UCIL) operates a numbermore » of underground mines in the Singhbhum Shear Zone of Jharkhand, and it is all processed at a single mill in Jaduguda. UCIL is attempting to aggrandize operations by establishing new mines and mills in other states, but the requisite permit-gathering and development time will defer production until at least 2009. A significant portion of India’s uranium comes from byproduct sources, but a number of these are derived from accumulated stores that are nearing exhaustion. A current maximum estimate of indigenous uranium production is 430t/yr (230t from mines and 200t from byproduct sources); whereas, the current uranium requirement for Indian PHWRs is 455t/yr (depending on plant capacity factor). This deficit is exacerbated by the additional requirements of the Indian weapons program. Present power generation capacity of Indian nuclear plants is 4350 MWe. The power generation target set by the Indian Department of Atomic Energy (DAE) is 20,000 MWe by the year 2020. It is expected that around half of this total will be provided by PHWRs using indigenously supplied uranium with the bulk of the remainder provided by breeder reactors or pressurized water reactors using imported low-enriched uranium.« less

Analysis of an Indirect Neutron Signature for Enhanced UF6 Cylinder Verification

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kulisek, Jonathan A.; McDonald, Benjamin S.; Smith, Leon E.

2017-02-21

The International Atomic Energy Agency (IAEA) currently uses handheld gamma-ray spectrometers combined with ultrasonic wall-thickness gauges to verify the declared enrichment of uranium hexafluoride (UF6) cylinders. The current method provides relatively low accuracy for the assay of 235U enrichment, especially for natural and depleted UF6. Furthermore, the current method provides no capability to assay the absolute mass of 235U in the cylinder due to the localized instrument geometry and limited penetration of the 186-keV gamma-ray signature from 235U. Also, the current verification process is a time-consuming component of on-site inspections at uranium enrichment plants. Toward the goal of a more-capablemore » cylinder assay method, the Pacific Northwest National Laboratory has developed the hybrid enrichment verification array (HEVA). HEVA measures both the traditional 186-keV direct signature and a non-traditional, high-energy neutron-induced signature (HEVANT). HEVANT enables full-volume assay of UF6 cylinders by exploiting the relatively larger mean free paths of the neutrons emitted from the UF6. In this work, Monte Carlo modeling is used as the basis for characterizing HEVANT in terms of the individual contributions to HEVANT from nuclides and hardware components. Monte Carlo modeling is also used to quantify the intrinsic efficiency of HEVA for neutron detection in a cylinder-assay geometry. Modeling predictions are validated against neutron-induced gamma-ray spectra from laboratory measurements and a relatively large population of Type 30B cylinders spanning a range of enrichments. Implications of the analysis and findings on the viability of HEVA for cylinder verification are discussed, such as the resistance of the HEVANT signature to manipulation by the nearby placement of neutron-conversion materials.« less

Organic geochemical analysis of sedimentary organic matter associated with uranium

USGS Publications Warehouse

Leventhal, J.S.; Daws, T.A.; Frye, J.S.

1986-01-01

Samples of sedimentary organic matter from several geologic environments and ages which are enriched in uranium (56 ppm to 12%) have been characterized. The three analytical techniqyes used to study the samples were Rock-Eval pyrolysis, pyrolysis-gas chromatography-mass spectrometry, and solid-state C-13 nuclear magnetic resonance (NMR) spectroscopy. In samples with low uranium content, the pyrolysis-gas chromatography products contain oxygenated functional groups (as hydroxyl) and molecules with both aliphatic and aromatic carbon atoms. These samples with low uranium content give measurable Rock-Eval hydrocarbon and organic-CO2 yields, and C-13 NMR values of > 30% aliphatic carbon. In contrast, uranium-rich samples have few hydrocarbon pyrolysis products, increased Rock-Eval organic-CO2 contents and > 70% aromatic carbon contents from C-13 NMR. The increase in aromaticity and decrease in hydrocarbon pyrolysis yield are related to the amount of uranium and the age of the uranium minerals, which correspond to the degree of radiation damage. The three analytical techniques give complementary results. Increase in Rock-Eval organic-CO2 yield correlates with uranium content for samples from the Grants uranium region. Calculations show that the amount of organic-CO2 corresponds to the quantity of uranium chemically reduced by the organic matter for the Grants uranium region samples. ?? 1986.

SEPARATION OF PLUTONIUM FROM URANIUM AND FISSION PRODUCTS

DOEpatents

Boyd, G.E.; Adamson, A.W.; Schubert, J.; Russell, E.R.

1958-10-01

A chromatographic adsorption process is presented for the separation of plutonium from other fission products formed by the irradiation of uranium. The plutonium and the lighter element fission products are adsorbed on a sulfonated phenol-formaldehyde resin bed from a nitric acid solution containing the dissolved uranium. Successive washes of sulfuric, phosphoric, and nitric acids remove the bulk of the fission products, then an eluate of dilute phosphoric and nitric acids removes the remaining plutonium and fission products. The plutonium is selectively removed by passing this solution through zirconium phosphate, from which the plutonium is dissolved with nitric acid. This process provides a convenient and efficient means for isolating plutonium.

Determination of irradiated reactor uranium in soil samples in Belarus using 236U as irradiated uranium tracer.

PubMed

Mironov, Vladislav P; Matusevich, Janna L; Kudrjashov, Vladimir P; Boulyga, Sergei F; Becker, J Sabine

2002-12-01

This work presents experimental results on the distribution of irradiated reactor uranium from fallout after the accident at Chernobyl Nuclear Power Plant (NPP) in comparison to natural uranium distribution in different soil types. Oxidation processes and vertical migration of irradiated uranium in soils typical of the 30 km relocation area around Chernobyl NPP were studied using 236U as the tracer for irradiated reactor uranium and inductively coupled plasma mass spectrometry as the analytical method for uranium isotope ratio measurements. Measurements of natural uranium yielded significant variations of its concentration in upper soil layers from 2 x 10(-7) g g(-1) to 3.4 x 10(-6) g g(-1). Concentrations of irradiated uranium in the upper 0-10 cm soil layers at the investigated sampling sites varied from 5 x 10(-12) g g(-1) to 2 x 10(-6) g g(-1) depending on the distance from Chernobyl NPP. In the majority of investigated soil profiles 78% to 97% of irradiated "Chernobyl" uranium is still contained in the upper 0-10 cm soil layers. The physical and chemical characteristics of the soil do not have any significant influence on processes of fuel particle destruction. Results obtained using carbonate leaching of 236U confirmed that more than 60% of irradiated "Chernobyl" uranium is still in a tetravalent form, ie. it is included in the fuel matrix (non-oxidized fuel UO2). The average value of the destruction rate of fuel particles determined for the Western radioactive trace (k = 0.030 +/- 0.005 yr(-1)) and for the Northern radioactive trace (k = 0.035 + 0.009 yr(-1)) coincide within experimental errors. Use of leaching of fission products in comparison to leaching of uranium for study of the destruction rate of fuel particles yielded poor coincidence due to the fact that use of fission products does not take into account differences in the chemical properties of fission products and fuel matrix (uranium).

Thermogravimetric Control of Intermediate Products in the Metallurgy of Uranium; CONTROL TERMOGRAVIMETRICO DE PRODUCTOS INTERMEDIOS DE LA METALURGIA DEL URANIO

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sanchez, L.G.; Cellini, R.F.

1959-01-01

The thermal decomposition of some intermediate compounds in the metallurgy of uranium such as uranium peroxide, ammonium uranate, ammonium uranium pentafluoride, uranium tetrafluoride, and UO/sub 2/, were studied using Chevenard's thermobalance. Some data on the pyrolysis of synthetic mixtures of intermediate compounds which may appear during the industrial processing are given. Thermogravimetric methods of control are suggested for use in uranium metallurgy. (tr-auth)

Scoping study to expedite development of a field deployable and portable instrument for UF6 enrichment assay

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chan, George; Valentine, John D.; Russo, Richard E.

The primary objective of the present study is to identity the most promising, viable technologies that are likely to culminate in an expedited development of the next-generation, field-deployable instrument for providing rapid, accurate, and precise enrichment assay of uranium hexafluoride (UF6). UF6 is typically involved, and is arguably the most important uranium compound, in uranium enrichment processes. As the first line of defense against proliferation, accurate analytical techniques to determine the uranium isotopic distribution in UF6 are critical for materials verification, accounting, and safeguards at enrichment plants. As nuclear fuel cycle technology becomes more prevalent around the world, international nuclearmore » safeguards and interest in UF6 enrichment assay has been growing. At present, laboratory-based mass spectrometry (MS), which offers the highest attainable analytical accuracy and precision, is the technique of choice for the analysis of stable and long-lived isotopes. Currently, the International Atomic Energy Agency (IAEA) monitors the production of enriched UF6 at declared facilities by collecting a small amount (between 1 to 10 g) of gaseous UF6 into a sample bottle, which is then shipped under chain of custody to a central laboratory (IAEA’s Nuclear Materials Analysis Laboratory) for high-precision isotopic assay by MS. The logistics are cumbersome and new shipping regulations are making it more difficult to transport UF6. Furthermore, the analysis is costly, and results are not available for some time after sample collection. Hence, the IAEA is challenged to develop effective safeguards approaches at enrichment plants. In-field isotopic analysis of UF6 has the potential to substantially reduce the time, logistics and expense of sample handling. However, current laboratory-based MS techniques require too much infrastructure and operator expertise for field deployment and operation. As outlined in the IAEA Department of Safeguards Long-Term R&D Plan, 2012–2023, one of the IAEA long-term R&D needs is to “develop tools and techniques to enable timely, potentially real-time, detection of HEU (Highly Enriched Uranium) production in LEU (Lowly Enriched Uranium) enrichment facilities” (Milestone 5.2). Because it is common that the next generation of analytical instruments is driven by technologies that are either currently available or just now emerging, one reasonable and practical approach to project the next generation of chemical instrumentation is to track the recent trends and to extrapolate them. This study adopted a similar approach, and an extensive literature review on existing and emerging technologies for UF6 enrichment assay was performed. The competitive advantages and current limitations of different analytical techniques for in-field UF6 enrichment assay were then compared, and the main gaps between needs and capabilities for their field use were examined. Subsequently, based on these results, technologies for the next-generation field-deployable instrument for UF6 enrichment assay were recommended. The study was organized in a way that a suite of assessment metric was first identified. Criteria used in this evaluation are presented in Section 1 of this report, and the most important ones are described briefly in the next few paragraphs. Because one driving force for in-field UF6 enrichment assay is related to the demanding transportation regulation for gaseous UF6, Section 2 contains a review of solid sorbents that convert and immobilized gaseous UF6 to a solid state, which is regarded as more transportation friendly and is less regulated. Furthermore, candidate solid sorbents, which show promise in mating with existing and emerging assay technologies, also factor into technology recommendations. Extensive literature reviews on existing and emerging technologies for UF6 enrichment assay, covering their scientific principles, instrument options, and current limitations are detailed in Sections 3 and 4, respectively. In Section 5, the technological gaps as well as start-of-the-art and commercial off-the-shelf components that can be adopted to expedite the development of a fieldable or portable UF6 enrichment-assay instrument are identified and discussed. Finally, based on the results of the review, requirements and recommendations for developing the next-generation field-deployable instrument for UF6 enrichment assay are presented in Section 6.« less

SOLVENT EXTRACTION PROCESS FOR SEPARATING URANIUM AND PLUTONIUM FROM AQUEOUS ACIDIC SOLUTIONS OF NEUTRON IRRADIATED URANIUM

DOEpatents

Bruce, F.R.

1962-07-24

A solvent extraction process was developed for separating actinide elements including plutonium and uranium from fission products. By this method the ion content of the acidic aqueous solution is adjusted so that it contains more equivalents of total metal ions than equivalents of nitrate ions. Under these conditions the extractability of fission products is greatly decreased. (AEC)

Uranium and its decay products in samples contaminated with uranium mine and mill waste

NASA Astrophysics Data System (ADS)

Benedik, L.; Klemencic, H.; Repinc, U.; Vrecek, P.

2003-05-01

The routine determination of the activity concentrations of uranium isotopes (^{238}U, ^{235}U and ^{234}U), thorium isotopes (^{212}Th, ^{230}TI, and ^{228}Th), ^{231}Pa, ^{226}Ra, ^{210}Pb and ^{210}Po in the environment is one of the most important tasks in uranium mining areas. Natural radionuclides contribute negligibly to the extemal radiation dose, but in the case of ingestion or inhalation can represent a very serious hazard. The objective of this study was to determine the activities of uranium and its decay products ^{230}Th, ^{231}Pa, ^{226}Ra, ^{210}Pb and ^{210}Po in sediments and water below sources of contamination (uranium mine, disposal sites and individual inflows) using gamma and alpha spectrometry, beta counting, the liquid scintillation technique and radiochemical neutron activation analysis.

Dating young groundwater with sulfur hexafluoride: Natural and anthropogenic sources of sulfur hexafluoride

USGS Publications Warehouse

Busenberg, Eurybiades; Plummer, Niel

2000-01-01

Sulfur hexafluoride (SF6) is primarily of anthropogenic origin but also occurs naturally. The troposphere concentration of SF6 has increased from a steady state value of 0.054±0.009 to more than 4 parts per trillion volume during the past 40 years. An analytical procedure was developed for measuring concentrations of SF6 to less than 0.01 fmol/L in water. Groundwater can be dated with SF6 if it is in equilibrium with atmospheric SF6 at the time of recharge and does not contain significant SF6 from other sources. The dating range of SF6 is currently 0 to 30 years. The tracer was successfully used to date shallow groundwater of the Atlantic Coastal Plain sand aquifers of the United States and springs issuing near the top of the Blue Ridge Mountains of Virginia. Significant concentrations of naturally occurring SF6 were found in some igneous, volcanic, and sedimentary rocks and in some hydrothermal fluids.

78 FR 20146 - Lost Creek ISR, LLC, Lost Creek Uranium In-Situ Recovery Project, Sweetwater County, Wyoming

Federal Register 2010, 2011, 2012, 2013, 2014

2013-04-03

... ISR, LLC, Lost Creek Uranium In-Situ Recovery Project, Sweetwater County, Wyoming AGENCY: Nuclear... to Source Materials License SUA-1598 for continued uranium production operations and in-situ recovery... identified in NUREG-1910, ``Generic Environmental Impact Statement for In-Situ Leach Uranium Milling...

Comparison of the Environment, Health, And Safety Characteristics of Advanced Thorium- Uranium and Uranium-Plutonium Fuel Cycles

NASA Astrophysics Data System (ADS)

Ault, Timothy M.

The environment, health, and safety properties of thorium-uranium-based (''thorium'') fuel cycles are estimated and compared to those of analogous uranium-plutonium-based (''uranium'') fuel cycle options. A structured assessment methodology for assessing and comparing fuel cycle is refined and applied to several reference fuel cycle options. Resource recovery as a measure of environmental sustainability for thorium is explored in depth in terms of resource availability, chemical processing requirements, and radiological impacts. A review of available experience and recent practices indicates that near-term thorium recovery will occur as a by-product of mining for other commodities, particularly titanium. The characterization of actively-mined global titanium, uranium, rare earth element, and iron deposits reveals that by-product thorium recovery would be sufficient to satisfy even the most intensive nuclear demand for thorium at least six times over. Chemical flowsheet analysis indicates that the consumption of strong acids and bases associated with thorium resource recovery is 3-4 times larger than for uranium recovery, with the comparison of other chemical types being less distinct. Radiologically, thorium recovery imparts about one order of magnitude larger of a collective occupational dose than uranium recovery. Moving to the entire fuel cycle, four fuel cycle options are compared: a limited-recycle (''modified-open'') uranium fuel cycle, a modified-open thorium fuel cycle, a full-recycle (''closed'') uranium fuel cycle, and a closed thorium fuel cycle. A combination of existing data and calculations using SCALE are used to develop material balances for the four fuel cycle options. The fuel cycle options are compared on the bases of resource sustainability, waste management (both low- and high-level waste, including used nuclear fuel), and occupational radiological impacts. At steady-state, occupational doses somewhat favor the closed thorium option while low-level waste volumes slightly favor the closed uranium option, although uncertainties are significant in both cases. The high-level waste properties (radioactivity, decay heat, and ingestion radiotoxicity) all significantly favor the closed fuel cycle options (especially the closed thorium option), but an alternative measure of key fission product inventories that drive risk in a repository slightly favors the uranium fuel cycles due to lower production of iodine-129. Resource requirements are much lower for the closed fuel cycle options and are relatively similar between thorium and uranium. In additional to the steady-state results, a variety of potential transition pathways are considered for both uranium and thorium fuel cycle end-states. For dose, low-level waste, and fission products contributing to repository risk, the differences among transition impacts largely reflected the steady-state differences. However, the HLW properties arrived at a distinctly opposite result in transition (strongly favoring uranium, whereas thorium was strongly favored at steady-state), because used present-day fuel is disposed without being recycled given that uranium-233, rather than plutonium, is the primarily fissile nuclide at the closed thorium fuel cycle's steady-state. Resource consumption was the only metric was strongly influenced by the specific transition pathway selected, favoring those pathways that more quickly arrived at steady-state through higher breeding ratio assumptions regardless of whether thorium or uranium was used.

Experience of on-site disposal of production uranium-graphite nuclear reactor.

PubMed

Pavliuk, Alexander O; Kotlyarevskiy, Sergey G; Bespala, Evgeny V; Zakharova, Elena V; Ermolaev, Vyacheslav M; Volkova, Anna G

2018-04-01

The paper reported the experience gained in the course of decommissioning EI-2 Production Uranium-Graphite Nuclear Reactor. EI-2 was a production Uranium-Graphite Nuclear Reactor located on the Production and Demonstration Center for Uranium-Graphite Reactors JSC (PDC UGR JSC) site of Seversk City, Tomsk Region, Russia. EI-2 commenced its operation in 1958, and was shut down on December 28, 1990, having operated for the period of 33 years all together. The extra pure grade graphite for the moderator, water for the coolant, and uranium metal for the fuel were used in the reactor. During the operation nitrogen gas was passed through the graphite stack of the reactor. In the process of decommissioning the PDC UGR JSC site the cavities in the reactor space were filled with clay-based materials. A specific composite barrier material based on clays and minerals of Siberian Region was developed for the purpose. Numerical modeling demonstrated the developed clay composite would make efficient geological barriers preventing release of radionuclides into the environment. Copyright © 2018 Elsevier Ltd. All rights reserved.

SEPARATION OF PLUTONIUM FROM URANIUM AND FISSION PRODUCTS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Boyd, G.E.; Adamson, A.W.; Schubert, J.

A chromatographic adsorption process is presented for the separation of plutonium from other fission products formed by the irradiation of uranium. The plutonium and the lighter element fission products are adsorbed on a sulfonated phenol-formaldehyde resin bed from a nitric acid solution containing the dissolved uranium. Successive washes of sulfuric, phosphoric, and nitric acids remove the bulk of the fission products, then an eluate of dilute phosphoric and nitric acids removes the remaining plutonium and fission products. The plutonium is selectively removed by passing this solution through zirconium phosphate, from which the plutonium is dissolved with nitric acid. This processmore » provides a convenient and efficient means for isolating plutonium.« less

Descriptive models of major uranium deposits in China - Some results of the Workshop on Uranium Resource Assessment sponsored by the International Atomic Energy Agency, Vienna, Austria, in cooperation with China National Nuclear Corporation, Beijing, and the U.S. Geological Survey, Denver, Colorado, and Reston, Virginia

USGS Publications Warehouse

Finch, W.I.; Feng, S.; Zuyi, C.; McCammon, R.B.

1993-01-01

Four major types of uranium deposits occur in China: granite, volcanic, sandstone, and carbonaceous-siliceous-pelitic rock. These types are major sources of uranium in many parts of the world and account for about 95 percent of Chinese production. Descriptive models for each of these types record the diagnostic regional and local geologic features of the deposits that are important to genetic studies, exploration, and resource assessment. A fifth type of uranium deposit, metasomatite, is also modeled because of its high potential for production. These five types of uranium deposits occur irregularly in five tectonic provinces distributed from the northwest through central to southern China. ?? 1993 Oxford University Press.

Production of fissioning uranium plasma to approximate gas-core reactor conditions

NASA Technical Reports Server (NTRS)

Lee, J. H.; Mcfarland, D. R.; Hohl, F.; Kim, K. H.

1974-01-01

The intense burst of neutrons from the d-d reaction in a plasma-focus apparatus is exploited to produce a fissioning uranium plasma. The plasma-focus apparatus consists of a pair of coaxial electrodes and is energized by a 25 kJ capacitor bank. A 15-g rod of 93% enriched U-235 is placed in the end of the center electrode where an intense electron beam impinges during the plasma-focus formation. The resulting uranium plasma is heated to about 5 eV. Fission reactions are induced in the uranium plasma by neutrons from the d-d reaction which were moderated by the polyethylene walls. The fission yield is determined by evaluating the gamma peaks of I-134, Cs-138, and other fission products, and it is found that more than 1,000,000 fissions are induced in the uranium for each focus formation, with at least 1% of these occurring in the uranium plasma.

Laser and gas centrifuge enrichment

NASA Astrophysics Data System (ADS)

Heinonen, Olli

2014-05-01

Principles of uranium isotope enrichment using various laser and gas centrifuge techniques are briefly discussed. Examples on production of high enriched uranium are given. Concerns regarding the possibility of using low end technologies to produce weapons grade uranium are explained. Based on current assessments commercial enrichment services are able to cover the global needs of enriched uranium in the foreseeable future.

IMPROVEMENTS IN OR RELATING TO THE PRODUCTION OF SINTERED URANIUM DIOXIDE

DOE Office of Scientific and Technical Information (OSTI.GOV)

Russell, L.E.; Harrison, J.D.L.; Brett, N.H.

A method is described for producing a dense sintered body of uranium dioxide or a mixture thereof with plutonium dioxide. Compacted uranium dioxide or a compacted uranium dioxide-plutonium dioxide mixture is heated to at least 1300 deg C in an atmosphere of carbon dioxide or carbon dioxide mixed with carbon monoxide. (R.J.S.)

Investigation of sulphur isotope variation due to different processes applied during uranium ore concentrate production.

PubMed

Krajkó, Judit; Varga, Zsolt; Wallenius, Maria; Mayer, Klaus; Konings, Rudy

The applicability and limitations of sulphur isotope ratio as a nuclear forensic signature have been studied. The typically applied leaching methods in uranium mining processes were simulated for five uranium ore samples and the n ( 34 S)/ n ( 32 S) ratios were measured. The sulphur isotope ratio variation during uranium ore concentrate (UOC) production was also followed using two real-life sample sets obtained from industrial UOC production facilities. Once the major source of sulphur is revealed, its appropriate application for origin assessment can be established. Our results confirm the previous assumption that process reagents have a significant effect on the n ( 34 S)/ n ( 32 S) ratio, thus the sulphur isotope ratio is in most cases a process-related signature.

Method of testing gas insulated systems for the presence of conducting particles utilizing a gas mixture of nitrogen and sulfur hexafluoride

DOEpatents

Wootton, Roy E.

1979-01-01

A method of testing a gas insulated system for the presence of conducting particles. The method includes inserting a gaseous mixture comprising about 98 volume percent nitrogen and about 2 volume percent sulfur hexafluoride into the gas insulated system at a pressure greater than 60 lb./sq. in. gauge, and then applying a test voltage to the system. If particles are present within the system, the gaseous mixture will break down, providing an indicator of the presence of the particles.

PROCESS FOR THE PRODUCTION OF AMMONIUM URANIUM FLUORIDE

DOEpatents

Ellis, A.S.; Mooney, R.B.

1953-08-25

This patent relates to the preparation of ammonium uranium fluoride. The process comprises adding a water soluble fluoride to an aqueous solution of a uranous compound containing an ammonium salt, and isolating the resulting precipitate. This patent relates to the manufacture of uranium tetnafluoride from ammonium uranium fluoride, NH/sub 4/UF/sub 5/. Uranium tetrafluoride is prepared by heating the ammonium uranium fluoride to a temperature at which dissociation occurs with liberation of ammonium fluoride. Preferably the process is carried out under reduced pressure, or in a current of an inert gas.

Process for alloying uranium and niobium

DOEpatents

Holcombe, Cressie E.; Northcutt, Jr., Walter G.; Masters, David R.; Chapman, Lloyd R.

1991-01-01

Alloys such as U-6Nb are prepared by forming a stacked sandwich array of uraniun sheets and niobium powder disposed in layers between the sheets, heating the array in a vacuum induction melting furnace to a temperature such as to melt the uranium, holding the resulting mixture at a temperature above the melting point of uranium until the niobium dissolves in the uranium, and casting the uranium-niobium solution. Compositional uniformity in the alloy product is enabled by use of the sandwich structure of uranium sheets and niobium powder.

PROCESSES FOR SEPARATING AND RECOVERING CONSTITUENTS OF NEUTRON IRRADIATED URANIUM

DOEpatents

Connick, R.E.; Gofman, J.W.; Pimentel, G.C.

1959-11-10

Processes are described for preparing plutonium, particularly processes of separating plutonium from uranium and fission products in neutron-irradiated uraniumcontaining matter. Specifically, plutonium solutions containing uranium, fission products and other impurities are contacted with reducing agents such as sulfur dioxide, uranous ion, hydroxyl ammonium chloride, hydrogen peroxide, and ferrous ion whereby the plutoninm is reduced to its fluoride-insoluble state. The reduced plutonium is then carried out of solution by precipitating niobic oxide therein. Uranium and certain fission products remain behind in the solution. Certain other fission products precipitate along with the plutonium. Subsequently, the plutonium and fission product precipitates are redissolved, and the solution is oxidized with oxidizing agents such as chlorine, peroxydisulfate ion in the presence of silver ion, permanganate ion, dichromate ion, ceric ion, and a bromate ion, whereby plutonium is oxidized to the fluoride-soluble state. The oxidized solution is once again treated with niobic oxide, thus precipitating the contamirant fission products along with the niobic oxide while the oxidized plutonium remains in solution. Plutonium is then recovered from the decontaminated solution.

Dupoly process for treatment of depleted uranium and production of beneficial end products

DOEpatents

Kalb, Paul D.; Adams, Jay W.; Lageraaen, Paul R.; Cooley, Carl R.

2000-02-29

The present invention provides a process of encapsulating depleted uranium by forming a homogenous mixture of depleted uranium and molten virgin or recycled thermoplastic polymer into desired shapes. Separate streams of depleted uranium and virgin or recycled thermoplastic polymer are simultaneously subjected to heating and mixing conditions. The heating and mixing conditions are provided by a thermokinetic mixer, continuous mixer or an extruder and preferably by a thermokinetic mixer or continuous mixer followed by an extruder. The resulting DUPoly shapes can be molded into radiation shielding material or can be used as counter weights for use in airplanes, helicopters, ships, missiles, armor or projectiles.

PROCESS FOR THE RECOVERY OF URANIUM FROM PHOSPHATIC ORE

DOEpatents

Long, R.L.

1959-04-14

A proccss is described for the recovery of uranium from phosphatic products derived from phosphatic ores. It has been discovered that certain alkyl phosphatic, derivatives can be employed in a direct solvent extraction operation to recover uranium from solid products, such as superphosphates, without first dissolving such solids. The organic extractants found suitable include alkyl derivatives of phosphoric, pyrophosphoric, phosof the derivative contains from 4 to 7 carbon atoms. A diluent such as kerosene is also used.

DECONTAMINATION OF URANIUM

DOEpatents

Spedding, F.H.; Butler, T.A.

1962-05-15

A process is given for separating fission products from uranium by extracting the former into molten aluminum. Phase isolation can be accomplished by selectively hydriding the uranium at between 200 and 300 deg C and separating the hydride powder from coarse particles of fissionproduct-containing aluminum. (AEC)

Routine inspection effort required for verification of a nuclear material production cutoff convention

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dougherty, D.; Fainberg, A.; Sanborn, J.

On 27 September 1993, President Clinton proposed {open_quotes}... a multilateral convention prohibiting the production of highly enriched uranium or plutonium for nuclear explosives purposes or outside of international safeguards.{close_quotes} The UN General Assembly subsequently adopted a resolution recommending negotiation of a non-discriminatory, multilateral, and internationally and effectively verifiable treaty (hereinafter referred to as {open_quotes}the Cutoff Convention{close_quotes}) banning the production of fissile material for nuclear weapons. The matter is now on the agenda of the Conference on Disarmament, although not yet under negotiation. This accord would, in effect, place all fissile material (defined as highly enriched uranium and plutonium) produced aftermore » entry into force (EIF) of the accord under international safeguards. {open_quotes}Production{close_quotes} would mean separation of the material in question from radioactive fission products, as in spent fuel reprocessing, or enrichment of uranium above the 20% level, which defines highly enriched uranium (HEU). Facilities where such production could occur would be safeguarded to verify that either such production is not occurring or that all material produced at these facilities is maintained under safeguards.« less

REGENERATION OF FISSION-PRODUCT-CONTAINING MAGNESIUM-THORIUM ALLOYS

DOEpatents

Chiotti, P.

1964-02-01

A process of regenerating a magnesium-thorium alloy contaminated with fission products, protactinium, and uranium is presented. A molten mixture of KCl--LiCl-MgCl/sub 2/ is added to the molten alloy whereby the alkali, alkaline parth, and rare earth fission products (including yttrium) and some of the thorium and uranium are chlorinated and

A METHOD OF PREPARING URANIUM DIOXIDE

DOEpatents

Scott, F.A.; Mudge, L.K.

1963-12-17

A process of purifying raw, in particular plutonium- and fission- products-containing, uranium dioxide is described. The uranium dioxide is dissolved in a molten chloride mixture containing potassium chloride plus sodium, lithium, magnesium, or lead chloride under anhydrous conditions; an electric current and a chlorinating gas are passed through the mixture whereby pure uranium dioxide is deposited on and at the same time partially redissolved from the cathode. (AEC)

Separation of Depleted Uranium From Soil

DTIC Science & Technology

2009-03-01

order to remove the metallic DU present in these soils. This procedure would re- duce the amount of time that metallic uranium could undergo corrosion ...slow corrosion is not sufficient to ignite the uranium . Unfired rod Weathered, unfired rod with yellow uranyl salt deposits Figure 1. Comparison...resulting in less downward movement. Interactions between uranium corrosion products and soil mineral and organic components can also affect

Process for recovering uranium from waste hydrocarbon oils containing the same. [Uranium contaminated lubricating oils from gaseous diffusion compressors

DOEpatents

Conrad, M.C.; Getz, P.A.; Hickman, J.E.; Payne, L.D.

1982-06-29

The invention is a process for the recovery of uranium from uranium-bearing hydrocarbon oils containing carboxylic acid as a degradation product. In one aspect, the invention comprises providing an emulsion of water and the oil, heating the same to a temperature effecting conversion of the emulsion to an organic phase and to an acidic aqueous phase containing uranium carboxylate, and recovering the uranium from the aqueous phase. The process is effective, simple and comparatively inexpensive. It avoids the use of toxic reagents and the formation of undesirable intermediates.

Harnessing redox activity for the formation of uranium tris(imido) compounds

NASA Astrophysics Data System (ADS)

Anderson, Nickolas H.; Odoh, Samuel O.; Yao, Yiyi; Williams, Ursula J.; Schaefer, Brian A.; Kiernicki, John J.; Lewis, Andrew J.; Goshert, Mitchell D.; Fanwick, Phillip E.; Schelter, Eric J.; Walensky, Justin R.; Gagliardi, Laura; Bart, Suzanne C.

2014-10-01

Classically, late transition-metal organometallic compounds promote multielectron processes solely through the change in oxidation state of the metal centre. In contrast, uranium typically undergoes single-electron chemistry. However, using redox-active ligands can engage multielectron reactivity at this metal in analogy to transition metals. Here we show that a redox-flexible pyridine(diimine) ligand can stabilize a series of highly reduced uranium coordination complexes by storing one, two or three electrons in the ligand. These species reduce organoazides easily to form uranium-nitrogen multiple bonds with the release of dinitrogen. The extent of ligand reduction dictates the formation of uranium mono-, bis- and tris(imido) products. Spectroscopic and structural characterization of these compounds supports the idea that electrons are stored in the ligand framework and used in subsequent reactivity. Computational analyses of the uranium imido products probed their molecular and electronic structures, which facilitated a comparison between the bonding in the tris(imido) structure and its tris(oxo) analogue.

Volatile fluoride process for separating plutonium from other materials

DOEpatents

Spedding, F. H.; Newton, A. S.

1959-04-14

The separation of plutonium from uranium and/or fission products by formation of the higher fluorides off uranium and/or plutonium is described. Neutronirradiated uranium metal is first converted to the hydride. This hydrided product is then treated with fluorine at about 315 deg C to form and volatilize UF/sub 6/ leaving plutonium behind. Thc plutonium may then be separated by reacting the residue with fluorine at about 5004DEC and collecting the volatile plutonium fluoride thus formed.

VOLATILE FLUORIDE PROCESS FOR SEPARATING PLUTONIUM FROM OTHER MATERIALS

DOEpatents

Spedding, F.H.; Newton, A.S.

1959-04-14

The separation of plutonium from uranium and/or tission products by formation of the higher fluorides of uranium and/or plutonium is discussed. Neutronirradiated uranium metal is first convcrted to the hydride. This hydrided product is then treatced with fluorine at about 315 deg C to form and volatilize UF/sup 6/ leaving plutonium behind. The plutonium may then be separated by reacting the residue with fluorine at about 500 deg C and collecting the volatile plutonium fluoride thus formed.

ARSENATE CARRIER PRECIPITATION METHOD OF SEPARATING PLUTONIUM FROM NEUTRON IRRADIATED URANIUM AND RADIOACTIVE FISSION PRODUCTS

DOEpatents

Thompson, S.G.; Miller, D.R.; James, R.A.

1961-06-20

A process is described for precipitating Pu from an aqueous solution as the arsenate, either per se or on a bismuth arsenate carrier, whereby a separation from uranium and fission products, if present in solution, is accomplished.

DUPoly process for treatment of depleted uranium and production of beneficial end products

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kalb, P.D.; Adams, J.W.; Lageraaen, P.R.

2000-02-29

The present invention provides a process of encapsulating depleted uranium by forming a homogeneous mixture of depleted uranium and molten virgin or recycled thermoplastic polymer into desired shapes. Separate streams of depleted uranium and virgin or recycled thermoplastic polymer are simultaneously subjected to heating and mixing conditions. The heating and mixing conditions are provided by a thermokinetic mixer, continuous mixer or an extruder and preferably by a thermokinetic mixer or continuous mixer followed by an extruder. The resulting DUPoly shapes can be molded into radiation shielding material or can be used as counter weights for use in airplanes, helicopters, ships,more » missiles, armor or projectiles.« less

Radioactivity of Consumer Products

NASA Astrophysics Data System (ADS)

Peterson, David; Jokisch, Derek; Fulmer, Philip

2006-11-01

A variety of consumer products and household items contain varying amounts of radioactivity. Examples of these items include: FiestaWare and similar glazed china, salt substitute, bananas, brazil nuts, lantern mantles, smoke detectors and depression glass. Many of these items contain natural sources of radioactivity such as Uranium, Thorium, Radium and Potassium. A few contain man-made sources like Americium. This presentation will detail the sources and relative radioactivity of these items (including demonstrations). Further, measurements of the isotopic ratios of Uranium-235 and Uranium-238 in several pieces of china will be compared to historical uses of natural and depleted Uranium. Finally, the presenters will discuss radiation safety as it pertains to the use of these items.

In-situ, time resolved monitoring of uranium in BFS:OPC grout. Part 2: Corrosion in water.

PubMed

Stitt, C A; Paraskevoulakos, C; Banos, A; Harker, N J; Hallam, K R; Pullin, H; Davenport, A; Street, S; Scott, T B

2018-06-18

To reflect potential conditions in a geological disposal facility, uranium was encapsulated in grout and submersed in de-ionised water for time periods between 2-47 weeks. Synchrotron X-ray Powder Diffraction and X-ray Tomography were used to identify the dominant corrosion products and measure their dimensions. Uranium dioxide was observed as the dominant corrosion product and time dependent thickness measurements were used to calculate oxidation rates. The effectiveness of physical and chemical grout properties to uranium corrosion and mobilisation is discussed and Inductively Coupled Plasma Mass Spectrometry was used to measure 238 U (aq) content in the residual water of several samples.

PROCESS OF PREPARING A FLUORIDE OF TETRAVLENT URANIUM

DOEpatents

Wheelwright, E.J.

1959-02-17

A method is described for producing a fluoride salt pf tetravalent uranium suitable for bomb reduction to metallic uranium. An aqueous solution of uranyl nitrate is treated with acetic acid and a nitrite-suppressor and then contacted with metallic lead whereby uranium is reduced from the hexavalent to the tetravalent state and soluble lead acetate is formed. Sulfate ions are then added to the solution to precipitate and remove the lead values. Hydrofluoric acid and alkali metal ions are then added causing the formation of an alkali metal uranium double-fluoride in which the uranium is in the tetravalent state. After recovery, this precipitate is suitable for using in the limited production of metallic uranium.

Industrial Production of Uranium; PRODUCCION INDUSTRIAL DE URANIO

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pedregal, J.D.

1956-01-01

The purity requirements of uranium and the necessity of purifying the uranium compounds previous to its metallurgical treatment are briefly discussed as an introduction to the different methods reduction to the metal. The methods which are used by the Junta de Energia Nuclear are indicated. (tr-auth)

PYROCHEMICAL DECONTAMINATION METHOD FOR REACTOR FUEL

DOEpatents

Buyers, A.G.

1959-06-30

A pyro-chemical method is presented for decontaminating neutron irradiated uranium and separating plutonium therefrom by contact in the molten state with a metal chloride salt. Uranium trichloride and uranium tetrachloride either alone or in admixture with alkaline metal and alkaline eanth metal fluorides under specified temperature and specified phase ratio conditions extract substantially all of the uranium from the irradiated uranium fuel together with certain fission products. The phases are then separated leaving purified uranium metal. The uranium and plutonium in the salt phase can be reduced to forin a highly decontaminated uraniumplutonium alloy. The present method possesses advantages for economically decontaminating irradiated nuclear fuel elements since irradiated fuel may be proccessed immediately after withdrawal from the reactor and the uranium need not be dissolved and later reduced to the metallic form. Accordingly, the uranium may be economically refabricated and reinserted into the reactor.

Modeling and control study of the NASA 0.3-meter transonic cryogenic tunnel for use with sulfur hexafluoride medium

NASA Technical Reports Server (NTRS)

Balakrishna, S.; Kilgore, W. Allen

1992-01-01

The NASA Langley 0.3-m Transonic Cryogenic Tunnel is to be modified to operate with sulfur hexafluoride gas while retaining its present capability to operate with nitrogen. The modified tunnel will provide high Reynolds number flow on aerodynamic models with two different test gases. The document details a study of the SF6 tunnel performance boundaries, thermodynamic modeling of the tunnel process, nonlinear dynamical simulation of math model to yield tunnel responses, the closed loop control requirements, control laws, and mechanization of the control laws on the microprocessor based controller.

Method for recovery of actinides from refractory oxides thereof using O.sub. F.sub.2

DOEpatents

Asprey, Larned B.; Eller, Phillip G.

1988-01-01

Method for recovery of actinides from nuclear waste material containing sintered and other oxides thereof using O.sub.2 F.sub.2 to generate the hexafluorides of the actinides present therein. The fluorinating agent, O.sub.2 F.sub.2, has been observed to perform the above-described tasks at sufficiently low temperatures that there is virtually no damage to the containment vessels. Moreover, the resulting actinide hexafluorides are not destroyed by high temperature reactions with the walls of the reaction vessel. Dioxygen difluoride is readily prepared, stored and transferred to the place of reaction.

Measurement of cardiac output using improved chromatographic analysis of sulfur hexafluoride (SF6).

PubMed

Klocke, F J; Roberts, D L; Farhi, E R; Naughton, B J; Sekovski, B; Klocke, R A

1977-06-01

A constant current variable frequency pulsed electron capture detector has been incorporated into the gas chromatographic analysis of trace amounts of sulfur hexafluoride (SF6) in water and blood. The resulting system offers a broader effective operating range than more conventional electron capture units and has been utilized for measurements of cardiac output employing constant-rate infusion of dissolved SF6. The SF6 technique has been validated against direct volumetric measurements of cardiac output in a canine right-heart bypass preparation and used subsequently for rapidly repeated measurements in conscious animals and man.

Method of cleaning plastics using super and subcritical media

DOEpatents

Sawan, Samuel P.; Spall, W. Dale; Talhi, Abdelhafid

1998-05-26

A method for treating a plastic, such as polyethylene or polypropylene, to remove at least a portion of at least one contaminant includes combining the plastic with a supercritical medium, such as carbon dioxide or sulfur hexafluoride, whereby at least a portion of the contaminant dissolves in the supercritical medium. Alternatively, the plastic can be combined with a suitable liquid medium, such as carbon dioxide or liquid sulfur hexafluoride. At least a portion of the medium, containing the dissolved contaminant, is separated from the plastic, thereby removing at least a portion of the contaminant from the plastic.

Method of cleaning plastics using super and subcritical media

DOEpatents

Sawan, S.P.; Spall, W.D.; Talhi, A.

1998-05-26

A method for treating a plastic, such as polyethylene or polypropylene, to remove at least a portion of at least one contaminant includes combining the plastic with a supercritical medium, such as carbon dioxide or sulfur hexafluoride, whereby at least a portion of the contaminant dissolves in the supercritical medium. Alternatively, the plastic can be combined with a suitable liquid medium, such as carbon dioxide or liquid sulfur hexafluoride. At least a portion of the medium, containing the dissolved contaminant, is separated from the plastic, thereby removing at least a portion of the contaminant from the plastic. 10 figs.

Uranium, its impact on the national and global energy mix; and its history, distribution, production, nuclear fuel-cycle, future, and relation to the environment

USGS Publications Warehouse

Finch, Warren Irvin

1997-01-01

The many aspects of uranium, a heavy radioactive metal used to generate electricity throughout the world, are briefly described in relatively simple terms intended for the lay reader. An adequate glossary of unfamiliar terms is given. Uranium is a new source of electrical energy developed since 1950, and how we harness energy from it is explained. It competes with the organic coal, oil, and gas fuels as shown graphically. Uranium resources and production for the world are tabulated and discussed by country and for various energy regions in the United States. Locations of major uranium deposits and power reactors in the United States are mapped. The nuclear fuel-cycle of uranium for a typical light-water reactor is illustrated at the front end-beginning with its natural geologic occurrence in rocks through discovery, mining, and milling; separation of the scarce isotope U-235, its enrichment, and manufacture into fuel rods for power reactors to generate electricity-and at the back end-the reprocessing and handling of the spent fuel. Environmental concerns with the entire fuel cycle are addressed. The future of the use of uranium in new, simplified, 'passively safe' reactors for the utility industry is examined. The present resource assessment of uranium in the United States is out of date, and a new assessment could aid the domestic uranium industry.

Method for the production of uranium chloride salt

DOEpatents

Westphal, Brian R.; Mariani, Robert D.

2013-07-02

A method for the production of UCl.sub.3 salt without the use of hazardous chemicals or multiple apparatuses for synthesis and purification is provided. Uranium metal is combined in a reaction vessel with a metal chloride and a eutectic salt- and heated to a first temperature under vacuum conditions to promote reaction of the uranium metal with the metal chloride for the production of a UCl.sub.3 salt. After the reaction has run substantially to completion, the furnace is heated to a second temperature under vacuum conditions. The second temperature is sufficiently high to selectively vaporize the chloride salts and distill them into a condenser region.

Uranium in Surface Waters and Sediments Affected by Historical Mining in the Denver West 1:100,000 Quadrangle, Colorado

USGS Publications Warehouse

Zielinski, Robert A.; Otton, James K.; Schumann, R. Randall; Wirt, Laurie

2008-01-01

Geochemical sampling of 82 stream waters and 87 stream sediments within mountainous areas immediately west of Denver, Colorado, was conducted by the U.S. Geological Survey in October 1994. The primary purpose was to evaluate regionally the effects of geology and past mining on the concentration and distribution of uranium. The study area contains uranium- and thorium-rich bedrock, numerous noneconomic occurrences of uranium minerals, and several uranium deposits of variable size and production history. During the sampling period, local streams had low discharge and were more susceptible to uranium-bearing acid drainage originating from historical mines of base- and precious-metal sulfides. Results indicated that the spatial distribution of Precambrian granites and metamorphic rocks strongly influences the concentration of uranium in stream sediments. Within-stream transport increases the dispersion of uranium- and thorium rich mineral grains derived primarily from granitic source rocks. Dissolved uranium occurs predominantly as uranyl carbonate complexes, and concentrations ranged from less than 1 to 65 micrograms per liter. Most values were less than 5 micrograms per liter, which is less than the current drinking water standard of 30 micrograms per liter and much less than locally applied aquatic-life toxicity standards of several hundred micrograms per liter. In local streams that are affected by uranium-bearing acid mine drainage, dissolved uranium is moderated by dilution and sorptive uptake by stream sediments. Sorbents include mineral alteration products and chemical precipitates of iron- and aluminum-oxyhydroxides, which form where acid drainage enters streams and is neutralized. Suspended uranium is relatively abundant in some stream segments affected by nearby acid drainage, which likely represents mobilization of these chemical precipitates. The 234U/238U activity ratio of acid drainage (0.95-1.0) is distinct from that of local surface waters (more than 1.05), and this distinctive isotopic composition may be preserved in iron-oxyhydroxide precipitates of acid drainage origin. The study area includes a particularly large vein-type uranium deposit (Schwartzwalder mine) with past uranium production. Stream water and sediment collected downstream from the mine's surface operations have locally anomalous concentrations of uranium. Fine-grained sediments downstream from the mine contain rare minute particles (10-20 micrometers) of uraninite, which is unstable in a stream environment and thus probably of recent origin related to mining. Additional rare particles of very fine grained (less than 5 micrometer) barite likely entered the stream as discharge from settling ponds in which barite precipitation was formerly used to scavenge dissolved radium from mine effluent.

Electron-beam-driven RI separator for SCRIT (ERIS) at RIKEN RI beam factory

NASA Astrophysics Data System (ADS)

Ohnishi, T.; Ichikawa, S.; Koizumi, K.; Kurita, K.; Miyashita, Y.; Ogawara, R.; Tamaki, S.; Togasaki, M.; Wakasugi, M.

2013-12-01

We constructed a radioactive isotope (RI) separator named ERIS (electron-beam-driven RI separator for SCRIT) for the SCRIT (Self-Confinement RI Target) electron scattering facility at RIKEN RI Beam Factory (RIBF). In ERIS, production rate of fission products in the photofission of uranium is estimated to be 2.2 ×1011 fissions/s with 30 g of uranium and a 1-kW electron beam. During the commissioning of ERIS, the mass resolution and overall efficiency, including ionization, extraction, and transmission, were found to be 1660 and 21%, respectively, using natural xenon gas. The preparation of uranium carbide (UC2) RI production targets is described from which a 132Sn beam was successfully separated in our first attempt at RI production.

Simulation of uranium and plutonium oxides compounds obtained in plasma

NASA Astrophysics Data System (ADS)

Novoselov, Ivan Yu.; Karengin, Alexander G.; Babaev, Renat G.

2018-03-01

The aim of this paper is to carry out thermodynamic simulation of mixed plutonium and uranium oxides compounds obtained after plasma treatment of plutonium and uranium nitrates and to determine optimal water-salt-organic mixture composition as well as conditions for their plasma treatment (temperature, air mass fraction). Authors conclude that it needs to complete the treatment of nitric solutions in form of water-salt-organic mixtures to guarantee energy saving obtainment of oxide compounds for mixed-oxide fuel and explain the choice of chemical composition of water-salt-organic mixture. It has been confirmed that temperature of 1200 °C is optimal to practice the process. Authors have demonstrated that condensed products after plasma treatment of water-salt-organic mixture contains targeted products (uranium and plutonium oxides) and gaseous products are environmental friendly. In conclusion basic operational modes for practicing the process are showed.

Review of the NURE assessment of the U.S. Gulf Coast Uranium Province

USGS Publications Warehouse

Hall, Susan M.

2013-01-01

Historic exploration and development were used to evaluate the reliability of domestic uranium reserves and potential resources estimated by the U.S. Department of Energy national uranium resource evaluation (NURE) program in the U.S. Gulf Coast Uranium Province. NURE estimated 87 million pounds of reserves in the $30/lb U3O8 cost category in the Coast Plain uranium resource region, most in the Gulf Coast Uranium Province. Since NURE, 40 million pounds of reserves have been mined, and 38 million pounds are estimated to remain in place as of 2012, accounting for all but 9 million pounds of U3O8 in the reserve or production categories in the NURE estimate. Considering the complexities and uncertainties of the analysis, this study indicates that the NURE reserve estimates for the province were accurate. An unconditional potential resource of 1.4 billion pounds of U3O8, 600 million pounds of U3O8 in the forward cost category of $30/lb U3O8 (1980 prices), was estimated in 106 favorable areas by the NURE program in the province. Removing potential resources from the non-productive Houston embayment, and those reserves estimated below historic and current mining depths reduces the unconditional potential resource 33% to about 930 million pounds of U3O8, and that in the $30/lb cost category 34% to 399 million pounds of U3O8. Based on production records and reserve estimates tabulated for the region, most of the production since 1980 is likely from the reserves identified by NURE. The potential resource predicted by NURE has not been developed, likely due to a variety of factors related to the low uranium prices that have prevailed since 1980.

Thermodynamic Simulation of Equilibrium Composition of Reaction Products at Dehydration of a Technological Channel in a Uranium-Graphite Reactor

NASA Astrophysics Data System (ADS)

Pavliuk, A. O.; Zagumennov, V. S.; Kotlyarevskiy, S. G.; Bespala, E. V.

2018-01-01

The problems of accumulation of nuclear fuel spills in the graphite stack in the course of operation of uranium-graphite nuclear reactors are considered. The results of thermodynamic analysis of the processes in the graphite stack at dehydration of a technological channel, fuel element shell unsealing and migration of fission products, and activation of stable nuclides in structural elements of the reactor and actinides inside the graphite moderator are given. The main chemical reactions and compounds that are produced in these modes in the reactor channel during its operation and that may be hazardous after its shutdown and decommissioning are presented. Thermodynamic simulation of the equilibrium composition is performed using the specialized code TERRA. The results of thermodynamic simulation of the equilibrium composition in different cases of technological channel dehydration in the course of the reactor operation show that, if the temperature inside the active core of the nuclear reactor increases to the melting temperature of the fuel element, oxides and carbides of nuclear fuel are produced. The mathematical model of the nonstationary heat transfer in a graphite stack of a uranium-graphite reactor in the case of the technological channel dehydration is presented. The results of calculated temperature evolution at the center of the fuel element, the replaceable graphite element, the air gap, and in the surface layer of the block graphite are given. The numerical results show that, in the case of dehydration of the technological channel in the uranium-graphite reactor with metallic uranium, the main reaction product is uranium dioxide UO2 in the condensed phase. Low probability of production of pyrophoric uranium compounds (UH3) in the graphite stack is proven, which allows one to disassemble the graphite stack without the risk of spontaneous graphite ignition in the course of decommissioning of the uranium-graphite nuclear reactor.

Removal of uranium from aqueous HF solutions

DOEpatents

Pulley, Howard; Seltzer, Steven F.

1980-01-01

This invention is a simple and effective method for removing uranium from aqueous HF solutions containing trace quantities of the same. The method comprises contacting the solution with particulate calcium fluoride to form uranium-bearing particulates, permitting the particulates to settle, and separting the solution from the settled particulates. The CaF.sub.2 is selected to have a nitrogen surface area in a selected range and is employed in an amount providing a calcium fluoride/uranium weight ratio in a selected range. As applied to dilute HF solutions containing 120 ppm uranium, the method removes at least 92% of the uranium, without introducing contaminants to the product solution.

Process for measuring degradation of sulfur hexafluoride in high voltage systems

DOEpatents

Sauers, Isidor

1986-01-01

This invention is a method of detecting the presence of toxic and corrosive by-products in high voltage systems produced by electrically induced degradation of SF.sub.6 insulating gas in the presence of certain impurities. It is an improvement over previous methods because it is extremely sensitive, detecting by-products present in parts per billion concentrations, and because the device employed is of a simple design and takes advantage of the by-products natural affinity for fluoride ions. The method employs an ion-molecule reaction cell in which negative ions of the by-products are produced by fluorine attachment. These ions are admitted to a negative ion mass spectrometer and identified by their spectra. This spectrometry technique is an improvement over conventional techniques because the negative ion peaks are strong and not obscured by a major ion spectra of the SF.sub.6 component as is the case in positive ion mass spectrometry.

Process for measuring degradation of sulfur hexafluoride in high voltage systems

DOEpatents

Sauers, I.

1985-04-23

This invention is a method of detecting the presence of toxic and corrosive by-products in high voltage systems produced by electrically induced degradation of SF/sub 6/ insulating gas in the presence of certain impurities. It is an improvement over previous methods because it is extremely sensitive, detecting by-products present in parts per billion concentrations, and because the device employed is of a simple design and takes advantage of the by-products natural affinity for fluoride ions. The method employs an ion-molecule reaction cell in which negative ions of the by-products are produced by fluorine attachment. These ions are admitted to a negative ion mass spectrometer and identified by their spectra. This spectrometry technique is an improvement over conventional techniques because the negative ion peaks are strong and not obscured by a major ion spectra of the SF/sub 6/ component as is the case in positive ion mass spectrometry.

Gulf War Illnesses: Understanding of Health Effects From Depleted Uranium Evolving But Safety Training Needed

DTIC Science & Technology

2000-03-01

against enemy munitions. Depleted uranium is a low- level radioactive heavy metal , and concerns have surfaced about whether exposure to it could be a...radioactive heavy metal , the potential for health effects are twofold: effects from radiation and effects from chemical toxicity. Two recent expert...depleted uranium safety training. Background Depleted uranium (DU), a low-level radioactive heavy metal , is a by- product of the process used to

Abandoned Uranium Mine (AUM) Points, Navajo Nation, 2016, US EPA Region 9

EPA Pesticide Factsheets

This GIS dataset contains point features of all Abandoned Uranium Mines (AUMs) on or within one mile of the Navajo Nation. Points are centroids developed from the Navajo Nation production mines polygon dataset that comprise of productive or unproductive Abandoned Uranium Mines. Attributes include mine names, aliases, links to AUM reports, indicators whether an AUM was mined above or below ground, indicators whether an AUM was mined above or below the local water table, and the region in which an AUM is located. This dataset contains 608 features.

Alternatives to argon for gas stopping volumes in the B194 neutron imager

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bleuel, D. L.; Anderson, S.; Caggiano, J. A.

2017-05-17

In a recent experiment at Lawrence Berkeley National Laboratory, the 40Ar(d,p)41Ar excitation function between 3-7 MeV was measured, confirming a previous estimation that there may be an intolerable radiation dose from 41Ar production by slowing to rest 6.74 MeV deuterons in the gas cell of the neutron imaging facility being constructed in B194. Gas alternatives to argon are considered, including helium, nitrogen (N2), neon, sulfur hexafluoride (SF6), krypton, and xenon, as well as high atomic number solid backings such as tantalum.

Preliminary report on uranium and thorium content of intrusive rocks in northeastern Washington and northern Idaho

DOE Office of Scientific and Technical Information (OSTI.GOV)

Castor, S.B.; Berry, M.R.; Robins, J.W.

1977-11-01

This study delineates favorable areas for uranium resources in northeastern Washington and northern Idaho by identifying granitic rocks with relatively large amounts of uranium and (or) thorium. Results are based on analysis of 344 rock samples. Uranium analyses obtained by gamma-ray spectrometric data correlate closely with fluorometric determinations. On the basis of cumulative frequency distribution curves, more than 8 ppM equivalent uranium and more than 20 ppM equivalent thorium are considered anomalous for granitic rocks in northeastern Washington and northern Idaho. Granitic rocks anomalously high in uranium and (or) thorium are concentrated in two northeast-trending belts. The most prominent, themore » Midnite-Hall Mountain belt, includes the Midnite and Sherwood uranium mines, and two lesser but productive areas farther north. This belt follows the contact between Precambrian and Paleozoic rocks, which is also the locus of the Kootenai arc fold belt. The second belt of anomalously radioactive granitic rocks is along the Republic graben, a prominent linear structure in an area with no recorded uranium production. Anomalously radioactive granitic rocks are generally massive quartz monzonite, alaskite, or pegmatite, which contain abundant quartz and potash feldspar. They are also characterized by pink potash feldspar, commonly as large phenocrysts, and by the presence of muscovite. Several uranium and thorium minerals have been identified in these rocks. The two belts of anomalously radioactive plutons are considered favorable for uranium resources. Deposits could occur in the intrusive rocks themselves or in favorable environments in adjacent rocks. 13 figs., 2 tables.« less

Use of Sodium Dithionite as Part of a More Efficient Groundwater Restoration Method Following In-situ Recovery of Uranium at the Smith-Ranch Highland Site in Wyoming

NASA Astrophysics Data System (ADS)

Harris, R.; Reimus, P. W.; Ware, D.; Williams, K.; Chu, D.; Perkins, G.; Migdissov, A. A.; Bonwell, C.

2017-12-01

Uranium is primarily mined for nuclear power production using an aqueous extraction technique called in-situ recovery (ISR). ISR can pollute groundwater with residual uranium and other heavy metals. Reverse osmosis and groundwater sweep are currently used to restore groundwater after ISR mining, but are not permanent solutions. Sodium dithionite is being tested as part of a method to more permanently restore groundwater after ISR mining at the Smith-Ranch Highland site in Wyoming. Sodium dithionite is a chemical reductant that can reduce sediments that were oxidized during ISR. The reduced sediments can reduce soluble uranium (VI) in the groundwater to insoluble uranium (IV). Laboratory studies that use sodium dithionite to treat sediments and waters from the site may help predict how it will behave during a field deployment. An aqueous batch experiment showed that sodium dithionite reduced uranium in post-mined untreated groundwater from 38 ppm to less than 1 ppm after 1 day. A sediment reduction batch experiment showed that sodium dithionite-treated sediments were capable of reducing uranium in post-mined untreated groundwater from 38 ppm to 2 ppm after 7 days. One column experiment is showing post-mined sodium dithionite-treated sediments are capable of reducing uranium in post-mined groundwater for over 30 pore volumes past the initial injection. While these results are promising for field deployments of sodium dithionite, another column experiment with sodium dithionite-treated sediments containing uranium rich organic matter is showing net production of uranium instead of uranium uptake. Sodium dithionite appears to liberate uranium from the organic matter. Another sediment reduction experiment is being conducted to further investigate this hypothesis. These experiments are helping guide plans for field deployments of sodium dithionite at uranium ISR mining sites.

Prediction of external corrosion for steel cylinders at the Paducah Gaseous Diffusion Plant: Application of an empirical method

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lyon, B.F.

1996-02-01

During the summer of 1995, ultrasonic wall thickness data were collected for 100 steel cylinders containing depleted uranium (DU) hexafluoride located at Paducah Gaseous Diffusion Plant (PGDP) in Paducah, Kentucky. The cylinders were selected for measurement to assess the condition of the more vulnerable portion of the cylinder inventory at PGDP. The purpose of this report is to apply the method used in Lyon to estimate the effects of corrosion for larger unsampled populations as a function of time. The scope of this report is limited and is not intended to represent the final analyses of available data. Future effortsmore » will include continuing analyses of available data to investigate defensible deviations from the conservative assumptions made to date. For each cylinder population considered, two basic types of analyses were conducted: (1) estimates were made of the number of cylinders as a function of time that will have a minimum wall thickness of either 0 mils (1 mil = 0.00 1 in.) or 250 mils and (2) the current minimum wall thickness distributions across cylinders were estimated for each cylinder population considered. Additional analyses were also performed investigating comparisons of the results for F and G yards with the results presented in Lyon (1995).« less

Neutron-rich isotope production using the uranium carbide multi-foil SPES target prototype

NASA Astrophysics Data System (ADS)

Scarpa, D.; Biasetto, L.; Corradetti, S.; Manzolaro, M.; Andrighetto, A.; Carturan, S.; Prete, G.; Zanonato, P.; Stracener, D. W.

2011-03-01

In the framework of the R&D program for the SPES (Selective Production of Exotic Species) project of the Istituto Nazionale di Fisica Nucleare (INFN), production yields of neutron-rich isotopes have been measured at the Holifield Radioactive Ion Beam Facility (HRIBF, Oak Ridge National Laboratory, USA). This experiment makes use of the multi-foil SPES target prototype composed of 7 uranium carbide discs, with excess of graphite (ratio C/ U = 4 . 77 isotopes of medium mass (between 72 and 141amu), produced via proton-induced fission of uranium using a 40MeV proton beam, have been collected and analyzed for the target heated at 2000 ° C target temperature.

Electrochemical method of producing eutectic uranium alloy and apparatus

DOEpatents

Horton, James A.; Hayden, H. Wayne

1995-01-01

An apparatus and method for continuous production of liquid uranium alloys through the electrolytic reduction of uranium chlorides. The apparatus includes an electrochemical cell formed from an anode shaped to form an electrolyte reservoir, a cathode comprising a metal, such as iron, capable of forming a eutectic uranium alloy having a melting point less than the melting point of pure uranium, and molten electrolyte in the reservoir comprising a chlorine or fluorine containing salt and uranium chloride. The method of the invention produces an eutectic uranium alloy by creating an electrolyte reservoir defined by a container comprising an anode, placing an electrolyte in the reservoir, the electrolyte comprising a chlorine or fluorine containing salt and uranium chloride in molten form, positioning a cathode in the reservoir where the cathode comprises a metal capable of forming an uranium alloy having a melting point less than the melting point of pure uranium, and applying a current between the cathode and the anode.

Electrochemical method of producing eutectic uranium alloy and apparatus

DOEpatents

Horton, J.A.; Hayden, H.W.

1995-01-10

An apparatus and method are disclosed for continuous production of liquid uranium alloys through the electrolytic reduction of uranium chlorides. The apparatus includes an electrochemical cell formed from an anode shaped to form an electrolyte reservoir, a cathode comprising a metal, such as iron, capable of forming a eutectic uranium alloy having a melting point less than the melting point of pure uranium, and molten electrolyte in the reservoir comprising a chlorine or fluorine containing salt and uranium chloride. The method of the invention produces an eutectic uranium alloy by creating an electrolyte reservoir defined by a container comprising an anode, placing an electrolyte in the reservoir, the electrolyte comprising a chlorine or fluorine containing salt and uranium chloride in molten form, positioning a cathode in the reservoir where the cathode comprises a metal capable of forming an uranium alloy having a melting point less than the melting point of pure uranium, and applying a current between the cathode and the anode. 2 figures.

Selective recovery of uranium from Ca-Mg uranates by chlorination

NASA Astrophysics Data System (ADS)

Pomiro, Federico J.; Gaviría, Juan P.; Quinteros, Raúl D.; Bohé, Ana E.

2017-07-01

A chlorination process is proposed for the uranium extraction and separation using Calciumsbnd Magnesium uranates such as starting reactants which were obtained by precipitation from uranyl nitrate solutions with calcium hydroxide. The study is based on thermodynamic and reaction analysis using chlorine gas as chlorination agent. The results showed that the chlorination reaction of Ca uranate is more feasible to occur than the Mg uranate. The products obtained after chlorination reactions were washed with deionized water to remove the chlorides produced and analyzed. The XRD patterns of the washed products indicated that the chlorination between 400 and 500 °C result in a single phase of calcium uranate (CaUO4) as reaction product. The formation of U3O8 and MgU3O10 was observed at temperatures between 600 °C and 700 °C for 8 hs. The optimal conditions to recover uranium were 3 l h-1 of chlorine and 10 hs of reaction at 700 °C being U3O8 the single uranium product obtained.

Progress on the chemical separation of fission fragments from 236Np produced by proton irradiation of natural uranium target

NASA Astrophysics Data System (ADS)

Larijani, C.; Jerome, S. M.; Lorusso, G.; Ivanov, P.; Russell, B.; Pearce, A. K.; Regan, P. H.

2017-11-01

The aim of the current work is to develop and validate a radiochemical separation scheme capable of separating both 236gNp and 236Pu from a uranium target of natural isotopic composition ( 1 g uranium) and 200 MBq of fission decay products. A target containing 1.2 g of UO2 was irradiated with a beam of 25 MeV protons with a typical beam current of 30 μA for 19 h in December 2013 at the University of Birmingham Cyclotron facility. Using literature values for the production cross-section for fusion of protons with uranium targets, we estimate that an upper limit of approximately 250 Bq of activity from the 236Np ground state was produced in this experiment. Using a radiochemical separation scheme, Np and Pu fractions were separated from the produced fission decay products, with analyses of the target-based final reaction products made using Inductively Couple Plasma Mass Spectrometry (ICP-MS) and high-resolution α particle and γ-ray spectrometry.

On the Nature of the Cherdyntsev-Chalov Effect

NASA Astrophysics Data System (ADS)

Timashev, S. F.

2018-06-01

It is shown that the Cherdyntsev-Chalov effect, usually presented as the separation of even isotopes of uranium upon their transition from the solid to the liquid phase, can include initiated acceleration of the radioactive decay of uranium-238 nuclei during the formation of cracks in geologically (seismic and volcanically) active zones of the Earth's crust. The fissuring of the solid-phase medium leads to an increase in mechanical tensile stress and the emergence of strong local electric fields, resulting in the injection of chemical-scale high-energy electrons into the aqueous phase of the cracks. Under these conditions, the e - catalytic decay of uranium-238 nucleus studied earlier can occur during the formation of metastable protactinium-238 nuclei with locally distorted nucleon structure, which subequently undergo β-decay with the formation of thorium-234 and helium-4 nuclei as products of the fission of the initial uranium-238 nucleus with a characteristic period of several years. The observed increased activity of uranium-234 nuclei that form during the subsequent β-decay of thorium and then protactinium is associated with the initiated fission of uranium-238. The possibility is discussed of developing thermal power by using existing wastes from uranium production that contain uranium-238 to activate this isotope through the mechanochemical processing of these wastes in aqueous media with the formation of 91 238 Pa isu , the half-life of which is several years.

Method for fluorination of actinide fluorides and oxyfluorides using O/sub 2/F/sub 2/

DOEpatents

Eller, P.G.; Malm, J.G.; Penneman, R.A.

1984-08-01

The present invention relates generally to methods of fluorination and more particularly to the use of O/sub 2/F/sub 2/ for the preparation of actinide hexafluorides, and for the extraction of deposited actinides and fluorides and oxyfluorides thereof from reaction vessels. The experiments set forth hereinabove demonstrate that the room temperature or below use of O/sub 2/F/sub 2/ will be highly beneficial for the preparation of pure actinide hexafluorides from their respective tetrafluorides without traces of HF being present as occurs using other fluorinating agents: and decontamination of equipment previously exposed to actinides: e.g., walls, feed lines, etc.

Evaluation of an optoacoustic based gas analysing device

NASA Astrophysics Data System (ADS)

Markmann, Janine; Lange, Birgit; Theisen-Kunde, Dirk; Danicke, Veit; Mayorov, Fedor; Eckert, Sebastian; Kettmann, Pascal; Brinkmann, Ralf

2017-07-01

The relative occurrence of volatile organic compounds in the human respiratory gas is disease-specific (ppb range). A prototype of a gas analysing device using two tuneable laser systems, an OPO-laser (2.5 to 10 μm) and a CO2-laser (9 to 11 μm), and an optoacoustic measurement cell was developed to detect concentrations in the ppb range. The sensitivity and resolution of the system was determined by test gas measurements, measuring ethylene and sulfur hexafluoride with the CO2-laser and butane with the OPO-laser. System sensitivity found to be 13 ppb for sulfur hexafluoride, 17 ppb for ethylene and <10 ppb for butane, with a resolution of 50 ppb at minimum for sulfur hexafluoride. Respiratory gas samples of 8 healthy volunteers were investigated by irradiation with 17 laser lines of the CO2-laser. Several of those lines overlap with strong absorption bands of ammonia. As it is known that ammonia concentration increases by age a separation of people <35 und >35 was striven for. To evaluate the data the first seven gas samples were used to train a discriminant analysis algorithm. The eighth subject was then assigned correctly to the group >35 years with the age of 49 years.

Nuclear waste viewed in a new light; a synchrotron study of uranium encapsulated in grout.

PubMed

Stitt, C A; Hart, M; Harker, N J; Hallam, K R; MacFarlane, J; Banos, A; Paraskevoulakos, C; Butcher, E; Padovani, C; Scott, T B

2015-03-21

How do you characterise the contents of a sealed nuclear waste package without breaking it open? This question is important when the contained corrosion products are potentially reactive with air and radioactive. Synchrotron X-rays have been used to perform micro-scale in-situ observation and characterisation of uranium encapsulated in grout; a simulation for a typical intermediate level waste storage packet. X-ray tomography and X-ray powder diffraction generated both qualitative and quantitative data from a grout-encapsulated uranium sample before, and after, deliberately constrained H2 corrosion. Tomographic reconstructions provided a means of assessing the extent, rates and character of the corrosion reactions by comparing the relative densities between the materials and the volume of reaction products. The oxidation of uranium in grout was found to follow the anoxic U+H2O oxidation regime, and the pore network within the grout was observed to influence the growth of uranium hydride sites across the metal surface. Powder diffraction analysis identified the corrosion products as UO2 and UH3, and permitted measurement of corrosion-induced strain. Together, X-ray tomography and diffraction provide means of accurately determining the types and extent of uranium corrosion occurring, thereby offering a future tool for isolating and studying the reactions occurring in real full-scale waste package systems. Copyright © 2014 Elsevier B.V. All rights reserved.

PROCESSING OF NEUTRON-IRRADIATED URANIUM

DOEpatents

Hopkins, H.H. Jr.

1960-09-01

An improved "Purex" process for separating uranium, plutonium, and fission products from nitric acid solutions of neutron-irradiated uranium is offered. Uranium is first extracted into tributyl phosphate (TBP) away from plutonium and fission products after adjustment of the acidity from 0.3 to 0.5 M and heating from 60 to 70 deg C. Coextracted plutonium, ruthenium, and fission products are fractionally removed from the TBP by three scrubbing steps with a 0.5 M nitric acid solution of ferrous sulfamate (FSA), from 3.5 to 5 M nitric acid, and water, respectively, and the purified uranium is finally recovered from the TBP by precipitation with an aqueous solution of oxalic acid. The plutonium in the 0.3 to 0.5 M acid solution is oxidized to the tetravalent state with sodium nitrite and extracted into TBP containing a small amount of dibutyl phosphate (DBP). Plutonium is then back-extracted from the TBP-DBP mixture with a nitric acid solution of FSA, reoxidized with sodium nitrite in the aqueous strip solution obtained, and once more extracted with TBP alone. Finally the plutonium is stripped from the TBP with dilute acid, and a portion of the strip solution thus obtained is recycled into the TBPDBP for further purification.

METHOD OF IMPREGNATING A POROUS MATERIAL

DOEpatents

Steele, G.N.

1960-06-01

A method of impregnating a porous body with an inorganic uranium- containing salt is outlined and comprises dissolving a water-soluble uranium- containing salt in water; saturating the intercommunicating pores of the porous body with the salt solution; infusing ammonia gas into the intercommunicating pores of the body, the ammonia gas in water chemically reacting with the water- soluble uranium-containing salt in the water solvent to form a nonwater-soluble uranium-containing precipitant; and evaporating the volatile unprecipitated products from the intercommunicating pores whereby the uranium-containing precipitate is uniformly distributed in the intercommunicating peres of the porous body.

PRODUCTION OF URANIUM

DOEpatents

Ruehle, A.E.; Stevenson, J.W.

1957-11-12

An improved process is described for the magnesium reduction of UF/sub 4/ to produce uranium metal. In the past, there have been undesirable premature reactions between the Mg and the bomb liner or the UF/sub 4/ before the actual ignition of the bomb reaction. Since these premature reactions impair the yield of uranium metal, they have been inhibited by forming a protective film upon the particles of Mg by reacting it with hydrated uranium tetrafluoride, sodium bifluoride, uranyl fluoride, or uranium trioxide. This may be accomplished by adding about 0.5 to 2% of the additive to the bomb charge.

The Problem of Preconcentration of Uranium Ores by Physical Processes; LES PROBLEMES DE LA PRECONCENTRATION DES MINERAIS D'URANIUM PAR VOIE PHYSIQUE. LE TRIAGE ELECTRONIQUE

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vuchot, L.; Ginocchio, A. et al.

1959-10-31

As uranium ores, like most other ores, are not definite substances which can be treated directly for the production of the metal, the ores must be concentrated. The common physical processes used for all ores, such as sieving, gravimetric separation, flotation, electromagnetic separation, and electrostatic separation, are applicable to the beneficiation of uranium. The radioactivity of uranium ores has led to a radiometric method for the concentration. This method is described in detail. As an example, the preconcentration of Forez ores is discussed. (J.S.R.)

An Investigation on the Persistence of Uranium Hydride during Storage of Simulant Nuclear Waste Packages.

PubMed

Stitt, C A; Harker, N J; Hallam, K R; Paraskevoulakos, C; Banos, A; Rennie, S; Jowsey, J; Scott, T B

2015-01-01

Synchrotron X-rays have been used to study the oxidation of uranium and uranium hydride when encapsulated in grout and stored in de-ionised water for 10 months. Periodic synchrotron X-ray tomography and X-ray powder diffraction have allowed measurement and identification of the arising corrosion products and the rates of corrosion. The oxidation rates of the uranium metal and uranium hydride were slower than empirically derived rates previously reported for each reactant in an anoxic water system, but without encapsulation in grout. This was attributed to the grout acting as a physical barrier limiting the access of oxidising species to the uranium surface. Uranium hydride was observed to persist throughout the 10 month storage period and industrial consequences of this observed persistence are discussed.

URANIUM RECOVERY AND PURIFICATION PROCESS AND PRODUCTION OF HIGH PURITY URANIUM TETRAFLUORIDE

DOEpatents

Bailes, R.H.; Long, R.S.; Grinstead, R.R.

1957-09-17

A process is described wherein an anionic exchange technique is employed to separate uramium from a large variety of impurities. Very efficient and economical purification of contamimated uranium can be achieved by treatment of the contaminated uranium to produce a solution containing a high concentration of chloride. Under these conditions the uranium exists as an aniomic chloride complex. Then the uranium chloride complex is adsorbed from the solution on an aniomic exchange resin, whereby a portion of the impurities remain in the solution and others are retained with the uramium by the resin. The adsorbed impurities are then removed by washing the resin with pure concentrated hydrochloric acid, after which operation the uranium is eluted with pure water yielding an acidic uranyl chloride solution of high purity.

ELECTROLYSIS OF THORIUM AND URANIUM

DOEpatents

Hansen, W.N.

1960-09-01

An electrolytic method is given for obtaining pure thorium, uranium, and thorium-uranium alloys. The electrolytic cell comprises a cathode composed of a metal selected from the class consisting of zinc, cadmium, tin, lead, antimony, and bismuth, an anode composed of at least one of the metals selected from the group consisting of thorium and uranium in an impure state, and an electrolyte composed of a fused salt containing at least one of the salts of the metals selected from the class consisting of thorium, uranium. zinc, cadmium, tin, lead, antimony, and bismuth. Electrolysis of the fused salt while the cathode is maintained in the molten condition deposits thorium, uranium, or thorium-uranium alloys in pure form in the molten cathode which thereafter may be separated from the molten cathode product by distillation.

An Investigation on the Persistence of Uranium Hydride during Storage of Simulant Nuclear Waste Packages

PubMed Central

Harker, N. J.; Hallam, K. R.; Paraskevoulakos, C.; Banos, A.; Rennie, S.; Jowsey, J.

2015-01-01

Synchrotron X-rays have been used to study the oxidation of uranium and uranium hydride when encapsulated in grout and stored in de-ionised water for 10 months. Periodic synchrotron X-ray tomography and X-ray powder diffraction have allowed measurement and identification of the arising corrosion products and the rates of corrosion. The oxidation rates of the uranium metal and uranium hydride were slower than empirically derived rates previously reported for each reactant in an anoxic water system, but without encapsulation in grout. This was attributed to the grout acting as a physical barrier limiting the access of oxidising species to the uranium surface. Uranium hydride was observed to persist throughout the 10 month storage period and industrial consequences of this observed persistence are discussed. PMID:26176551

[Effect of the chelator BPCBG on the decorporation of uranium in vivo and uranium-induced damage of human renal tubular epithelial cells in vitro].

PubMed

Bao, Yi-zhong; Wang, Dan; Hu, Yu-xing; Xu, Ai-hong; Sun, Mei-zhen; Chen, Hong-hong

2011-11-01

This study is to assess the efficacy of BPCBG on the decorporation of uranium (VI) and protecting human renal proximal tubular epithelial cells (HK-2) against uranium-induced damage. BPCBG at different doses was injected intramuscularly to male SD rats immediately after a single intraperitoneal injection of UO2(CH3COO)2. Twenty-four hours later uranium contents in urine, kidneys and femurs were measured by ICP-MS. After HK-2 cells were exposed to UO2(CH3COO)2 immediately or for 24 h followed by BPCBG treatment at different doses for another 24 or 48 h, the uranium contents in HK-2 cells were measured by ICP-MS, the cell survival was assayed by cell counting kit-8 assay, formation of micronuclei was determined by the cytokinesis-block (CB) micronucleus assay and the production of intracellular reactive oxygen species (ROS) was detected by 2',7'-dichlorofluorescin diacetate (DCFH-DA) oxidation. DTPA-CaNa3 was used as control. It was found that BPCBG at dosages of 60, 120, and 600 micromol kg(-1) resulted in 37%-61% increase in 24 h-urinary uranium excretion, and significantly decreased the amount of uranium retention in kidney and bone to 41%-31% and 86%-42% of uranium-treated group, respectively. After HK-2 cells that had been pre-treated with UO2(CH3COO)2 for 24 h were treated with the chelators for another 24 h, 55%-60% of the intracellular uranium was removed by 10-250 micromol L(-1) of BPCBG. Treatment of uranium-treated HK-2 cells with BPCBG significantly enhanced the cell survival, decreased the formation of micronuclei and inhibited the production of intracellular ROS. Although DTPA-CaNa3 markedly reduced the uranium retention in kidney of rats and HK-2 cells, its efficacy of uranium removal from body was significantly lower than that of BPCBG and it could not protect uranium-induced cell damage. It can be concluded that BPCBG effectively decorporated the uranium from UO2(CH3COO)2-treated rats and HK-2 cells, which was better than DTPA-CaNa3. It could also scavenge the uranium-induced intracellular ROS and protect against the uranium-induced cell damage. BPCBG is worth further investigation.

RECOVERY OF Pu VALUES BY FLUORINATION AND FRACTIONATION

DOEpatents

Brown, H.S.; Webster, D.S.

1959-01-20

A method is presented for the concentration and recovery of plutonium by fluorination and fractionation. A metallic mass containing uranium and plutonium is heated to 250 C and contacted with a stream of elemental fluorine. After fluorination of the metallic mass, the rcaction products are withdrawn and subjected to a distillation treatment to separate the fluorination products of uranium and to obtain a residue containing the fluorination products of plutonium.

800-MeV proton irradiation of thorium and depleted uranium targets

DOE Office of Scientific and Technical Information (OSTI.GOV)

Russell, G.J.; Brun, T.O.; Pitcher, E.J.

As part of the Los Alamos Fertile-to-Fissile-Conversion (FERFICON) program in the late 1980`s, thick targets of the fertile materials thorium and depleted uranium were bombarded by 800-MeV protons to produce the fissile materials {sup 233}U and {sup 239}Pu, respectively. The amount of {sup 233}U made was determined by measuring the {sup 233}Pa activity, and the yield of {sup 239}Pu was deduced by measuring the activity of {sup 239}Np. For the thorium target, 4 spallation products and 34 fission products were also measured. For the depleted uranium target, 3 spallation products and 16 fission products were also measured. The number ofmore » fissions in each target was deduced from fission product mass-yield curves. In actuality, axial distributions of the products were measured, and the distributions were then integrated over the target volume to obtain the total number of products for each reaction.« less

Investigating Freshwater Periphyton Community Response to Uranium with Phospholipid Fatty Acid and Denaturing Gradient Gel Electrophoresis Analyses

DOE Office of Scientific and Technical Information (OSTI.GOV)

Small, Jack A.; Bunn, Amoret L.; McKinstry, Craig A.

2008-04-01

Periphyton communities can be used as monitors of ecosystem health and as indicators of contamination in lotic systems. Measures of biomass, community structure and genetic diversity were used to investigate impacts of uranium exposure on periphyton. Laboratory exposures of periphyton in river water amended with uranium were performed for 5 days, followed by 2 days of uranium depuration in unamended river water. Productivity as measured by biomass was not affected by concentrations up to 100 µg L-1 uranium. Phospholipid fatty acid (PLFA) profiles and denaturing gradient gel electrophoresis (DGGE) banding patterns found no changes in community or genetic structure relatedmore » to uranium exposure. We suggest that the periphyton community as a whole is not impacted by exposures of uranium up to a dose of 100 µg L-1. These findings have significance for the assessment and prediction of uranium impacts on aquatic ecosystems.« less

METHOD OF OPERATING NUCLEAR REACTORS

DOEpatents

Untermyer, S.

1958-10-14

A method is presented for obtaining enhanced utilization of natural uranium in heavy water moderated nuclear reactors by charging the reactor with an equal number of fuel elements formed of natural uranium and of fuel elements formed of uranium depleted in U/sup 235/ to the extent that the combination will just support a chain reaction. The reactor is operated until the rate of burnup of plutonium equals its rate of production, the fuel elements are processed to recover plutonium, the depleted uranium is discarded, and the remaining uranium is formed into fuel elements. These fuel elements are charged into a reactor along with an equal number of fuel elements formed of uranium depleted in U/sup 235/ to the extent that the combination will just support a chain reaction, and reuse of the uranium is continued as aforesaid until it wlll no longer support a chain reaction when combined with an equal quantity of natural uranium.

Applied technology for mine waste water decontamination in the uranium ores extraction from Romania

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bejenaru, C.; Filip, G.; Vacariu, V.T.

1996-12-31

The exploitation of uranium ores in Romania is carried out in underground mines. In all exploited uranium deposits, mine waste waters results and will still result after the closure of uranium ore extraction activity. The mine waters are radioactively contaminated with uranium and its decay products being a hazard both for underground waters as for the environment. This paper present the results of research work carried out by authors for uranium elimination from waste waters as the problems involved during the exploitation process of the existent equipment as its maintenance in good experimental conditions. The main waste water characteristics aremore » discussed: solids as suspension, uranium, radium, mineral salts, pH, etc. The moist suitable way to eliminate uranium from mine waste waters is the ion exchange process based on ion exchangers in fluidized bed. A flowsheet is given with main advantages resulted.« less

Biota of uranium mill tailings near the Black Hills

Treesearch

Mark A. Rumble

1982-01-01

Reclamation" often implies the enhancement of the land as wildlife habitat or for other productive uses. However, there are situations where revegetation to stabilize erosion is the only desired goal. Uranium mining and mill sites may fall into this later category. Data pertaining to plant and animal components on revegetated uranium mill tailings was collected....

METHOD FOR SEPARATION OF PLUTONIUM FROM URANIUM AND FISSION PRODUCTS BY SOLVENT EXTRACTION

DOEpatents

Seaborg, G.T.; Blaedel, W.J.; Walling, M.T. Jr.

1960-08-23

A process is given for separating from each other uranium, plutonium, and fission products in an aqueous nitric acid solution by the so-called Redox process. The plutonium is first oxidized to the hexavalent state, e.g., with a water-soluble dichromate or sodium bismuthate, preferably together with a holding oxidant such as potassium bromate. potassium permanganate, or an excess of the oxidizing agent. The solution is then contacted with a water-immiscible organic solvent, preferably hexone. whereby uranium and plutonium are extracted while the fission products remain in the aqueous solution. The separated organic phase is then contacted with an aqueous solution of a reducing agent, with or without a holding reductant (e.g., with a ferrous salt plus hydrazine or with ferrous sulfamate), whereby plutonium is reduced to the trivalent state and back- extracted into the aqueous solution. The uranium may finally be back-extracted from the organic solvent (e.g., with a 0.1 N nitric acid).

Method for making a uranium chloride salt product

DOEpatents

Miller, William E [Naperville, IL; Tomczuk, Zygmunt [Lockport, IL

2004-10-05

The subject apparatus provides a means to produce UCl.sub.3 in large quantities without incurring corrosion of the containment vessel or associated apparatus. Gaseous Cl is injected into a lower layer of Cd where CdCl.sub.2 is formed. Due to is lower density, the CdCl.sub.2 rises through the Cd layer into a layer of molten LiCl--KCL salt where a rotatable basket containing uranium ingots is suspended. The CdCl.sub.2 reacts with the uranium to form UCl.sub.3 and Cd. Due to density differences, the Cd sinks down to the liquid Cd layer and is reused. The UCl.sub.3 combines with the molten salt. During production the temperature is maintained at about 600.degree. C. while after the uranium has been depleted the salt temperature is lowered, the molten salt is pressure siphoned from the vessel, and the salt product LiCl--KCl-30 mol % UCl.sub.3 is solidified.

ALLOY COATINGS AND METHOD OF APPLYING

DOEpatents

Eubank, L.D.; Boller, E.R.

1958-08-26

A method for providing uranium articles with a pro tective coating by a single dip coating process is presented. The uranium article is dipped into a molten zinc bath containing a small percentage of aluminum. The resultant product is a uranium article covered with a thin undercoat consisting of a uranium-aluminum alloy with a small amount of zinc, and an outer layer consisting of zinc and aluminum. The article may be used as is, or aluminum sheathing may then be bonded to the aluminum zinc outer layer.

Method of increasing the deterrent to proliferation of nuclear fuels

DOEpatents

Rampolla, Donald S.

1982-01-01

A process of recycling protactinium-231 to enhance the utilization of radioactively hot uranium-232 in nuclear fuel for the purpose of making both fresh and spent fuel more resistant to proliferation. The uranium-232 may be obtained by the irradiation of protactinium-231 which is normally found in the spent fuel rods of a thorium base nuclear reactor. The production of protactinium-231 and uranium-232 would be made possible by the use of the thorium uranium-233 fuel cycle in power reactors.

Magnesium transport extraction of transuranium elements from LWR fuel

DOEpatents

Ackerman, John P.; Battles, James E.; Johnson, Terry R.; Miller, William E.; Pierce, R. Dean

1992-01-01

A process of separating transuranium actinide values from uranium values present in spent nuclear oxide fuels which contain rare earth and noble metal fission products. The oxide fuel is reduced with Ca metal in the presence of CaCl.sub.2 and a U-Fe alloy containing not less than about 84% by weight uranium at a temperature in the range of from about 800.degree. C. to about 850.degree. C. to produce additional uranium metal which dissolves in the U-Fe alloy raising the uranium concentration and having transuranium actinide metals and rare earth fission product metals and the noble metal fission products dissolved therein. The CaCl.sub.2 having CaO and fission products of alkali metals and the alkali earth metals and iodine dissolved therein is separated and electrolytically treated with a carbon electrode to reduce the CaO to Ca metal while converting the carbon electrode to CO and CO.sub.2. The Ca metal and CaCl.sub.2 is recycled to reduce additional oxide fuel. The U-Fe alloy having transuranium actinide metals and rare earth fission product metals and the noble metal fission products dissolved therein is contacted with Mg metal which takes up the actinide and rare earth fission product metals. The U-Fe alloy retains the noble metal fission products and is stored while the Mg is distilled and recycled leaving the transuranium actinide and rare earth fission products isolated.

Actinide metal processing

DOEpatents

Sauer, Nancy N.; Watkin, John G.

1992-01-01

A process of converting an actinide metal such as thorium, uranium, or plnium to an actinide oxide material by admixing the actinide metal in an aqueous medium with a hypochlorite as an oxidizing agent for sufficient time to form the actinide oxide material and recovering the actinide oxide material is provided together with a low temperature process of preparing an actinide oxide nitrate such as uranyl nitrte. Additionally, a composition of matter comprising the reaction product of uranium metal and sodium hypochlorite is provided, the reaction product being an essentially insoluble uranium oxide material suitable for disposal or long term storage.

Actinide metal processing

DOEpatents

Sauer, N.N.; Watkin, J.G.

1992-03-24

A process for converting an actinide metal such as thorium, uranium, or plutonium to an actinide oxide material by admixing the actinide metal in an aqueous medium with a hypochlorite as an oxidizing agent for sufficient time to form the actinide oxide material and recovering the actinide oxide material is described together with a low temperature process for preparing an actinide oxide nitrate such as uranyl nitrate. Additionally, a composition of matter comprising the reaction product of uranium metal and sodium hypochlorite is provided, the reaction product being an essentially insoluble uranium oxide material suitable for disposal or long term storage.

PROCESS FOR PRODUCTION OF URANIUM

DOEpatents

Crawford, J.W.C.

1959-09-29

A process is described for the production of uranium by the autothermic reduction of an anhydrous uranium halide with an alkaline earth metal, preferably magnesium One feature is the initial reduction step which is brought about by locally bringing to reaction temperature a portion of a mixture of the reactants in an open reaction vessel having in contact with the mixture a lining of substantial thickness composed of calcium fluoride. The lining is prepared by coating the interior surface with a plastic mixture of calcium fluoride and water and subsequently heating the coating in situ until at last the exposed surface is substantially anhydrous.

Uranium chloride extraction of transuranium elements from LWR fuel

DOEpatents

Miller, W.E.; Ackerman, J.P.; Battles, J.E.; Johnson, T.R.; Pierce, R.D.

1992-08-25

A process of separating transuranium actinide values from uranium values present in spent nuclear oxide fuels containing rare earth and noble metal fission products as well as other fission products is disclosed. The oxide fuel is reduced with Ca metal in the presence of Ca chloride and a U-Fe alloy which is liquid at about 800 C to dissolve uranium metal and the noble metal fission product metals and transuranium actinide metals and rare earth fission product metals leaving Ca chloride having CaO and fission products of alkali metals and the alkali earth metals and iodine dissolved therein. The Ca chloride and CaO and the fission products contained therein are separated from the U-Fe alloy and the metal values dissolved therein. The U-Fe alloy having dissolved therein reduced metals from the spent nuclear fuel is contacted with a mixture of one or more alkali metal or alkaline earth metal halides selected from the class consisting of alkali metal or alkaline earth metal and Fe or U halide or a combination thereof to transfer transuranium actinide metals and rare earth metals to the halide salt leaving the uranium and some noble metal fission products in the U-Fe alloy and thereafter separating the halide salt and the transuranium metals dissolved therein from the U-Fe alloy and the metals dissolved therein. 1 figure.

Uranium chloride extraction of transuranium elements from LWR fuel

DOEpatents

Miller, William E.; Ackerman, John P.; Battles, James E.; Johnson, Terry R.; Pierce, R. Dean

1992-01-01

A process of separating transuranium actinide values from uranium values present in spent nuclear oxide fuels containing rare earth and noble metal fission products as well as other fission products is disclosed. The oxide fuel is reduced with Ca metal in the presence of Ca chloride and a U-Fe alloy which is liquid at about 800.degree. C. to dissolve uranium metal and the noble metal fission product metals and transuranium actinide metals and rare earth fission product metals leaving Ca chloride having CaO and fission products of alkali metals and the alkali earth metals and iodine dissolved therein. The Ca chloride and CaO and the fission products contained therein are separated from the U-Fe alloy and the metal values dissolved therein. The U-Fe alloy having dissolved therein reduced metals from the spent nuclear fuel is contacted with a mixture of one or more alkali metal or alkaline earth metal halides selected from the class consisting of alkali metal or alkaline earth metal and Fe or U halide or a combination thereof to transfer transuranium actinide metals and rare earth metals to the halide salt leaving the uranium and some noble metal fission products in the U-Fe alloy and thereafter separating the halide salt and the transuranium metals dissolved therein from the U-Fe alloy and the metals dissolved therein.

Nuclear reactor fuel structure containing uranium alloy wires embedded in a metallic matrix plate

DOEpatents

Travelli, A.

1985-10-25

A flat or curved plate structure, to be used as fuel in a nuclear reactor, comprises elongated fissionable wires or strips embedded in a metallic continuous non-fissionable matrix plate. The wires or strips are made predominantly of a malleable uranium alloy, such as uranium silicide, uranium gallide or uranium germanide. The matrix plate is made predominantly of aluminum or an aluminum alloy. The wires or strips are located in a single row at the midsurface of the plate, parallel with one another and with the length dimension of the plate. The wires or strips are separated from each other, and from the surface of the plate, by sufficient thicknesses of matrix material, to provide structural integrity and effective fission product retention, under neutron irradiation. This construction makes it safely feasible to provide a high uranium density, so that the uranium enrichment with uranium 235 may be reduced below about 20%, to deter the reprocessing of the uranium for use in nuclear weapons.

Nuclear reactor fuel structure containing uranium alloy wires embedded in a metallic matrix plate

DOEpatents

Travelli, Armando

1988-01-01

A flat or curved plate structure, to be used as fuel in a nuclear reactor, comprises elongated fissionable wires or strips embedded in a metallic continuous non-fissionable matrix plate. The wires or strips are made predominantly of a malleable uranium alloy, such as uranium silicide, uranium gallide or uranium germanide. The matrix plate is made predominantly of aluminum or an aluminum alloy. The wires or strips are located in a single row at the midsurface of the plate, parallel with one another and with the length dimension of the plate. The wires or strips are separated from each other, and from the surface of the plate, by sufficient thicknesses of matrix material, to provide structural integrity and effective fission product retention, under neutron irradiation. This construction makes it safely feasible to provide a high uranium density, so that the uranium enrichment with uranium 235 may be reduced below about 20%, to deter the reprocessing of the uranium for use in nuclear weapons.

FISSION PRODUCT REMOVAL FROM ORGANIC SOLUTIONS

DOEpatents

Moore, R.H.

1960-05-10

The decontamination of organic solvents from fission products and in particular the treatment of solvents that were used for the extraction of uranium and/or plutonium from aqueous acid solutions of neutron-irradiated uranium are treated. The process broadly comprises heating manganese carbonate in air to a temperature of between 300 and 500 deg C whereby manganese dioxide is formed; mixing the manganese dioxide with the fission product-containing organic solvent to be treated whereby the fission products are precipitated on the manganese dioxide; and separating the fission product-containing manganese dioxide from the solvent.

Effect of sulfur hexafluoride gas and post-annealing treatment for inductively coupled plasma etched barium titanate thin films

PubMed Central

2014-01-01

Aerosol deposition- (AD) derived barium titanate (BTO) micropatterns are etched via SF6/O2/Ar plasmas using inductively coupled plasma (ICP) etching technology. The reaction mechanisms of the sulfur hexafluoride on BTO thin films and the effects of annealing treatment are verified through X-ray photoelectron spectroscopy (XPS) analysis, which confirms the accumulation of reaction products on the etched surface due to the low volatility of the reaction products, such as Ba and Ti fluorides, and these residues could be completely removed by the post-annealing treatment. The exact peak positions and chemicals shifts of Ba 3d, Ti 2p, O 1 s, and F 1 s are deduced by fitting the XPS narrow-scan spectra on as-deposited, etched, and post-annealed BTO surfaces. Compared to the as-deposited BTOs, the etched Ba 3d 5/2 , Ba 3d 3/2 , Ti 2p 3/2 , Ti 2p 1/2 , and O 1 s peaks shift towards higher binding energy regions by amounts of 0.55, 0.45, 0.4, 0.35, and 0.85 eV, respectively. A comparison of the as-deposited film with the post-annealed film after etching revealed that there are no significant differences in the fitted XPS narrow-scan spectra except for the slight chemical shift in the O 1 s peak due to the oxygen vacancy compensation in O2-excessive atmosphere. It is inferred that the electrical properties of the etched BTO film can be restored by post-annealing treatment after the etching process. Moreover, the relative permittivity and loss tangent of the post-annealed BTO thin films are remarkably improved by 232% and 2,695%, respectively. PMID:25249824

Studies on separation and purification of fission (99)Mo from neutron activated uranium aluminum alloy.

PubMed

Rao, Ankita; Kumar Sharma, Abhishek; Kumar, Pradeep; Charyulu, M M; Tomar, B S; Ramakumar, K L

2014-07-01

A new method has been developed for separation and purification of fission (99)Mo from neutron activated uranium-aluminum alloy. Alkali dissolution of the irradiated target (100mg) results in aluminum along with (99)Mo and a few fission products passing into solution, while most of the fission products, activation products and uranium remain undissolved. Subsequent purification steps involve precipitation of aluminum as Al(OH)3, iodine as AgI/AgIO3 and molybdenum as Mo-α-benzoin oxime. Ruthenium is separated by volatilization as RuO4 and final purification of (99)Mo was carried out using anion exchange method. The radiochemical yield of fission (99)Mo was found to be >80% and the purity of the product was in conformity with the international pharmacopoeia standards. Copyright © 2014 Elsevier Ltd. All rights reserved.

Continuous process electrorefiner

DOEpatents

Herceg, Joseph E [Naperville, IL; Saiveau, James G [Hickory Hills, IL; Krajtl, Lubomir [Woodridge, IL

2006-08-29

A new device is provided for the electrorefining of uranium in spent metallic nuclear fuels by the separation of unreacted zirconium, noble metal fission products, transuranic elements, and uranium from spent fuel rods. The process comprises an electrorefiner cell. The cell includes a drum-shaped cathode horizontally immersed about half-way into an electrolyte salt bath. A conveyor belt comprising segmented perforated metal plates transports spent fuel into the salt bath. The anode comprises the conveyor belt, the containment vessel, and the spent fuel. Uranium and transuranic elements such as plutonium (Pu) are oxidized at the anode, and, subsequently, the uranium is reduced to uranium metal at the cathode. A mechanical cutter above the surface of the salt bath removes the deposited uranium metal from the cathode.

Review of the NURE Assessment of the U.S. Gulf Coast Uranium Province

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hall, Susan M., E-mail: SusanHall@usgs.gov

2013-09-15

Historic exploration and development were used to evaluate the reliability of domestic uranium reserves and potential resources estimated by the U.S. Department of Energy national uranium resource evaluation (NURE) program in the U.S. Gulf Coast Uranium Province. NURE estimated 87 million pounds of reserves in themore » $$30/lb U{sub 3}O{sub 8} cost category in the Coast Plain uranium resource region, most in the Gulf Coast Uranium Province. Since NURE, 40 million pounds of reserves have been mined, and 38 million pounds are estimated to remain in place as of 2012, accounting for all but 9 million pounds of U{sub 3}O{sub 8} in the reserve or production categories in the NURE estimate. Considering the complexities and uncertainties of the analysis, this study indicates that the NURE reserve estimates for the province were accurate. An unconditional potential resource of 1.4 billion pounds of U{sub 3}O{sub 8}, 600 million pounds of U{sub 3}O{sub 8} in the forward cost category of $$30/lb U{sub 3}O{sub 8} (1980 prices), was estimated in 106 favorable areas by the NURE program in the province. Removing potential resources from the non-productive Houston embayment, and those reserves estimated below historic and current mining depths reduces the unconditional potential resource 33% to about 930 million pounds of U{sub 3}O{sub 8}, and that in the $30/lb cost category 34% to 399 million pounds of U{sub 3}O{sub 8}. Based on production records and reserve estimates tabulated for the region, most of the production since 1980 is likely from the reserves identified by NURE. The potential resource predicted by NURE has not been developed, likely due to a variety of factors related to the low uranium prices that have prevailed since 1980.« less

Critical analysis of world uranium resources

USGS Publications Warehouse

Hall, Susan; Coleman, Margaret

2013-01-01

The U.S. Department of Energy, Energy Information Administration (EIA) joined with the U.S. Department of the Interior, U.S. Geological Survey (USGS) to analyze the world uranium supply and demand balance. To evaluate short-term primary supply (0–15 years), the analysis focused on Reasonably Assured Resources (RAR), which are resources projected with a high degree of geologic assurance and considered to be economically feasible to mine. Such resources include uranium resources from mines currently in production as well as resources that are in the stages of feasibility or of being permitted. Sources of secondary supply for uranium, such as stockpiles and reprocessed fuel, were also examined. To evaluate long-term primary supply, estimates of uranium from unconventional and from undiscovered resources were analyzed. At 2010 rates of consumption, uranium resources identified in operating or developing mines would fuel the world nuclear fleet for about 30 years. However, projections currently predict an increase in uranium requirements tied to expansion of nuclear energy worldwide. Under a low-demand scenario, requirements through the period ending in 2035 are about 2.1 million tU. In the low demand case, uranium identified in existing and developing mines is adequate to supply requirements. However, whether or not these identified resources will be developed rapidly enough to provide an uninterrupted fuel supply to expanded nuclear facilities could not be determined. On the basis of a scenario of high demand through 2035, 2.6 million tU is required and identified resources in operating or developing mines is inadequate. Beyond 2035, when requirements could exceed resources in these developing properties, other sources will need to be developed from less well-assured resources, deposits not yet at the prefeasibility stage, resources that are currently subeconomic, secondary sources, undiscovered conventional resources, and unconventional uranium supplies. This report’s analysis of 141 mines that are operating or are being actively developed identifies 2.7 million tU of in-situ uranium resources worldwide, approximately 2.1 million tU recoverable after mining and milling losses were deducted. Sixty-four operating mines report a total of 1.4 million tU of in-situ RAR (about 1 million tU recoverable). Seventy-seven developing mines/production centers report 1.3 million tU in-situ Reasonably Assured Resources (RAR) (about 1.1 million tU recoverable), which have a reasonable chance of producing uranium within 5 years. Most of the production is projected to come from conventional underground or open pit mines as opposed to in-situ leach mines. Production capacity in operating mines is about 76,000 tU/yr, and in developing mines is estimated at greater than 52,000 tU/yr. Production capacity in operating mines should be considered a maximum as mines seldom produce up to licensed capacity due to operational difficulties. In 2010, worldwide mines operated at 70 percent of licensed capacity, and production has never exceeded 89 percent of capacity. The capacity in developing mines is not always reported. In this study 35 percent of developing mines did not report a target licensed capacity, so estimates of future capacity may be too low. The Organisation for Economic Co-operation and Development’s Nuclear Energy Agency (NEA) and International Atomic Energy Agency (IAEA) estimate an additional 1.4 million tU economically recoverable resources, beyond that identified in operating or developing mines identified in this report. As well, 0.5 million tU in subeconomic resources, and 2.3 million tU in the geologically less certain inferred category are identified worldwide. These agencies estimate 2.2 million tU in secondary sources such as government and commercial stockpiles and re-enriched uranium tails. They also estimate that unconventional uranium supplies (uraniferous phosphate and black shale deposits) may contain up to 7.6 million tU. Although unconventional resources are currently subeconomic, the improvement of extraction techniques or the production of coproducts may make extraction of uranium from these types of deposits profitable. A large undiscovered resource base is reported by these agencies, however this class of resource should be considered speculative and will require intensive exploration programs to adequately define them as mineable. These resources may all contribute to uranium supply that would fuel the world nuclear fleet well beyond that calculated in this report. Production of resources in both operating and developing uranium mines is subject to uncertainties caused by technical, legal, regulatory, and financial challenges that combined to create long timelines between deposit discovery and mine production. This analysis indicates that mine development is proceeding too slowly to fully meet requirements for an expanded nuclear power reactor fleet in the near future (to 2035), and unless adequate secondary or unconventional resources can be identified, imbalances in supply and demand may occur.

PRODUCTION OF URANIUM AND THORIUM COMPOUNDS

DOEpatents

Arden, T.V.; Burstall, F.H.; Linstead, R.P.; Wells, R.A.

1955-12-27

Compounds of Th and U are extracted with an organic solvent in the presence of an adsorbent substance which has greater retentivity for impurities present than for the uranium and/or thorium. The preferred adsorbent material is noted as being cellulose. The uranium and thoriumcontaining substances treated are preferably in the form of dissolved nitrates, and the preferred organic solvent is diethyl ether.

PROCESS FOR CONTINUOUSLY SEPARATING IRRADIATION PRODUCTS OF THORIUM

DOEpatents

Hatch, L.P.; Miles, F.T.; Sheehan, T.V.; Wiswall, R.H.; Heus, R.J.

1959-07-01

A method is presented for separating uranium-233 and protactinium from thorium-232 containing compositions which comprises irradiating finely divided particles of said thorium with a neutron flux to form uranium-233 and protactinium, heating the neutron-irradiated composition in a fluorine and hydrogen atmosphere to form volatile fluorides of uranium and protactinium and thereafter separating said volatile fluorides from the thorium.

QUANTITATIVE DETERMINATION OF THE URANIUM CONTENT OF URANIUM ORES TECHNOLOGICAL PRODUCTS BY ION EXCHANGE-COMPLEXON SEPARATION (in Hungarian)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fodor, M.

An ion exchange-complexion separation meihod was developed for the removal of interfering elements in the determination of the uranium content of recovery solutions. By adding (ethylenediamine)tetraacetic acid to the solution, most of the interfering elements can be brought into an anionic complex. Adjusting the soluiion to pH 7 and letting it pass through an Amberlite IRC-50 type cation exchanger of hydrogen form, the uranium remains on the column whereas the interfering elements pass into the effluent. The method was successfully applied in analyzing the recovery solutions of uranium ores. (auth)

Separation by solvent extraction

DOEpatents

Holt, Jr., Charles H.

1976-04-06

17. A process for separating fission product values from uranium and plutonium values contained in an aqueous solution, comprising adding an oxidizing agent to said solution to secure uranium and plutonium in their hexavalent state; contacting said aqueous solution with a substantially water-immiscible organic solvent while agitating and maintaining the temperature at from -1.degree. to -2.degree. C. until the major part of the water present is frozen; continuously separating a solid ice phase as it is formed; separating a remaining aqueous liquid phase containing fission product values and a solvent phase containing plutonium and uranium values from each other; melting at least the last obtained part of said ice phase and adding it to said separated liquid phase; and treating the resulting liquid with a new supply of solvent whereby it is practically depleted of uranium and plutonium.

PROCESS FOR THE SEPARATION OF HEAVY METALS

DOEpatents

Gofman, J.W.; Connick, R.E.; Wahl, A.C.

1959-01-27

A method is presented for thc separation of plutonium from uranium and the fission products with which it is associated. The method is based on the fact that hexavalent plutonium forms an insoluble complex precipitate with sodium acetate, as does the uranyl ion, while reduced plutonium is not precipitated by sodium acetate. Several embodiments are shown, e.g., a solution containing plutonium and uranium in the hexavalent state may be contacted with sodium acetate causing the formation of a sodium uranyl acetate precipitate which carries the plutonium values while the fission products remain in solution. If the original solution is treated with a reducing agent, so that the plutonium is reduced while the uranium remains in the hexavalent state, and sodium and acetate ions are added, the uranium will precipitutc while the plutonium remains in solution effecting separation of the Pu from urarium.

Fermentation and Hydrogen Metabolism Affect Uranium Reduction by Clostridia

DOE PAGES

Gao, Weimin; Francis, Arokiasamy J.

2013-01-01

Previously, it has been shown that not only is uranium reduction under fermentation condition common among clostridia species, but also the strains differed in the extent of their capability and the pH of the culture significantly affected uranium(VI) reduction. In this study, using HPLC and GC techniques, metabolic properties of those clostridial strains active in uranium reduction under fermentation conditions have been characterized and their effects on capability variance of uranium reduction discussed. Then, the relationship between hydrogen metabolism and uranium reduction has been further explored and the important role played by hydrogenase in uranium(VI) and iron(III) reduction by clostridiamore » demonstrated. When hydrogen was provided as the headspace gas, uranium(VI) reduction occurred in the presence of whole cells of clostridia. This is in contrast to that of nitrogen as the headspace gas. Without clostridia cells, hydrogen alone could not result in uranium(VI) reduction. In alignment with this observation, it was also found that either copper(II) addition or iron depletion in the medium could compromise uranium reduction by clostridia. In the end, a comprehensive model was proposed to explain uranium reduction by clostridia and its relationship to the overall metabolism especially hydrogen (H 2 ) production.« less

Method for fluorination of actinide fluorides and oxyfluorides thereof using O.sub.2 F.sub.2

DOEpatents

Eller, Phillip G.; Malm, John G.; Penneman, Robert A.

1988-01-01

Method for fluorination of actinides and fluorides and oxyfluorides thereof using O.sub.2 F.sub.2 which generates actinide hexafluorides, and for removal of actinides and compounds thereof from surfaces upon which they appear as unwanted deposits. The fluorinating agent, O.sub.2 F.sub.2, has been observed to readily perform the above-described tasks at sufficiently low temperatures that there is virtually no damage to the containment vessels. Moreover, the resulting actinide hexafluorides are thereby not destroyed by high temperature reactions with the walls of the reaction vessel. Dioxygen difluoride is easily prepared, stored and transferred to the desired place of reaction.

Method for recovery of actinides from actinide-bearing scrap and waste nuclear material using O/sub 2/F/sub 2/

DOEpatents

Asprey, L.B.; Eller, P.G.

1984-09-12

Method for recovery of actinides from nuclear waste material containing sintered and other oxides thereof and from scrap materials containing the metal actinides using O/sub 2/F/sub 2/ to generate the hexafluorides of the actinides present therein. The fluorinating agent, O/sub 2/F/sub 2/, has been observed to perform the above-described tasks at sufficiently low temperatures that there is virtually no damage to the containment vessels. Moreover, the resulting actinide hexafluorides are not detroyed by high temperature reactions with the walls of the reaction vessel. Dioxygen difluoride is readily prepared, stored and transferred to the place of reaction.

Method for fluorination of actinide fluorides and oxyfluorides thereof using O[sub 2]F[sub 2

DOEpatents

Eller, P.G.; Malm, J.G.; Penneman, R.A.

1988-11-08

Method is described for fluorination of actinides and fluorides and oxyfluorides thereof using O[sub 2]F[sub 2] which generates actinide hexafluorides, and for removal of actinides and compounds thereof from surfaces upon which they appear as unwanted deposits. The fluorinating agent, O[sub 2]F[sub 2], has been observed to readily perform the above-described tasks at sufficiently low temperatures that there is virtually no damage to the containment vessels. Moreover, the resulting actinide hexafluorides are thereby not destroyed by high temperature reactions with the walls of the reaction vessel. Dioxygen difluoride is easily prepared, stored and transferred to the desired place of reaction.

SEPARATION OF PLUTONIUM FROM URANIUM AND FISSION PRODUCTS BY ADSORPTION

DOEpatents

Seaborg, G.T.; Willard, J.E.

1958-01-01

A method is presented for the separation of plutonium from solutions containing that element in a valence state not higher than 41 together with uranium ions and fission products. This separation is accomplished by contacting the solutions with diatomaceous earth which preferentially adsorbs the plutonium present. Also mentioned as effective for this adsorbtive separation are silica gel, filler's earth and alumina.

Kr ion irradiation study of the depleted-uranium alloys

NASA Astrophysics Data System (ADS)

Gan, J.; Keiser, D. D.; Miller, B. D.; Kirk, M. A.; Rest, J.; Allen, T. R.; Wachs, D. M.

2010-12-01

Fuel development for the reduced enrichment research and test reactor (RERTR) program is tasked with the development of new low enrichment uranium nuclear fuels that can be employed to replace existing high enrichment uranium fuels currently used in some research reactors throughout the world. For dispersion type fuels, radiation stability of the fuel-cladding interaction product has a strong impact on fuel performance. Three depleted-uranium alloys are cast for the radiation stability studies of the fuel-cladding interaction product using Kr ion irradiation to investigate radiation damage from fission products. SEM analysis indicates the presence of the phases of interest: U(Al, Si) 3, (U, Mo)(Al, Si) 3, UMo 2Al 20, U 6Mo 4Al 43 and UAl 4. Irradiations of TEM disc samples were conducted with 500 keV Kr ions at 200 °C to ion doses up to 2.5 × 10 19 ions/m 2 (˜10 dpa) with an Kr ion flux of 10 16 ions/m 2/s (˜4.0 × 10 -3 dpa/s). Microstructural evolution of the phases relevant to fuel-cladding interaction products was investigated using transmission electron microscopy.

Solubility classification of airborne uranium products collected at the perimeter of the Allied Chemical Plant, Metropolis, Illinois

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kalkwarf, D.R.

1980-05-01

Airborne uranium products were collected at the perimeter of the uranium-conversion plant operated by the Allied Chemical Corporation at Metropolis, Illinois, and the dissolution rates of these products were classified in terms of the ICRP Task Group Lung Model. Assignments were based on measurements of the dissolution half-times exhibited by uranium components of the dust samples as they dissolved in simulated lung fluid at 37/sup 0/C. Based on three trials, the dissolution behavior of dust with aerodynamic equivalent diameter (AED) less than 5.5 ..mu..m and collected nearest the closest residence to the plant was classified 0.40 D, 0.60 Y. Basedmore » on two trials, the dissolution behavior of dust with AED greater than 5.5 ..mu..m and collected at this location was classified 0.37 D, 0.63 Y. Based on one trial, the dissolution behavior of dust with AED less than 5.5 ..mu..m and collected at a location on the opposite side of the plant was classified 0.68 D, 0.32 Y. There was some evidence for adsorption of dissolved uranium onto other dust components during dissolution, and preliminary dissolution trials are recommended for future samples in order to optimize the fluid replacement schedule.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dale, Gregory E.

There is currently a serious shortage of 99Mo, from which to generate the medically significant isotope 99mTc. Most of the world's supply comes from the fission of highly enriched uranium targets--this is a proliferation concern. This document focuses on the technology involved in two alternative methods: electron accelerator production of 99Mo from the 100Mo(γ,n) 99Mo reaction and production of 99Mo as a fission product in a subcritical, DT accelerator-driven low enriched uranium salt solution.

Non-Invasive Acoustic-Based Monitoring of Heavy Water and Uranium Process Solutions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pantea, Cristian; Sinha, Dipen N.; Lakis, Rollin Evan

This presentation includes slides on Project Goals; Heavy Water Production Monitoring: A New Challenge for the IAEA; Noninvasive Measurements in SFAI Cell; Large Scatter in Literature Values; Large Scatter in Literature Values; Highest Precision Sound Speed Data Available: New Standard in H/D; ~400 pts of data; Noninvasive Measurements in SFAI Cell; New funding from NA241 SGTech; Uranium Solution Monitoring: Inspired by IAEA Challenge in Kazakhstan; Non-Invasive Acoustic-Based Monitoring of Uranium in Solutions; Non-Invasive Acoustic-Based Monitoring of Uranium in Solutions; and finally a summary.

Electrolytic process for preparing uranium metal

DOEpatents

Haas, Paul A.

1990-01-01

An electrolytic process for making uranium from uranium oxide using Cl.sub.2 anode product from an electrolytic cell to react with UO.sub.2 to form uranium chlorides. The chlorides are used in low concentrations in a melt comprising fluorides and chlorides of potassium, sodium and barium in the electrolytic cell. The electrolysis produces Cl.sub.2 at the anode that reacts with UO.sub.2 in the feed reactor to form soluble UCl.sub.4, available for a continuous process in the electrolytic cell, rather than having insoluble UO.sub.2 fouling the cell.

Final Technical Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sobecky, Patricia A; Taillefert, Martial

This final technical report describes results and findings from a research project to examine the role of microbial phosphohydrolase enzymes in naturally occurring subsurface microorganisms for the purpose of promoting the immobilization of the radionuclide uranium through the production of insoluble uranium phosphate minerals. The research project investigated the microbial mechanisms and the physical and chemical processes promoting uranium biomineralization and sequestration in oxygenated subsurface soils. Uranium biomineralization under aerobic conditions can provide a secondary biobarrier strategy to immobilize radionuclides should the metal precipitates formed by microbial dissimilatory mechanisms remobilize due to a change in redox state.

Characterization of the Environmentally Induced Chemical Transformations of Uranium Tetrafluoride

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wellons, M.

A key challenge with nuclear safeguards environmental sampling is identification of the materials post release due to subsequent chemical reactions with ambient water and oxygen. Uranium Tetrafluoride (UF4) is of interest as an intermediate in both the upstream and downstream portions of uranium feedstock and metal production processes used in nuclear fuel production; however minimal published research exists relating to UF4 hydrolysis. FY16 efforts were dedicated to in-situ Raman spectroscopy and X-ray diffraction characterization of UF4 during exposure to various relative humidity conditions. This effort mapped several hydrolysis reaction pathways and identified both intermediate, and terminal progeny species.

An Overview of Process Monitoring Related to the Production of Uranium Ore Concentrate

DOE Office of Scientific and Technical Information (OSTI.GOV)

McGinnis, Brent

2014-04-01

Uranium ore concentrate (UOC) in various chemical forms, is a high-value commodity in the commercial nuclear market, is a potential target for illicit acquisition, by both State and non-State actors. With the global expansion of uranium production capacity, control of UOC is emerging as a potentially weak link in the nuclear supply chain. Its protection, control and management thus pose a key challenge for the international community, including States, regulatory authorities and industry. This report evaluates current process monitoring practice and makes recommendations for utilization of existing or new techniques for managing the inventory and tracking this material.

SEPARATION OF PLUTONIUM FROM FISSION PRODUCTS BY A COLLOID REMOVAL PROCESS

DOEpatents

Schubert, J.

1960-05-24

A method is given for separating plutonium from uranium fission products. An acidic aqueous solution containing plutonium and uranium fission products is subjected to a process for separating ionic values from colloidal matter suspended therein while the pH of the solution is maintained between 0 and 4. Certain of the fission products, and in particular, zirconium, niobium, lanthanum, and barium are in a colloidal state within this pH range, while plutonium remains in an ionic form, Dialysis, ultracontrifugation, and ultrafiltration are suitable methods of separating plutonium ions from the colloids.

The measurement of U(VI) and Np(IV) mass transfer in a single stage centrifugal contactor

NASA Astrophysics Data System (ADS)

May, I.; Birkett, E. J.; Denniss, I. S.; Gaubert, E. T.; Jobson, M.

2000-07-01

BNFL currently operates two reprocessing plants for the conversion of spent nuclear fuel into uranium and plutonium products for fabrication into uranium oxide and mixed uranium and plutonium oxide (MOX) fuels. To safeguard the future commercial viability of this process, BNFL is developing novel single cycle flowsheets that can be operated in conjunction with intensified centrifugal contactors.

Ionic Liquids as templating agents in formation of uranium-containing nanomaterials

DOEpatents

Visser, Ann E; Bridges, Nicholas J

2014-06-10

A method for forming nanoparticles containing uranium oxide is described. The method includes combining a uranium-containing feedstock with an ionic liquid to form a mixture and holding the mixture at an elevated temperature for a period of time to form the product nanoparticles. The method can be carried out at low temperatures, for instance less than about 300.degree. C.

TRANSURANIC ELEMENT, COMPOSITION THEREOF, AND METHODS FOR PRODUCING SEPARATING AND PURIFYING SAME

DOEpatents

Wahl, A.C.

1961-09-19

A process of separating plutonium from fission products contained in an aqueous solution is described. Plutonium, in the tri- or tetravalent state, and the fission products are coprecipitated on lanthanum fluoride, lanthanum oxalate, cerous fluoride, cerous phosphate, ceric iodate, zirconyl phosphate, thorium iodate, or thorium fluoride. The precipitate is dissolved in acid, and the plutonium is oxidized to the hexavalent state. The fission products are selectively precipitated on a carrier of the above group but different from that used for the coprecipitation. The plutonium in the solution, after removal of the fission product precipitate, is reduced to at least the tetravalent state and precipitated on lanthanum fluoride, lanthanum phosphate, lanthanum oxalate, lanthanum hydroxide, cerous fluoride, cerous phosphate, cerous oxalate, cerous hydroxide, ceric iodate, zirconyl phosphate, zirconyl iodate, zirconium hydroxide, thorium fluoride, thorium oxalate, thorium iodate, thorium peroxide, uranium iodate, uranium oxalate, or uranium peroxide, again using a different carrier than that used for the precipitation of the fission products.

URANIUM RECOVERY FROM COMPOSITE UF$sub 4$ REDUCTION BOMB WASTES

DOE Office of Scientific and Technical Information (OSTI.GOV)

Johnson, E R; Doyle, R L; Coleman, J R

1954-01-28

A number of techniques have been investigated on a laboratory-scale for separating uranium from fluorides during the recovery of uranium from UF4 reduction bomb wastes (C-oxide) by an HCl leach - NH4OH precipitation process. Among these are included adsorption of fluorides from filtered leach liquors, fractional precipitation of fluorides and uranium, complexing of fluorides into forms soluble in slightly acid solutions, and fluoride volatilization from the uranium concentrate. Solubility studies of CaF2 and MgF2 in aqueous hydrochloric acid at various acidities and temperatures were also conducted. A description of the production-scale processing of C-oxide in the FMPC scrap plant hasmore » been included.« less

Global Uranium And Thorium Resources: Are They Adequate To Satisfy Demand Over The Next Half Century?

NASA Astrophysics Data System (ADS)

Lambert, I. B.

2012-04-01

This presentation will consider the adequacy of global uranium and thorium resources to meet realistic nuclear power demand scenarios over the next half century. It is presented on behalf of, and based on evaluations by, the Uranium Group - a joint initiative of the OECD Nuclear Energy Agency and the International Atomic Energy Agency, of which the author is a Vice Chair. The Uranium Group produces a biennial report on Uranium Resources, Production and Demand based on information from some 40 countries involved in the nuclear fuel cycle, which also briefly reviews thorium resources. Uranium: In 2008, world production of uranium amounted to almost 44,000 tonnes (tU). This supplied approximately three-quarters of world reactor requirements (approx. 59,000 tU), the remainder being met by previously mined uranium (so-called secondary sources). Information on availability of secondary sources - which include uranium from excess inventories, dismantling nuclear warheads, tails and spent fuel reprocessing - is incomplete, but such sources are expected to decrease in market importance after 2013. In 2008, the total world Reasonably Assured plus Inferred Resources of uranium (recoverable at less than 130/kgU) amounted to 5.4 million tonnes. In addition, it is clear that there are vast amounts of uranium recoverable at higher costs in known deposits, plus many as yet undiscovered deposits. The Uranium Group has concluded that the uranium resource base is more than adequate to meet projected high-case requirements for nuclear power for at least half a century. This conclusion does not assume increasing replacement of uranium by fuels from reprocessing current reactor wastes, or by thorium, nor greater reactor efficiencies, which are likely to ameliorate future uranium demand. However, progressively increasing quantities of uranium will need to be mined, against a backdrop of the relatively small number of producing facilities around the world, geopolitical uncertainties and strong opposition to growth of nuclear power in a number of quarters - it is vital that the market provides incentives for exploration and development of environmentally sustainable mining operations. Thorium: World Reasonably Assured plus Inferred Resources of thorium are estimated at over 2.2 million tonnes, in hard rock and heavy mineral sand deposits. At least double this amount is considered to occur in as yet undiscovered thorium deposits. Currently, demand for thorium is insignificant, but even a major shift to thorium-fueled reactors would not make significant inroads into the huge resource base over the next half century.

EVALUATION OF AUSTRALIAN RUM JUNGLE URANIUM CONCENTRATE FOR USE AS NLO REFINERY FEED

DOE Office of Scientific and Technical Information (OSTI.GOV)

Collopy, T.J.; Huntington, C.W.; Blum, J.F.

1956-01-20

A laboratory evaluation of Australian Rum Jungle uranium concentrate showed that the uracium can be satisfactorily extracted by 33.5% TBP-kerosene from an aqueous acid slurry of the material, and that impurities in the aqueous uranyl nitrate product obtained by re-extraetion from the organic phase approach NL0 tolerance specifications. The uranium values in the organic product were not completely re-extracted at room temperatare (l0th stage organic, 1.6 g/l U); however, it was assumed that reextraction will be complete under pulse column conditions (150 deg F). The results of the Pilot Plant evaluation of Rum Jungle uranium concentrate (Lot No. 1) indicatedmore » that this material can be processed employing NLO refinery conditions. The aqueous uranyl nitrate product from the test met all impurity specifications except those for manganese and nickel. The high chloride content of this lot of concentrate will mske blending necessary in order to meet NLO feed material specifications. The blending will alan lessen the tendencies toward metallic contamination of the OK liquor observed in these tests. (auth)« less

Biogeochemical behaviour and bioremediation of uranium in waters of abandoned mines.

PubMed

Mkandawire, Martin

2013-11-01

The discharges of uranium and associated radionuclides as well as heavy metals and metalloids from waste and tailing dumps in abandoned uranium mining and processing sites pose contamination risks to surface and groundwater. Although many more are being planned for nuclear energy purposes, most of the abandoned uranium mines are a legacy of uranium production that fuelled arms race during the cold war of the last century. Since the end of cold war, there have been efforts to rehabilitate the mining sites, initially, using classical remediation techniques based on high chemical and civil engineering. Recently, bioremediation technology has been sought as alternatives to the classical approach due to reasons, which include: (a) high demand of sites requiring remediation; (b) the economic implication of running and maintaining the facilities due to high energy and work force demand; and (c) the pattern and characteristics of contaminant discharges in most of the former uranium mining and processing sites prevents the use of classical methods. This review discusses risks of uranium contamination from abandoned uranium mines from the biogeochemical point of view and the potential and limitation of uranium bioremediation technique as alternative to classical approach in abandoned uranium mining and processing sites.

Validation of reference materials for uranium radiochronometry in the frame of nuclear forensic investigations

DOE PAGES

Varga, Z.; Mayer, K.; Bonamici, C. E.; ...

2015-05-11

The results of a joint effort by expert nuclear forensic laboratories in the area of age dating of uranium, i.e. the elapsed time since the last chemical purification of the material are presented and discussed. Completely separated uranium materials of known production date were distributed among the laboratories, and the samples were dated according to routine laboratory procedures by the measurement of the ²²⁰Th/²³⁴U ratio. The measurement results were in good agreement with the known production date showing that the concept for preparing uranium age dating reference material based on complete separation is valid. Detailed knowledge of the laboratory proceduresmore » used for uranium age dating allows the identification of possible improvements in the current protocols and the development of improved practice in the future. The availability of age dating reference materials as well as the evolvement of the age dating best-practice protocol will increase the relevance and applicability of age dating as part of the tool-kit available for nuclear forensic investigations.« less

Validation of reference materials for uranium radiochronometry in the frame of nuclear forensic investigations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Varga, Z.; Mayer, K.; Bonamici, C. E.

The results of a joint effort by expert nuclear forensic laboratories in the area of age dating of uranium, i.e. the elapsed time since the last chemical purification of the material are presented and discussed. Completely separated uranium materials of known production date were distributed among the laboratories, and the samples were dated according to routine laboratory procedures by the measurement of the ²²⁰Th/²³⁴U ratio. The measurement results were in good agreement with the known production date showing that the concept for preparing uranium age dating reference material based on complete separation is valid. Detailed knowledge of the laboratory proceduresmore » used for uranium age dating allows the identification of possible improvements in the current protocols and the development of improved practice in the future. The availability of age dating reference materials as well as the evolvement of the age dating best-practice protocol will increase the relevance and applicability of age dating as part of the tool-kit available for nuclear forensic investigations.« less

Production and Evaluation of 236gNp and Reference Materials for Naturally Occurring Radioactive Materials

NASA Astrophysics Data System (ADS)

Larijani, Cyrus Kouroush

This thesis is based on the development of a radiochemical separation scheme capable of separating both 236gNp and 236Pu from a uranium target of natural isotopic composition ( 1 g uranium) and 200 MBq of fission decay products. The isobaric distribution of fission residues produced following the bombardment of a natural uranium target with a beam of 25 MeV protons has been evaluated. Decay analysis of thirteen isobarically distinct fission residues were carried out using high-resolution gamma-ray spectrometry at the UK National Physical Laboratory. Stoichiometric abundances were calculated via the determination of absolute activity concentrations associated with the longest-lived members of each isobaric chain. This technique was validated by computational modelling of likely sequential decay processes through an isobaric decay chain. The results were largely in agreement with previously published values for neutron bombardments on natural uranium at energies of 14 MeV. Higher relative yields of products with mass numbers A 110-130 were found, consistent with the increasing yield of these radionuclides as the bombarding energy is increased. Using literature values for the production cross-section for fusion of protons with uranium targets, it is estimated that an upper limit of approximately 250 Bq of activity from the 236Np ground state was produced in this experiment. Using a radiochemical separation scheme, Np and Pu fractions were separated from the produced fission decay products, with analyses of the target-based final reaction products made using Inductively Couple Plasma Mass Spectrometry (ICP-MS) and high-resolution alpha and gamma-ray spectrometry. In a separate research theme, reliable measurement of Naturally Occurring Radioactive Materials is of significance in order to comply with environmental regulations and for radiological protection purposes. The thesis describes the standardisation of three reference materials, namely Sand, Tuff and TiO2 which can serve as quality control materials to achieve traceability, method validation and instrument calibration. The sample preparation, material characterization via gamma, alpha and Inductively Coupled Plasma Mass Spectrometry (ICP-MS) and the assignment of values for both the 4n Thorium and 4n + 2 Uranium decay series are presented.

Processing of LEU targets for {sup 99}Mo production--testing and modification of the Cintichem process

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wu, D.; Landsberger, S.; Buchholz, B.

1995-09-01

Recent experimental results on testing and modification of the Cintichem process to allow substitution of low enriched uranium (LEU) for high enriched uranium (HEU) targets are presented in this report. The main focus is on {sup 99}Mo recovery and purification by its precipitation with {alpha}-benzoin oxime. Parameters that were studied include concentrations of nitric and sulfuric acids, partial neutralization of the acids, molybdenum and uranium concentrations, and the ratio of {alpha}-benzoin oxime to molybdenum. Decontamination factors for uranium, neptunium, and various fission products were measured. Experiments with tracer levels of irradiated LEU were conducted for testing the {sup 99}Mo recoverymore » and purification during each step of the Cintichem process. Improving the process with additional processing steps was also attempted. The results indicate that the conversion of molybdenum chemical processing from HEU to LEU targets is possible.« less

An off-line method to characterize the fission product release from uranium carbide-target prototypes developed for SPIRAL2 project

NASA Astrophysics Data System (ADS)

Hy, B.; Barré-Boscher, N.; Özgümüs, A.; Roussière, B.; Tusseau-Nenez, S.; Lau, C.; Cheikh Mhamed, M.; Raynaud, M.; Said, A.; Kolos, K.; Cottereau, E.; Essabaa, S.; Tougait, O.; Pasturel, M.

2012-10-01

In the context of radioactive ion beams, fission targets, often based on uranium compounds, have been used for more than 50 years at isotope separator on line facilities. The development of several projects of second generation facilities aiming at intensities two or three orders of magnitude higher than today puts an emphasis on the properties of the uranium fission targets. A study, driven by Institut de Physique Nucléaire d'Orsay (IPNO), has been started within the SPIRAL2 project to try and fully understand the behavior of these targets. In this paper, we have focused on five uranium carbide based targets. We present an off-line method to characterize their fission product release and the results are examined in conjunction with physical characteristics of each material such as the microstructure, the porosity and the chemical composition.

Design and analysis of a natural-gradient ground-water tracer test in a freshwater tidal wetland, West Branch Canal Creek, Aberdeen Proving Ground, Maryland

USGS Publications Warehouse

Olsen, Lisa D.; Tenbus, Frederick J.

2005-01-01

A natural-gradient ground-water tracer test was designed and conducted in a tidal freshwater wetland at West Branch Canal Creek, Aberdeen Proving Ground, Maryland. The objectives of the test were to characterize solute transport at the site, obtain data to more accurately determine the ground-water velocity in the upper wetland sediments, and to compare a conservative, ionic tracer (bromide) to a volatile tracer (sulfur hexafluoride) to ascertain whether volatilization could be an important process in attenuating volatile organic compounds in the ground water. The tracer test was conducted within the upper peat unit of a layer of wetland sediments that also includes a lower clayey unit; the combined layer overlies an aquifer. The area selected for the test was thought to have an above-average rate of ground-water discharge based on ground-water head distributions and near-surface detections of volatile organic compounds measured in previous studies. Because ground-water velocities in the wetland sediments were expected to be slow compared to the underlying aquifer, the test was designed to be conducted on a small scale. Ninety-seven ?-inch-diameter inverted-screen stainless-steel piezometers were installed in a cylindrical array within approximately 25 cubic feet (2.3 cubic meters) of wetland sediments, in an area with a vertically upward hydraulic gradient. Fluorescein dye was used to qualitatively evaluate the hydrologic integrity of the tracer array before the start of the tracer test, including verifying the absence of hydraulic short-circuiting due to nonnatural vertical conduits potentially created during piezometer installation. Bromide and sulfur hexafluoride tracers (0.139 liter of solution containing 100,000 milligrams per liter of bromide ion and 23.3 milligrams per liter of sulfur hexafluoride) were co-injected and monitored to generate a dataset that could be used to evaluate solute transport in three dimensions. Piezometers were sampled 2 to 15 times each, from July 1998 through September 1999, to assess background conditions and monitor tracer movement. During the test, 644 samples were analyzed for fluorescein, 617 samples were analyzed for bromide with an ion-selective electrode, 213 samples were analyzed for bromide with colorimetric methods, and 603 samples were analyzed for sulfur hexafluoride, including samples collected prior to tracer injection to determine background concentrations. Additional samples were analyzed for volatile organic compounds (96 samples) and methane (37 samples) to determine the distribution of these contaminants and the extent of methanogenic conditions within the tracer array; however, these data were not used for the analysis of the test. During the tracer test, the fluorescein dye, bromide, and sulfur hexafluoride were transported predominantly in the upward direction, although all three tracers also moved outward in all directions from the injection point, and it is likely that some tracer mass moved beyond the lateral edges of the array. An analysis of the tracer-test data was performed through the use of breakthrough curves and isoconcentration contour plots. Results show that movement of the fluorescein dye, a non-conservative tracer, was retarded compared to the other two tracers, likely as a result of sorption onto the wetland sediments. Suspected loss of tracer mass along the lateral edges of the array prevented a straightforward quantitative analysis of tracer transport and ground-water velocity from the bromide and sulfur-hexafluoride data. In addition, the initial density of the bromide/sulfur hexafluoride solution (calculated to be 1.097 grams per milli2 Ground-Water Tracer Test, West Branch Canal Creek, Aberdeen Proving Ground, MD liter) could have caused the solution to sink below the injection point before undergoing dilution and moving back up into the array. For these reasons, the data analysis in this report was performed largely through qualitative method

PROCESSES OF CHLORINATION OF URANIUM OXIDES

DOEpatents

Rosenfeld, S.

1958-09-16

An improvement is described in the process fur making UCl/sub 4/ from uranium oxide and carbon tetrachloride. In that process, oxides of uranium are contacted with carbon tetrachloride vapor at an elevated temperature. It has been fuund that the reaction product and yield are improved if the uranlum oxide charge is disposed in flat trays in the reaction zone, to a depth of not more than 1/2 centimeter.

10. VIEW OF DEPLETED URANIUM INGOT AND MOLD IN FOUNDRY. ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

10. VIEW OF DEPLETED URANIUM INGOT AND MOLD IN FOUNDRY. (11/11/56) - Rocky Flats Plant, Non-Nuclear Production Facility, South of Cottonwood Avenue, west of Seventh Avenue & east of Building 460, Golden, Jefferson County, CO

Uranium(VI) interactions with mackinawite in the presence and absence of bicarbonate and oxygen.

PubMed

Gallegos, Tanya J; Fuller, Christopher C; Webb, Samuel M; Betterton, William

2013-07-02

Mackinawite, Fe(II)S, samples loaded with uranium (10(-5), 10(-4), and 10(-3) mol U/g FeS) at pH 5, 7, and 9, were characterized using X-ray absorption spectroscopy and X-ray diffraction to determine the effects of pH, bicarbonate, and oxidation on uptake. Under anoxic conditions, a 5 g/L suspension of mackinawite lowered 5 × 10(-5) M uranium(VI) to below 30 ppb (1.26 × 10(-7) M) U. Between 82 and 88% of the uranium removed from solution by mackinawite was U(IV) and was nearly completely reduced to U(IV) when 0.012 M bicarbonate was added. Near-neighbor coordination consisting of uranium-oxygen and uranium-uranium distances indicates the formation of uraninite in the presence and absence of bicarbonate, suggesting reductive precipitation as the dominant removal mechanism. Following equilibration in air, mackinawite was oxidized to mainly goethite and sulfur and about 76% of U(IV) was reoxidized to U(VI) with coordination of uranium to axial and equatorial oxygen, similar to uranyl. Additionally, uranium-iron distances, typical of coprecipitation of uranium with iron oxides, and uranium-sulfur distances indicating bidentate coordination of U(VI) to sulfate were evident. The affinity of mackinawite and its oxidation products for U(VI) provides impetus for further study of mackinawite as a potential reactive medium for remediation of uranium-contaminated water.

Fission Product Yields from {sup 232}Th, {sup 238}U, and {sup 235}U Using 14 MeV Neutrons

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pierson, B.D., E-mail: bpnuke@umich.edu; Pacific Northwest National Laboratory, P.O. Box 999, Richland, WA 99352; Greenwood, L.R.

Neutron-induced fission yield studies using deuterium-tritium fusion-produced 14 MeV neutrons have not yet directly measured fission yields from fission products with half-lives on the order of seconds (far from the line of nuclear stability). Fundamental data of this nature are important for improving and validating the current models of the nuclear fission process. Cyclic neutron activation analysis (CNAA) was performed on three actinide targets–thorium-oxide, depleted uranium metal, and highly enriched uranium metal–at the University of Michigan's Neutron Science Laboratory (UM-NSL) using a pneumatic system and Thermo-Scientific D711 accelerator-based fusion neutron generator. This was done to measure the fission yields ofmore » short-lived fission products and to examine the differences between the delayed fission product signatures of the three actinides. The measured data were compared against previously published results for {sup 89}Kr, −90, and −92 and {sup 138}Xe, −139, and −140. The average percent deviation of the measured values from the Evaluated Nuclear Data Files VII.1 (ENDF/B-VII.1) for thorium, depleted-uranium, and highly-enriched uranium were −10.2%, 4.5%, and −12.9%, respectively. In addition to the measurements of the six known fission products, 23 new fission yield measurements from {sup 84}As to {sup 146}La are presented.« less

VOLATILE CHLORIDE PROCESS FOR THE RECOVERY OF METAL VALUES

DOEpatents

Hanley, W.R.

1959-01-01

A process is presented for recovering uranium, iron, and aluminum from centain shale type ores which contain uranium in minute quantities. The ore is heated wiih a chlorinating agent. such as chlorine, to form a volatilized stream of metal chlorides. The chloride stream is then passed through granular alumina which preferentially absorbs the volatile uranium chloride and from which the uranium may later be recovered. The remaining volatilized chlorides, chiefly those of iron and aluminum, are further treated to recover chlorine gas for recycle, and to recover ferric oxide and aluminum oxide as valuable by-products.

FUSED SALT PROCESS FOR RECOVERY OF VALUES FROM USED NUCLEAR REACTOR FUELS

DOEpatents

Moore, R.H.

1960-08-01

A process is given for recovering plutonium from a neutron-irradiated uranium mass (oxide or alloy) by dissolving the mass in an about equimolar alkali metalaluminum double chloride, adding aluminum metal to the mixture obtained at a temperature of between 260 and 860 deg C, and separating a uranium-containing metal phase and a plutonium-chloride- and fission-product chloridecontaining salt phase. Dissolution can be expedited by passing carbon tetrachloride vapors through the double salt. Separation without reduction of plutonium from neutron- bombarded uranium and that of cerium from uranium are also discussed.

METHOD OF SEPARATION

DOEpatents

Boyd, G.E.

1958-08-26

A process is presented fer separating uranium, plutonium, and fission products ions from uranyl nitrate solutions having a pH value between 1 and 3 obtained by dissolving neutron irradiated uranium. The method consists in passing such solutions through a bed of cation exchange resin, which may be a sulfonated phenol formaidehyde type. Following the adsorption step the resin is first treated with a solution of 0.2M to 0.3M sulfuric acid to desorb the uranium. Fission product ions are then desorbed by treating the resin in phosphoric acid and 1M in nitric acid. Lastly, the plutonium may be desorbed by treating the resin with a solution approximately 0.8M in phosphoric acid and 1M in nitric acid.

Environmental site description for a Uranium Atomic Vapor Laser Isotope Separation (U-AVLIS) production plant at the Paducah Gaseous Diffusion Plant site

DOE Office of Scientific and Technical Information (OSTI.GOV)

Marmer, G.J.; Dunn, C.P.; Moeller, K.L.

Uranium enrichment in the United States has utilized a diffusion process to preferentially enrich the U-235 isotope in the uranium product. The U-AVLIS process is based on electrostatic extraction of photoionized U-235 atoms from an atomic vapor stream created by electron-beam vaporization of uranium metal alloy. The U-235 atoms are ionized when precisely tuned laser light -- of appropriate power, spectral, and temporal characteristics -- illuminates the uranium vapor and selectively photoionizes the U-235 isotope. A programmatic document for use in screening DOE site to locate a U-AVLIS production plant was developed and implemented in two parts. The first partmore » consisted of a series of screening analyses, based on exclusionary and other criteria, that identified a reasonable number of candidate sites. These sites were subjected to a more rigorous and detailed comparative analysis for the purpose of developing a short list of reasonable alternative sites for later environmental examination. This environmental site description (ESD) provides a detailed description of the PGDP site and vicinity suitable for use in an environmental impact statement (EIS). The report is based on existing literature, data collected at the site, and information collected by Argonne National Laboratory (ANL) staff during a site visit. 65 refs., 15 tabs.« less

Determination of the elemental concentration of uranium and thorium in the products and by-products of amang tin tailings process

NASA Astrophysics Data System (ADS)

Alnour, I. A.; Wagiran, H.; Ibrahim, N.; Hamzah, S.; Elias, M. S.

2017-01-01

Amang or tin tailing is processed into concentrated ores and other economical valuable minerals such as monazite, zircon, xenotime, ilmenite etc. Besides that, the tailings from these ores may have a significant potential source of radiation exposure to amang plants' workers. This study was conducted to determine the elemental concentration of uranium and thorium in mineral samples collected from five amang tailing factories. The concentration of uranium and thorium was carried out by using instrumental neutron activation analysis (INAA) relative technique. The concentration of uranium and thorium in ppm obtained in this study are as follows: raw (189-1064) and (622-4965); monazite (1076-1988) and (3467-33578); xenotime 4053 and 5540; zircon (309-3090) and (387-6339); ilmenite (104-583) and (88-1205); rutile (212-889) and (44-1119); pyrite (7-43) and (9-132); and waste (5-338) and (9-1218) respectively. The analysis results shows that the monazite, xenotime and zircon have high content of uranium and thorium, whereas ilmenite, rutile, pyrite and waste have lower concentration compare with raw materials after tailing process. The highest values of uranium and thorium concentrations (4053 ± 428 ppm and 33578 ± 873 ppm, respectively) were observed in xenotime and monazite; whereas the lowest value was 5.48 ± 0.86 ppm of uranium recorded in waste (sand) and 9 ± 0.32 ppm of thorium for waste (sand) and pyrite.

METHOD OF SEPARATING URANIUM, PLUTONIUM AND FISSION PRODUCTS BY BROMINATION AND DISTILLATION

DOEpatents

Jaffey, A.H.; Seaborg, G.T.

1958-12-23

The method for separation of plutonium from uranium and radioactive fission products obtained by neutron irradiation of uranlum consists of reacting the lrradiated material with either bromine, hydrogen bromide, alumlnum bromide, or sulfur and bromine at an elevated temperature to form the bromides of all the elements, then recovering substantlally pure plutonium bromide by dlstillatlon in combinatlon with selective condensatlon at prescribed temperature and pressure.

Traversing probe system

DOEpatents

Mashburn, Douglas N.; Stevens, Richard H.; Woodall, Harold C.

1977-01-01

This invention comprises a rotatable annular probe-positioner which carries at least one radially disposed sensing probe, such as a Pitot tube having a right-angled tip. The positioner can be coaxially and rotatably mounted within a compressor casing or the like and then actuated to orient the sensing probe as required to make measurements at selected stations in the annulus between the positioner and compressor casing. The positioner can be actuated to (a) selectively move the probe along its own axis, (b) adjust the yaw angle of the right-angled probe tip, and (c) revolve the probe about the axis common to the positioner and casing. A cam plate engages a cam-follower portion of the probe and normally rotates with the positioner. The positioner includes a first-motor-driven ring gear which effects slidable movement of the probe by rotating the positioner at a time when an external pneumatic cylinder is actuated to engage the cam plate and hold it stationary. When the pneumatic cylinder is not actuated, this ring gear can be driven to revolve the positioner and thus the probe to a desired circumferential location about the above-mentioned common axis. A second motor-driven ring gear included in the positioner can be driven to rotate the probe about its axis, thus adjusting the yaw angle of the probe tip. The positioner can be used in highly corrosive atmosphere, such as gaseous uranium hexafluoride.

Energy balance for uranium recovery from seawater

DOE Office of Scientific and Technical Information (OSTI.GOV)

Schneider, E.; Lindner, H.

The energy return on investment (EROI) of an energy resource is the ratio of the energy it ultimately produces to the energy used to recover it. EROI is a key viability measure for a new recovery technology, particularly in its early stages of development when financial cost assessment would be premature or highly uncertain. This paper estimates the EROI of uranium recovery from seawater via a braid adsorbent technology. In this paper, the energy cost of obtaining uranium from seawater is assessed by breaking the production chain into three processes: adsorbent production, adsorbent deployment and mooring, and uranium elution andmore » purification. Both direct and embodied energy inputs are considered. Direct energy is the energy used by the processes themselves, while embodied energy is used to fabricate their material, equipment or chemical inputs. If the uranium is used in a once-through fuel cycle, the braid adsorbent technology EROI ranges from 12 to 27, depending on still-uncertain performance and system design parameters. It is highly sensitive to the adsorbent capacity in grams of U captured per kg of adsorbent as well as to potential economies in chemical use. This compares to an EROI of ca. 300 for contemporary terrestrial mining. It is important to note that these figures only consider the mineral extraction step in the fuel cycle. At a reference performance level of 2.76 g U recovered per kg adsorbent immersed, the largest energy consumers are the chemicals used in adsorbent production (63%), anchor chain mooring system fabrication and operations (17%), and unit processes in the adsorbent production step (12%). (authors)« less

The Military Significance of Small Uranium Enrichment Facilities Fed with Low-Enrichment Uranium (Redacted)

DTIC Science & Technology

1969-12-01

a five-year supply of enriched uranium for reactor fuel . Nevertheless, it seems clear that some foreign enrichment developments are approaching a...produc- tion of fissile material could powerfully influence the assessment of risks and benefits of a nuclear weapons development program . Since... program is likely to include the production of its own relatively pure fissile plutonium. This would involve more rapid cycling and reprocessing of fuel

SINTERING METAL OXIDES

DOEpatents

Roake, W.E.

1960-09-13

A process is given for producing uranium dioxide material of great density by preparing a compacted mixture of uranium dioxide and from 1 to 3 wt.% of calcium hydride, heating the mixture to at least 675 deg C for decomposition of the hydride and then for sintering, preferably in a vacuum, at from 1550 to 2000 deg C. Calcium metal is formed, some uranium is reduced by the calcium to the metal and a product of high density is obtained.

DPASV analytical technique for ppb level uranium analysis

NASA Astrophysics Data System (ADS)

Pal, Sangita; Singha, Mousumi; Meena, Sher Singh

2018-04-01

Determining uranium in ppb level is considered to be most crucial for reuse of water originated in nuclear industries at the time of decontamination of plant effluents generated during uranium (fuel) production, fuel rod fabrication, application in nuclear reactors and comparatively small amount of effluents obtained during laboratory research and developmental work. Higher level of uranium in percentage level can be analyzed through gravimetry, titration etc, whereas inductively coupled plasma-atomic energy spectroscopy (ICP-AES), fluorimeter are well suited for ppm level. For ppb level of uranium, inductively coupled plasma - mass spectroscopy (ICP-MS) or Differential Pulse Anodic Stripping Voltammetry (DPASV) serve the purpose. High precision, accuracy and sensitivity are the crucial for uranium analysis in trace (ppb) level, which are satisfied by ICP-MS and stripping voltammeter. Voltammeter has been found to be less expensive, requires low maintenance and is convenient for measuring uranium in presence of large number of other ions in the waste effluent. In this paper, necessity of uranium concentration quantification for recovery as well as safe disposal of plant effluent, working mechanism of voltammeter w.r.t. uranium analysis in ppb level with its standard deviation and a data comparison with ICP-MS has been represented.

Potential Aquifer Vulnerability in Regions Down-Gradient from ...

EPA Pesticide Factsheets

Sandstone-hosted roll-front uranium ore deposits originate when U(VI) dissolved in groundwater is reduced and precipitated as insoluble U(IV) minerals. Groundwater redox geochemistry, aqueous complexation, and solute migration are instrumental in leaching uranium from source rocks and transporting it in low concentrations to a chemical redox interface where it is deposited in an ore zone typically containing the uranium minerals uraninite, pitchblende, and/or coffinite; various iron sulfides; native selenium; clays; and calcite. In situ recovery (ISR) of these uranium ores is a process of contacting the uranium mineral deposit with leaching (lixiviant) fluids via injection of the lixiviant into wells drilled into the subsurface aquifer that hosts uranium ore, while other extraction wells pump the dissolved uranium after dissolution of the uranium minerals. Environmental concerns during and after ISR include water quality impacts from: 1) potential excursions of leaching solutions away from the injection zone into down-dip, underlying, or overlying aquifers; 2) potential migration of uranium and its decay products (e.g., Ra, Rn, Pb); and, 3) potential migration of redox-sensitive trace metals (e.g., Fe, Mn, Mo, Se, V), metalloids (e.g., As), and anions (e.g., sulfate). This review describes the geochemical processes that control roll-front uranium transport and fate in groundwater systems, identifies potential aquifer vulnerabilities to ISR operations, identifies

Preliminary study of uranium favorability of the Boulder batholith, Montana

DOE Office of Scientific and Technical Information (OSTI.GOV)

Castor, S.B.; Robins, J.W.

1978-01-01

The Boulder batholith of southwestern Montana is a composite Late Cretaceous intrusive mass, mostly composed of quartz monzonite and granodiorite. This study was not restricted to the plutonic rocks; it also includes younger rocks that overlie the batholith, and older rocks that it intrudes. The Boulder batholith area has good overall potential for economic uranium deposits, because its geology is similar to that of areas that contain economic deposits elsewhere in the world, and because at least 35 uranium occurrences of several different types are present. Potential is greatest for the occurrence of small uranium deposits in chalcedony veins andmore » base-metal sulfide veins. Three areas may be favorable for large, low-grade deposits consisting of a number of closely spaced chalcedony veins and enriched wall rock; the Mooney claims, the Boulder area, and the Clancy area. In addition, there is a good possibility of by-product uranium production from phosphatic black shales in the project area. The potential for uranium deposits in breccia masses that cut prebatholith rocks, in manganese-quartz veins near Butte, and in a shear zone that cuts Tertiary rhyolite near Helena cannot be determined on the basis of available information. Low-grade, disseminated, primary uranium concentrations similar to porphyry deposits proposed by Armstrong (1974) may exist in the Boulder batholith, but the primary uranium content of most batholith rocks is low. The geologic environment adjacent to the Boulder batholith is similar in places to that at the Midnite mine in Washington. Some igneous rocks in the project area contain more than 10 ppM U/sub 3/O/sub 8/, and some metasedimentary rocks near the batholith contain reductants such as sulfides and carbonaceous material.« less

Pyroprocessing of Fast Flux Test Facility Nuclear Fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

B.R. Westphal; G.L. Fredrickson; G.G. Galbreth

Used nuclear fuel from the Fast Flux Test Facility (FFTF) was recently transferred to the Idaho National Laboratory and processed by pyroprocessing in the Fuel Conditioning Facility. Approximately 213 kg of uranium from sodium-bonded metallic FFTF fuel was processed over a one year period with the equipment previously used for the processing of EBR-II used fuel. The peak burnup of the FFTF fuel ranged from 10 to 15 atom% for the 900+ chopped elements processed. Fifteen low-enriched uranium ingots were cast following the electrorefining and distillation operations to recover approximately 192 kg of uranium. A material balance on the primarymore » fuel constituents, uranium and zirconium, during the FFTF campaign will be presented along with a brief description of operating parameters. Recoverable uranium during the pyroprocessing of FFTF nuclear fuel was greater than 95% while the purity of the final electrorefined uranium products exceeded 99%.« less

Pyroprocessing of fast flux test facility nuclear fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Westphal, B.R.; Wurth, L.A.; Fredrickson, G.L.

Used nuclear fuel from the Fast Flux Test Facility (FFTF) was recently transferred to the Idaho National Laboratory and processed by pyroprocessing in the Fuel Conditioning Facility. Approximately 213 kg of uranium from sodium-bonded metallic FFTF fuel was processed over a one year period with the equipment previously used for the processing of EBR-II used fuel. The peak burnup of the FFTF fuel ranged from 10 to 15 atom% for the 900+ chopped elements processed. Fifteen low-enriched uranium ingots were cast following the electrorefining and distillation operations to recover approximately 192 kg of uranium. A material balance on the primarymore » fuel constituents, uranium and zirconium, during the FFTF campaign will be presented along with a brief description of operating parameters. Recoverable uranium during the pyroprocessing of FFTF nuclear fuel was greater than 95% while the purity of the final electro-refined uranium products exceeded 99%. (authors)« less

Uranium(VI) interactions with mackinawite in the presence and absence of bicarbonate and oxygen

USGS Publications Warehouse

Gallegos, Tanya J.; Fuller, Christopher C.; Webb, Samuel M.; Betterton, William J.

2013-01-01

Mackinawite, Fe(II)S, samples loaded with uranium (10-5, 10-4, and 10-3 mol U/g FeS) at pH 5, 7, and 9, were characterized using X-ray absorption spectroscopy and X-ray diffraction to determine the effects of pH, bicarbonate, and oxidation on uptake. Under anoxic conditions, a 5 g/L suspension of mackinawite lowered 5 × 10-5 M uranium(VI) to below 30 ppb (1.26 × 10-7 M) U. Between 82 and 88% of the uranium removed from solution by mackinawite was U(IV) and was nearly completely reduced to U(IV) when 0.012 M bicarbonate was added. Near-neighbor coordination consisting of uranium–oxygen and uranium-uranium distances indicates the formation of uraninite in the presence and absence of bicarbonate, suggesting reductive precipitation as the dominant removal mechanism. Following equilibration in air, mackinawite was oxidized to mainly goethite and sulfur and about 76% of U(IV) was reoxidized to U(VI) with coordination of uranium to axial and equatorial oxygen, similar to uranyl. Additionally, uranium-iron distances, typical of coprecipitation of uranium with iron oxides, and uranium-sulfur distances indicating bidentate coordination of U(VI) to sulfate were evident. The affinity of mackinawite and its oxidation products for U(VI) provides impetus for further study of mackinawite as a potential reactive medium for remediation of uranium-contaminated water.

Sintering Uranium Dioxide of Domestic Production. Report No. 78; SINTERIZACION DE DIOXIDO DE URANIO DE PRODUCCION NACIONAL. Informe No. 78

DOE Office of Scientific and Technical Information (OSTI.GOV)

Carrea, A.J.

1963-01-01

After a brief indication of the uranium- oxygen equilibrium and the methods for the preparation of UO/sub 2/, the sintering of UO/sub 2/ is considered. The effects of various sintering atmospheres on the properties of the product are discussed and tabulated. The method used for the processing of domestic ores for the preparation of UO/sub 2/ and the fabricition of the sintered UO/sub 2/are described. The properties of the product obtained are illustrated graphically. (J.S.R.)

Enhanced fuel production in thorium/lithium hybrid blankets utilizing uranium multipliers

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pitulski, R.H.

1979-10-01

A consistent neutronics analysis is performed to determine the effectiveness of uranium bearing neutron multiplier zones on increasing the production of U/sup 233/ in thorium/lithium blankets for use in a tokamak fusion-fission hybrid reactor. The nuclear performance of these blankets is evaluated as a function of zone thicknesses and exposure by using the coupled transport burnup code ANISN-CINDER-HIC. Various parameters such as U/sup 233/, Pu/sup 239/, and H/sup 3/ production rates, the blanket energy multiplication, isotopic composition of the fuels, and neutron leakages into the various zones are evaluated during a 5 year (6 MW.y.m/sup -2/) exposure period. Although themore » results of this study were obtained for a tokomak magnetic fusion device, the qualitative behavior associated with the use of the uranium bearing neutron multiplier should be applicable to all fusion-fission hybrids.« less

Application of the DART Code for the Assessment of Advanced Fuel Behavior

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rest, J.; Totev, T.

2007-07-01

The Dispersion Analysis Research Tool (DART) code is a dispersion fuel analysis code that contains mechanistically-based fuel and reaction-product swelling models, a one dimensional heat transfer analysis, and mechanical deformation models. DART has been used to simulate the irradiation behavior of uranium oxide, uranium silicide, and uranium molybdenum aluminum dispersion fuels, as well as their monolithic counterparts. The thermal-mechanical DART code has been validated against RERTR tests performed in the ATR for irradiation data on interaction thickness, fuel, matrix, and reaction product volume fractions, and plate thickness changes. The DART fission gas behavior model has been validated against UO{sub 2}more » fission gas release data as well as measured fission gas-bubble size distributions. Here DART is utilized to analyze various aspects of the observed bubble growth in U-Mo/Al interaction product. (authors)« less

Mortality (1950–1999) and cancer incidence (1969–1999) of workers in the Port Hope cohort study exposed to a unique combination of radium, uranium and γ-ray doses

PubMed Central

Zablotska, Lydia B; Lane, Rachel S D; Frost, Stanley E

2013-01-01

Objectives Uranium processing workers are exposed to uranium and radium compounds from the ore dust and to γ-ray radiation, but less to radon decay products (RDP), typical of the uranium miners. We examined the risks of these exposures in a cohort of workers from Port Hope radium and uranium refinery and processing plant. Design A retrospective cohort study with carefully documented exposures, which allowed separation of those with primary exposures to radium and uranium. Settings Port Hope, Ontario, Canada, uranium processors with no mining experience. Participants 3000 male and female workers first employed (1932–1980) and followed for mortality (1950–1999) and cancer incidence (1969–1999). Outcome measures Cohort mortality and incidence were compared with the general Canadian population. Poisson regression was used to evaluate the association between cumulative RDP exposures and γ-ray doses and causes of death and cancers potentially related to radium and uranium processing. Results Overall, workers had lower mortality and cancer incidence compared with the general Canadian population. In analyses restricted to men (n=2645), the person-year weighted mean cumulative RDP exposure was 15.9 working level months (WLM) and the mean cumulative whole-body γ-ray dose was 134.4 millisieverts. We observed small, non-statistically significant increases in radiation risks of mortality and incidence of lung cancer due to RDP exposures (excess relative risks/100 WLM=0.21, 95% CI <−0.45 to 1.59 and 0.77, 95% CI <−0.19 to 3.39, respectively), with similar risks for those exposed to radium and uranium. All other causes of death and cancer incidence were not significantly associated with RDP exposures or γ-ray doses or a combination of both. Conclusions In one of the largest cohort studies of workers exposed to radium, uranium and γ-ray doses, no significant radiation-associated risks were observed for any cancer site or cause of death. Continued follow-up and pooling with other cohorts of workers exposed to by-products of radium and uranium processing could provide valuable insight into occupational risks and suspected differences in risk with uranium miners. PMID:23449746

Determination of impurities in uranium matrices by time-of-flight ICP-MS using matrix-matched method

DOE Office of Scientific and Technical Information (OSTI.GOV)

Buerger, Stefan; Riciputi, Lee R; Bostick, Debra A

2007-01-01

The analysis of impurities in uranium matrices is performed in a variety of fields, e.g. for quality control in the production stream converting uranium ores to fuels, as element signatures in nuclear forensics and safeguards, and for non-proliferation control. We have investigated the capabilities of time-of-flight ICP-MS for the analysis of impurities in uranium matrices using a matrix-matched method. The method was applied to the New Brunswick Laboratory CRM 124(1-7) series. For the seven certified reference materials, an overall precision and accuracy of approximately 5% and 14%, respectively, were obtained for 18 analyzed elements.

METHOD OF PRODUCING URANIUM

DOEpatents

Foster, L.S.; Magel, T.T.

1958-05-13

A modified process is described for the production of uranium metal by means of a bomb reduction of UF/sub 4/. Difficulty is sometimes experienced in obtaining complete separation of the uranium from the slag when the process is carried out on a snnall scale, i.e., for the production of 10 grams of U or less. Complete separation may be obtained by incorporating in the reaction mixture a quantity of MnCl/sub 2/, so that this compound is reduced along with the UF/sub 4/ . As a result a U--Mn alloy is formed which has a melting point lower than that of pure U, and consequently the metal remains molten for a longer period allowing more complete separation from the slag.

ADSORPTION-BISMUTH PHOSPHATE METHOD FOR SEPARATING PLUTONIUM

DOEpatents

Russell, E.R.; Adamson, A.W.; Boyd, G.E.

1960-06-28

A process is given for separating plutonium from uranium and fission products. Plutonium and uranium are adsorbed by a cation exchange resin, plutonium is eluted from the adsorbent, and then, after oxidation to the hexavalent state, the plutonium is contacted with a bismuth phosphate carrier precipitate.

Helium and Sulfur Hexafluoride in Musical Instruments

NASA Astrophysics Data System (ADS)

Forinash, Kyle; Dixon, Cory L.

2014-11-01

The effects of inhaled helium on the human voice were investigated in a recent article in The Physics Teacher.1 As mentioned in that article, demonstrations of the effect are a popular classroom activity. If the number of YouTube videos is any indication, the effects of sulfur hexafluoride on the human voice are equally popular. However, there appears to be little information available on the effects of either of these gases on musical instruments.2 We describe here the results of a student project that involved measuring the frequency shifts in an organ pipe, a trumpet, and a trombone as the result of filling the instruments with these two gases. The project was one of several possible end-of-semester projects required in an elective science of sound course for non-science majors.

PRECIPITATION METHOD OF SEPARATING PLUTONIUM FROM CONTAMINATING ELEMENTS

DOEpatents

Sutton, J.B.

1958-02-18

This patent relates to an improved method for the decontamination of plutonium. The process consists broadly in an improvement in a method for recovering plutonium from radioactive uranium fission products in aqueous solutions by decontamination steps including byproduct carrier precipitation comprising the step of introducing a preformed aqueous slurry of a hydroxide of a metal of group IV B into any aqueous acidic solution which contains the plutonium in the hexavalent state, radioactive uranium fission products contaminant and a by-product carrier precipitate and separating the metal hydroxide and by-product precipitate from the solution. The process of this invention is especially useful in the separation of plutonium from radioactive zirconium and columbium fission products.

Selective Extraction of Uranium from Liquid or Supercritical Carbon Dioxide

DOE Office of Scientific and Technical Information (OSTI.GOV)

Farawila, Anne F.; O'Hara, Matthew J.; Wai, Chien M.

2012-07-31

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 -CO2) system was designed tomore » mitigate these problems. Indeed, TBP nitric acid complexes are highly soluble in l/sc -CO2 and are capable of extracting uranium directly from UO2, UO3 and U3O8 powders. This eliminates the need for total acid dissolution of the irradiated fuel. Furthermore, since CO2 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-HNO3 loaded Sc-CO2 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-CO2 is then presented. Next, the uranium extraction efficiencies and kinetics is demonstrated for different oxides and under different physical and chemical conditions: l/sc -CO2 pressure and temperature, TBP/HNO3 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-CO2 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-CO2 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-CO2 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.« less

Evaluation of a measurement system for Uranium electrodeposition control to radiopharmaceuticals production

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tufic Madi Filho; Adonis Marcelo Saliba Silva; Jose Patricio Nahuel Cardenas

2015-07-01

For 2016, studies by international bodies forecast a crisis in the supply of Molybdenum ({sup 99}Mo), which is the generator of {sup 99m}Tc, widely used for medical diagnoses and treatments. As a result, many countries are making efforts to prevent this crisis. Brazil is developing the Brazilian Multipurpose Reactor (RMB) project, under the responsibility of the National Nuclear Energy Commission (CNEN). The RMB is a nuclear reactor for research and production of radioisotopes used in the production of radiopharmaceuticals and radioactive sources, broadly used in industrial and research areas in Brazil. Electrodeposition of uranium is a common practice to createmore » samples for alpha spectrometry and this methodology may be an alternative way to produce targets of low enriched uranium (LEU) to fabricate radiopharmaceuticals, as {sup 99}Mo, used for cancer diagnosis. To study the electrodeposition, a solution of 10 mM uranyl nitrate, in 2-propanol, containing uranium enriched to 2.4% in {sup 235}U, with pH = 1, was prepared and measurements with an alpha spectrometer were performed. These studies are justified by the need to produce {sup 99}Mo since, despite using molybdenum in bulk, Brazil is totally dependent on its import. In this project, we intend to obtain a process that may be technologically feasible to control the radiation targets for {sup 99}Mo production. (authors)« less

METAL COMPOSITIONS

DOEpatents

Seybolt, A.U.

1959-02-01

Alloys of uranium which are strong, hard, and machinable are presented, These alloys of uranium contain bctween 0.1 to 5.0% by weight of at least one noble metal such as rhodium, palladium, and gold. The alloys may be heat treated to obtain a product with iniproved tensile and compression strengths,

9. VIEW OF FOUNDRY FURNACE, DEPLETED URANIUM INGOTS, BERYLLIUM INGOTS, ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

9. VIEW OF FOUNDRY FURNACE, DEPLETED URANIUM INGOTS, BERYLLIUM INGOTS, AND ALUMINUM SHAPES WERE PRODUCED IN THE FOUNDRY. (10/30/56) - Rocky Flats Plant, Non-Nuclear Production Facility, South of Cottonwood Avenue, west of Seventh Avenue & east of Building 460, Golden, Jefferson County, CO

Sustainability of uranium mining and milling: toward quantifying resources and eco-efficiency.

PubMed

Mudd, Gavin M; Diesendorf, Mark

2008-04-01

The mining of uranium has long been a controversial public issue, and a renewed debate has emerged on the potential for nuclear power to help mitigate against climate change. The central thesis of pro-nuclear advocates is the lower carbon intensity of nuclear energy compared to fossil fuels, although there remains very little detailed analysis of the true carbon costs of nuclear energy. In this paper, we compile and analyze a range of data on uranium mining and milling, including uranium resources as well as sustainability metrics such as energy and water consumption and carbon emissions with respect to uranium production-arguably the first time for modern projects. The extent of economically recoverable uranium resources is clearly linked to exploration, technology, and economics but also inextricably to environmental costs such as energy/water/chemicals consumption, greenhouse gas emissions, and social issues. Overall, the data clearly show the sensitivity of sustainability assessments to the ore grade of the uranium deposit being mined and that significant gaps remain in complete sustainability reporting and accounting. This paper is a case study of the energy, water, and carbon costs of uranium mining and milling within the context of the nuclear energy chain.

PREPARATION OF URANIUM TRIOXIDE

DOEpatents

Buckingham, J.S.

1959-09-01

The production of uranium trioxide from aqueous solutions of uranyl nitrate is discussed. The uranium trioxide is produced by adding sulfur or a sulfur-containing compound, such as thiourea, sulfamic acid, sulfuric acid, and ammonium sulfate, to the uranyl solution in an amount of about 0.5% by weight of the uranyl nitrate hexahydrate, evaporating the solution to dryness, and calcining the dry residue. The trioxide obtained by this method furnished a dioxide with a considerably higher reactivity with hydrogen fluoride than a trioxide prepared without the sulfur additive.

Formation of Neogenic Ores on the Dump-Heaps of Old Uranium Mines and on the Mine-Head of Mines under Exploitation; FORMATION DE MINERAUX NEOGENES SUR LES HALDES D'ANCIENNES MINES D'URANIUM ET SUR LE CARREAU DES MINES EN EXPLOITATION

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chervet, J.

1960-01-01

The major degradations suffered by primary and secondary uranium ores under the weathering action of air and water are assessed. Pyritic ores were found to be the most vunerable. The interactions between pynite oxidation products and urantferous compounds often lead to the formation of neogentc ores. (C.J.G.)

SIERRA ANCHA WILDERNESS, ARIZONA.

USGS Publications Warehouse

Wrucke, Chester T.; Light, Thomas D.

1984-01-01

Mineral surveys show that the Sierra Ancha Wilderness in Arizona has demonstrated resources of uranium, asbestos, and iron; probable and substantiated resource potential for uranium, asbestos, and iron; and a probable resource potential for fluorspar. Uranium resources occur in vein and strata-bound deposits in siltstone that underlies much of the wilderness. Deposits of long-staple chrysotile asbestos are likely in parts of the wilderness adjacent to known areas of asbestos production. Magnetite deposits in the wilderness form a small iron resource. No fossil fuel resources were identified in this study.

Natural Transmutation of Actinides via the Fission Reaction in the Closed Thorium-Uranium-Plutonium Fuel Cycle

NASA Astrophysics Data System (ADS)

Marshalkin, V. Ye.; Povyshev, V. M.

2017-12-01

It is shown for a closed thorium-uranium-plutonium fuel cycle that, upon processing of one metric ton of irradiated fuel after each four-year campaign, the radioactive wastes contain 54 kg of fission products, 0.8 kg of thorium, 0.10 kg of uranium isotopes, 0.005 kg of plutonium isotopes, 0.002 kg of neptunium, and "trace" amounts of americium and curium isotopes. This qualitatively simplifies the handling of high-level wastes in nuclear power engineering.

Potential aquifer vulnerability in regions down-gradient from uranium in situ recovery (ISR) sites.

PubMed

Saunders, James A; Pivetz, Bruce E; Voorhies, Nathan; Wilkin, Richard T

2016-12-01

Sandstone-hosted roll-front uranium ore deposits originate when U(VI) dissolved in groundwater is reduced and precipitated as insoluble U(IV) minerals. Groundwater redox geochemistry, aqueous complexation, and solute migration are important in leaching uranium from source rocks and transporting it in low concentrations to a chemical redox interface where it is deposited in an ore zone typically containing the uranium minerals uraninite, pitchblende, and/or coffinite; various iron sulfides; native selenium; clays; and calcite. In situ recovery (ISR) of uranium ores is a process of contacting the uranium mineral deposit with leaching and oxidizing (lixiviant) fluids via injection of the lixiviant into wells drilled into the subsurface aquifer that hosts uranium ore, while other extraction wells pump the dissolved uranium after dissolution of the uranium minerals. Environmental concerns during and after ISR include water quality degradation from: 1) potential excursions of leaching solutions away from the injection zone into down-gradient, underlying, or overlying aquifers; 2) potential migration of uranium and its decay products (e.g., Ra, Rn, Pb); and, 3) potential mobilization and migration of redox-sensitive trace metals (e.g., Fe, Mn, Mo, Se, V), metalloids (e.g., As), and anions (e.g., sulfate). This review describes the geochemical processes that control roll-front uranium transport and fate in groundwater systems, identifies potential aquifer vulnerabilities to ISR operations, identifies data gaps in mitigating these vulnerabilities, and discusses the hydrogeological characterization involved in developing a monitoring program. Published by Elsevier Ltd.

Identification of bacteria synthesizing ribosomal RNA in response to uranium addition during biostimulation at the Rifle, CO Integrated Field Research site

DOE PAGES

McGuinness, Lora R.; Wilkins, Michael J.; Williams, Kenneth H.; ...

2015-09-18

Understanding which organisms are capable of reducing uranium at historically contaminated sites provides crucial information needed to evaluate treatment options and outcomes. One approach is determination of the bacteria which directly respond to uranium addition. In this research, uranium amendments were made to groundwater samples from a site of ongoing biostimulation with acetate. The active microbes in the planktonic phase were deduced by monitoring ribosomes production via RT-PCR. The results indicated several microorganisms were synthesizing ribosomes in proportion with uranium amendment up to 2 μM. Concentrations of U (VI) >2 μM were generally found to inhibit ribosome synthesis. Two activemore » bacteria responding to uranium addition in the field were close relatives of Desulfobacter postgateii and Geobacter bemidjiensis. Since RNA content often increases with growth rate, our findings suggest it is possible to rapidly elucidate active bacteria responding to the addition of uranium in field samples and provides a more targeted approach to stimulate specific populations to enhance radionuclide reduction in contaminated sites.« less

Identification of Bacteria Synthesizing Ribosomal RNA in Response to Uranium Addition During Biostimulation at the Rifle, CO Integrated Field Research Site

PubMed Central

McGuinness, Lora R.; Wilkins, Michael J.; Williams, Kenneth H.; Long, Philip E.; Kerkhof, Lee J.

2015-01-01

Understanding which organisms are capable of reducing uranium at historically contaminated sites provides crucial information needed to evaluate treatment options and outcomes. One approach is determination of the bacteria which directly respond to uranium addition. In this study, uranium amendments were made to groundwater samples from a site of ongoing biostimulation with acetate. The active microbes in the planktonic phase were deduced by monitoring ribosomes production via RT-PCR. The results indicated several microorganisms were synthesizing ribosomes in proportion with uranium amendment up to 2 μM. Concentrations of U (VI) >2 μM were generally found to inhibit ribosome synthesis. Two active bacteria responding to uranium addition in the field were close relatives of Desulfobacter postgateii and Geobacter bemidjiensis. Since RNA content often increases with growth rate, our findings suggest it is possible to rapidly elucidate active bacteria responding to the addition of uranium in field samples and provides a more targeted approach to stimulate specific populations to enhance radionuclide reduction in contaminated sites. PMID:26382047

DOE Office of Scientific and Technical Information (OSTI.GOV)

BEHAR, Christophe; GUIBERTEAU, Philippe; DUPERRET, Bernard

This paper describes the D&D program that is being implemented at France's High Enrichment Gaseous Diffusion Plant, which was designed to supply France's Military with Highly Enriched Uranium. This plant was definitively shut down in June 1996, following French President Jacques Chirac's decision to end production of Highly Enriched Uranium and dismantle the corresponding facilities.

Variably Saturated Flow and Multicomponent Biogeochemical Reactive Transport Modeling of a Uranium Bioremediation Field Experiment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yabusaki, Steven B.; Fang, Yilin; Williams, Kenneth H.

2011-11-01

Field experiments at a former uranium mill tailings site have identified the potential for stimulating indigenous bacteria to catalyze the conversion of aqueous uranium in the +6 oxidation state to immobile solid-associated uranium in the +4 oxidation state. This effectively removes uranium from solution resulting in groundwater concentrations below actionable standards. Three-dimensional, coupled variably-saturated flow and biogeochemical reactive transport modeling of a 2008 in situ uranium bioremediation field experiment is used to better understand the interplay of transport rates and biogeochemical reaction rates that determine the location and magnitude of key reaction products. A comprehensive reaction network, developed largely throughmore » previous 1-D modeling studies, was used to simulate the impacts on uranium behavior of pulsed acetate amendment, seasonal water table variation, spatially-variable physical (hydraulic conductivity, porosity) and geochemical (reactive surface area) material properties. A principal challenge is the mechanistic representation of biologically-mediated terminal electron acceptor process (TEAP) reactions whose products significantly alter geochemical controls on uranium mobility through increases in pH, alkalinity, exchangeable cations, and highly reactive reduction products. In general, these simulations of the 2008 Big Rusty acetate biostimulation field experiment in Rifle, Colorado confirmed previously identified behaviors including (1) initial dominance by iron reducing bacteria that concomitantly reduce aqueous U(VI), (2) sulfate reducing bacteria that become dominant after {approx}30 days and outcompete iron reducers for the acetate electron donor, (3) continuing iron-reducer activity and U(VI) bioreduction during dominantly sulfate reducing conditions, and (4) lower apparent U(VI) removal from groundwater during dominantly sulfate reducing conditions. New knowledge on simultaneously active metal and sulfate reducers has been incorporated into the modeling. In this case, an initially small population of slow growing sulfate reducers is active from the initiation of biostimulation. Three-dimensional, variably saturated flow modeling was used to address impacts of a falling water table during acetate injection. These impacts included a significant reduction in aquifer saturated thickness and isolation of residual reactants and products, as well as unmitigated uranium, in the newly unsaturated vadose zone. High permeability sandy gravel structures resulted in locally high flow rates in the vicinity of injection wells that increased acetate dilution. In downgradient locations, these structures created preferential flow paths for acetate delivery that enhanced local zones of TEAP reactivity and subsidiary reactions. Conversely, smaller transport rates associated with the lower permeability lithofacies (e.g., fine) and vadose zone were shown to limit acetate access and reaction. Once accessed by acetate, however, these same zones limited subsequent acetate dilution and provided longer residence times that resulted in higher concentrations of TEAP products when terminal electron donors and acceptors were not limiting. Finally, facies-based porosity and reactive surface area variations were shown to affect aqueous uranium concentration distributions; however, the ranges were sufficiently small to preserve general trends. Large computer memory and high computational performance were required to simulate the detailed coupled process models for multiple biogeochemical components in highly resolved heterogeneous materials for the 110-day field experiment and 50 days of post-biostimulation behavior. In this case, a highly-scalable subsurface simulator operating on 128 processor cores for 12 hours was used to simulate each realization. An equivalent simulation without parallel processing would have taken 60 days, assuming sufficient memory was available.« less

SEPARATION OF URANIUM, PLUTONIUM, AND FISSION PRODUCTS

DOEpatents

Spence, R.; Lister, M.W.

1958-12-16

Uranium and plutonium can be separated from neutron-lrradiated uranium by a process consisting of dissolvlng the lrradiated material in nitric acid, saturating the solution with a nitrate salt such as ammonium nitrate, rendering the solution substantially neutral with a base such as ammonia, adding a reducing agent such as hydroxylamine to change plutonium to the trivalent state, treating the solution with a substantially water immiscible organic solvent such as dibutoxy diethylether to selectively extract the uranium, maklng the residual aqueous solutlon acid with nitric acid, adding an oxidizing agent such as ammonlum bromate to oxidize the plutonium to the hexavalent state, and selectlvely extracting the plutonium by means of an immlscible solvent, such as dibutoxy dlethyletber.

BASIC PEROXIDE PRECIPITATION METHOD OF SEPARATING PLUTONIUM FROM CONTAMINANTS

DOEpatents

Seaborg, G.T.; Perlman, I.

1959-02-10

A process is described for the separation from each other of uranyl values, tetravalent plutonium values and fission products contained in an aqueous acidic solution. First the pH of the solution is adjusted to between 2.5 and 8 and hydrogen peroxide is then added to the solution causing precipitation of uranium peroxide which carries any plutonium values present, while the fission products remain in solution. Separation of the uranium and plutonium values is then effected by dissolving the peroxide precipitate in an acidic solution and incorporating a second carrier precipitate, selective for plutonium. The plutonium values are thus carried from the solution while the uranium remains flissolved. The second carrier precipitate may be selected from among the group consisting of rare earth fluorides, and oxalates, zirconium phosphate, and bismuth lihosphate.

The uranium deposit at the Yellow Canary claims, Daggett County, Utah

USGS Publications Warehouse

Wilmarth, V.R.; Vickers, R.C.; McKeown, F.A.; Beroni, E.P.

1952-01-01

The Yellow Canary claims uranium deposit is on the west side of Red Creek Canyon in the northern part of the Uinta Mountains, Daggett County, Utah. The claims have been developed by two adits, three open cuts, and several hundred deep of bulldozer trenches. No uranium ore has been produced from this deposit. The uranium deposit at the Yellow Canary claims is in the Red Creek quartzite of pre-Cambrian age. The formation is composed of intercalated beds of quartzite, hornblendite, garnet schist, staurolite schist, and quartz-mica schist and is intruded by diorite dikes. A thick unit of highly fractured white quatrzite at the top of the formation contains tyutamunite as coatings on fracture surfaces. The tyutamunite is associated with carnotite, volborthite, iron oxides, azurite, malachite, brochantite, and hyalite. The secondary uranium and vanadium minerals are believed to be alteration products of primary minerals. The uranium content of 15 samples from this property ranged from 0.000 to 0.57 percent.

SOLVENT EXTRACTION PROCESS FOR THE SEPARATION OF URANIUM AND THORIUM FROM PROTACTINIUM AND FISSION PRODUCTS

DOEpatents

Rainey, R.H.; Moore, J.G.

1962-08-14

A liquid-liquid extraction process was developed for recovering thorium and uranium values from a neutron irradiated thorium composition. They are separated from a solvent extraction system comprising a first end extraction stage for introducing an aqueous feed containing thorium and uranium into the system consisting of a plurality of intermediate extractiorr stages and a second end extractron stage for introducing an aqueous immiscible selective organic solvent for thorium and uranium in countercurrent contact therein with the aqueous feed. A nitrate iondeficient aqueous feed solution containing thorium and uranium was introduced into the first end extraction stage in countercurrent contact with the organic solvent entering the system from the second end extraction stage while intro ducing an aqueous solution of salting nitric acid into any one of the intermediate extraction stages of the system. The resultant thorium and uranium-laden organic solvent was removed at a point preceding the first end extraction stage of the system. (AEC)

TENORM (Technologically Enhanced Naturally Occurring Radioactive Materials)

MedlinePlus

... and Titanium Mining Wastes Rare Earths Mining Wastes Uranium Mining Wastes Copper Mining and Production Wastes Bauxite and Alumina Production Wastes Energy production Oil and Gas Production Wastes Coal Combustion Residuals ​Water ...

Uranium-Loaded Water Treatment Resins: 'Equivalent Feed' at NRC and Agreement State-Licensed Uranium Recovery Facilities - 12094

DOE Office of Scientific and Technical Information (OSTI.GOV)

Camper, Larry W.; Michalak, Paul; Cohen, Stephen

Community Water Systems (CWSs) are required to remove uranium from drinking water to meet EPA standards. Similarly, mining operations are required to remove uranium from their dewatering discharges to meet permitted surface water discharge limits. Ion exchange (IX) is the primary treatment strategy used by these operations, which loads uranium onto resin beads. Presently, uranium-loaded resin from CWSs and mining operations can be disposed as a waste product or processed by NRC- or Agreement State-licensed uranium recovery facilities if that licensed facility has applied for and received permission to process 'alternate feed'. The disposal of uranium-loaded resin is costly andmore » the cost to amend a uranium recovery license to accept alternate feed can be a strong disincentive to commercial uranium recovery facilities. In response to this issue, the NRC issued a Regulatory Issue Summary (RIS) to clarify the agency's policy that uranium-loaded resin from CWSs and mining operations can be processed by NRC- or Agreement State-licensed uranium recovery facilities without the need for an alternate feed license amendment when these resins are essentially the same, chemically and physically, to resins that licensed uranium recovery facilities currently use (i.e., equivalent feed). NRC staff is clarifying its current alternate feed policy to declare IX resins as equivalent feed. This clarification is necessary to alleviate a regulatory and financial burden on facilities that filter uranium using IX resin, such as CWSs and mine dewatering operations. Disposing of those resins in a licensed facility could be 40 to 50 percent of the total operations and maintenance (O and M) cost for a CWS. Allowing uranium recovery facilities to treat these resins without requiring a license amendment lowers O and M costs and captures a valuable natural resource. (authors)« less

Nuclear energy in Europe: uranium flow modeling and fuel cycle scenario trade-offs from a sustainability perspective.

PubMed

Tendall, Danielle M; Binder, Claudia R

2011-03-15

The European nuclear fuel cycle (covering the EU-27, Switzerland and Ukraine) was modeled using material flow analysis (MFA).The analysis was based on publicly available data from nuclear energy agencies and industries, national trade offices, and nongovernmental organizations. Military uranium was not considered due to lack of accessible data. Nuclear fuel cycle scenarios varying spent fuel reprocessing, depleted uranium re-enrichment, enrichment assays, and use of fast neutron reactors, were established. They were then assessed according to environmental, economic and social criteria such as resource depletion, waste production, chemical and radiation emissions, costs, and proliferation risks. The most preferable scenario in the short term is a combination of reduced tails assay and enrichment grade, allowing a 17.9% reduction of uranium demand without significantly increasing environmental, economic, or social risks. In the long term, fast reactors could theoretically achieve a 99.4% decrease in uranium demand and nuclear waste production. However, this involves important costs and proliferation risks. Increasing material efficiency is not systematically correlated with the reduction of other risks. This suggests that an overall optimization of the nuclear fuel cycle is difficult to obtain. Therefore, criteria must be weighted according to stakeholder interests in order to determine the most sustainable solution. This paper models the flows of uranium and associated materials in Europe, and provides a decision support tool for identifying the trade-offs of the alternative nuclear fuel cycles considered.

Uranium: Prices, rise, then fall

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pool, T.C.

Uranium prices hit eight-year highs in both market tiers,more » $$16.60/lb U{sub 3}O{sub 8} for non-former Soviet Union (FSU) origin and $$15.50 for FSU origin during mid 1996. However, they declined to $14.70 and $13.90, respectively, by the end of the year. Increased uranium prices continue to encourage new production and restarts of production facilities presently on standby. Australia scrapped its {open_quotes}three-mine{close_quotes} policy following the ouster of the Labor party in a March election. The move opens the way for increasing competition with Canada`s low-cost producers. Other events in the industry during 1996 that have current or potential impacts on the market include: approval of legislation outlining the ground rules for privatization of the US Enrichment Corp. (USEC) and the subsequent sales of converted Russian highly enriched uranium (HEU) from its nuclear weapons program, announcement of sales plans for converted US HEU and other surplus material through either the Department of Energy or USEC, and continuation of quotas for uranium from the FSU in the United States and Europe. In Canada, permitting activities continued on the Cigar Lake and McArthur River projects; and construction commenced on the McClean Lake mill.« less

Cancer and non-cancer mortality among French uranium cycle workers: the TRACY cohort.

PubMed

Samson, Eric; Piot, Irwin; Zhivin, Sergey; Richardson, David B; Laroche, Pierre; Serond, Ana-Paula; Laurier, Dominique; Laurent, Olivier

2016-04-05

The health effects of internal contamination by radionuclides, and notably by uranium, are poorly characterised. New cohorts of uranium workers are needed to better examine these effects. This paper analyses for the first time the mortality profile of the French cohort of uranium cycle workers. It considers mortality from cancer and non-cancer causes. The cohort includes workers employed at least 6 months between 1958 and 2006 in French companies involved in the production of nuclear fuel. Vital status and causes of death were collected from French national registries. Workers were followed-up from 1 January 1968 to 31 December 2008. Standardised mortality ratios (SMRs) were computed based on mortality rates for the French general population. The cohort includes 12,649 workers (88% men). The average length of follow-up is 27 years and the mean age at the end of the study is 60 years. Large mortality deficits are observed for non-cancer causes of death such as non-cancer respiratory diseases (SMR=0.51 (0.41 to 0.63)) and circulatory diseases (SMR=0.68 (0.62 to 0.74)). A mortality deficit of lower magnitude is also observed for all cancers combined (SMR (95% CI): 0.76 (0.71 to 0.81)). Pleural mesothelioma is elevated (SMR=2.04 (1.19 to 3.27)). A healthy worker effect is observed in this new cohort of workers involved in the uranium cycle. Collection of individual information on internal uranium exposure as well as other risk factors is underway, to allow for the investigation of uranium-related risks. Published by the BMJ Publishing Group Limited. For permission to use (where not already granted under a licence) please go to http://www.bmj.com/company/products-services/rights-and-licensing/

Improving the Estimates of Waste from the Recycling of Used Nuclear Fuel - 13410

DOE Office of Scientific and Technical Information (OSTI.GOV)

Phillips, Chris; Willis, William; Carter, Robert

2013-07-01

Estimates are presented of wastes arising from the reprocessing of 50 GWD/tonne, 5 year and 50 year cooled used nuclear fuel (UNF) from Light Water Reactors (LWRs), using the 'NUEX' solvent extraction process. NUEX is a fourth generation aqueous based reprocessing system, comprising shearing and dissolution in nitric acid of the UNF, separation of uranium and mixed uranium-plutonium using solvent extraction in a development of the PUREX process using tri-n-butyl phosphate in a kerosene diluent, purification of the plutonium and uranium-plutonium products, and conversion of them to uranium trioxide and mixed uranium-plutonium dioxides respectively. These products are suitable for usemore » as new LWR uranium oxide and mixed oxide fuel, respectively. Each unit process is described and the wastes that it produces are identified and quantified. Quantification of the process wastes was achieved by use of a detailed process model developed using the Aspen Custom Modeler suite of software and based on both first principles equilibrium and rate data, plus practical experience and data from the industrial scale Thermal Oxide Reprocessing Plant (THORP) at the Sellafield nuclear site in the United Kingdom. By feeding this model with the known concentrations of all species in the incoming UNF, the species and their concentrations in all product and waste streams were produced as the output. By using these data, along with a defined set of assumptions, including regulatory requirements, it was possible to calculate the waste forms, their radioactivities, volumes and quantities. Quantification of secondary wastes, such as plant maintenance, housekeeping and clean-up wastes, was achieved by reviewing actual operating experience from THORP during its hot operation from 1994 to the present time. This work was carried out under a contract from the United States Department of Energy (DOE) and, so as to enable DOE to make valid comparisons with other similar work, a number of assumptions were agreed. These include an assumed reprocessing capacity of 800 tonnes per year, the requirement to remove as waste forms the volatile fission products carbon-14, iodine-129, krypton-85, tritium and ruthenium-106, the restriction of discharge of any water from the facility unless it meets US Environmental Protection Agency drinking water standards, no intentional blending of wastes to lower their classification, and the requirement for the recovered uranium to be sufficiently free from fission products and neutron-absorbing species to allow it to be re-enriched and recycled as nuclear fuel. The results from this work showed that over 99.9% of the radioactivity in the UNF can be concentrated via reprocessing into a fission-product-containing vitrified product, bottles of compressed krypton storage and a cement grout containing the tritium, that together have a volume of only about one eighth the volume of the original UNF. The other waste forms have larger volumes than the original UNF but contain only the remaining 0.1% of the radioactivity. (authors)« less

Thorium and Uranium in the Rock Raw Materials Used For the Production of Building Materials

NASA Astrophysics Data System (ADS)

Pękala, Agnieszka

2017-10-01

Thorium and uranium are constant components of all soils and most minerals thereby rock raw materials. They belong to the particularly dangerous elements because of their natural radioactivity. Evaluation of the content of the radioactive elements in the rock raw materials seems to be necessary in the early stage of the raw material evaluation. The rock formations operated from deposits often are accumulated in landfills and slag heaps where the concentration of the radioactive elements can be many times higher than under natural conditions. In addition, this phenomenon may refer to buildings where rock raw materials are often the main components of the construction materials. The global control system of construction products draws particular attention to the elimination of used construction products containing excessive quantities of the natural radioactive elements. In the presented study were determined the content of thorium and uranium in rock raw materials coming from the Bełachatów lignite deposit. The Bełchatów lignite deposit extracts mainly lignite and secondary numerous accompanying minerals with the raw material importance. In the course of the field works within the framework of the carried out work has been tested 92 samples of rocks of varied petrographic composition. There were carried out analyses of the content of the radioactive elements for 50 samples of limestone of the Jurassic age, 18 samples of kaolinite clays, and 24 samples of siliceous raw materials, represented by opoka-rocks, diatomites, gaizes and clastic rocks. The measurement of content of the natural radioactive elements thorium and uranium based on measuring the frequency counts of gamma quantum, recorded separately in measuring channels. At the same time performed measurements on volume patterns radioactive: thorium and uranium. The studies were carried out in Mazar spectrometer on the powdered material. Standardly performed ten measuring cycles, after which were calculated the concentration of radioactive elements in the sample. The highest concentration of thorium and uranium has been found in the clayey raw material. Their value was respectively from 8 to 12 mg/kg for thorium and from 2.3 to 3.5 mg/kg for uranium. In carbonate sediments the content of thorium was at the level from 0.5 to 2.1 mg/kg and uranium from 0.5-2.2 mg/kg. From a group of the siliceous raw materials the diatomite had a highest concentrations of radioactive elements where the content of thorium was from 1.5 to 1.8 mg/kg and uranium from 1.3 to 1.7 mg/kg.

A modern depleted uranium manufacturing facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zagula, T.A.

1995-07-01

The Specific Manufacturing Capabilities (SMC) Project located at the Idaho National Engineering Laboratory (INEL) and operated by Lockheed Martin Idaho Technologies Co. (LMIT) for the Department of Energy (DOE) manufactures depleted uranium for use in the U.S. Army MIA2 Abrams Heavy Tank Armor Program. Since 1986, SMC has fabricated more than 12 million pounds of depleted uranium (DU) products in a multitude of shapes and sizes with varying metallurgical properties while maintaining security, environmental, health and safety requirements. During initial facility design in the early 1980`s, emphasis on employee safety, radiation control and environmental consciousness was gaining momentum throughout themore » DOE complex. This fact coupled with security and production requirements forced design efforts to focus on incorporating automation, local containment and computerized material accountability at all work stations. The result was a fully automated production facility engineered to manufacture DU armor packages with virtually no human contact while maintaining security, traceability and quality requirements. This hands off approach to handling depleted uranium resulted in minimal radiation exposures and employee injuries. Construction of the manufacturing facility was complete in early 1986 with the first armor package certified in October 1986. Rolling facility construction was completed in 1987 with the first certified plate produced in the fall of 1988. Since 1988 the rolling and manufacturing facilities have delivered more than 2600 armor packages on schedule with 100% final product quality acceptance. During this period there was an annual average of only 2.2 lost time incidents and a single individual maximum radiation exposure of 150 mrem. SMC is an example of designing and operating a facility that meets regulatory requirements with respect to national security, radiation control and personnel safety while achieving production schedules and product quality.« less

Comparison of actinide production in traveling wave and pressurized water reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Osborne, A.G.; Smith, T.A.; Deinert, M.R.

The geopolitical problems associated with civilian nuclear energy production arise in part from the accumulation of transuranics in spent nuclear fuel. A traveling wave reactor is a type of breed-burn reactor that could, if feasible, reduce the overall production of transuranics. In one possible configuration, a cylinder of natural or depleted uranium would be subjected to a fast neutron flux at one end. The neutrons would transmute the uranium, producing plutonium and higher actinides. Under the right conditions, the reactor could become critical, at which point a self-stabilizing fission wave would form and propagate down the length of the reactormore » cylinder. The neutrons from the fission wave would burn the fissile nuclides and transmute uranium ahead of the wave to produce additional fuel. Fission waves in uranium are driven largely by the production and fission of {sup 239}Pu. Simulations have shown that the fuel burnup can reach values greater than 400 MWd/kgIHM, before fission products poison the reaction. In this work we compare the production of plutonium and minor actinides produced in a fission wave to that of a UOX fueled light water reactor, both on an energy normalized basis. The nuclide concentrations in the spent traveling wave reactor fuel are computed using a one-group diffusion model and are verified using Monte Carlo simulations. In the case of the pressurized water reactor, a multi-group collision probability model is used to generate the nuclide quantities. We find that the traveling wave reactor produces about 0.187 g/MWd/kgIHM of transuranics compared to 0.413 g/MWd/kgIHM for a pressurized water reactor running fuel enriched to 4.95 % and burned to 50 MWd/kgIHM. (authors)« less

PDRD (SR13046) TRITIUM PRODUCTION FINAL REPORT

DOE Office of Scientific and Technical Information (OSTI.GOV)

Smith, P.; Sheetz, S.

Utilizing the results of Texas A&M University (TAMU) senior design projects on tritium production in four different small modular reactors (SMR), the Savannah River National Laboratory’s (SRNL) developed an optimization model evaluating tritium production versus uranium utilization under a FY2013 plant directed research development (PDRD) project. The model is a tool that can evaluate varying scenarios and various reactor designs to maximize the production of tritium per unit of unobligated United States (US) origin uranium that is in limited supply. The primary module in the model compares the consumption of uranium for various production reactors against the base case ofmore » Watts Bar I running a nominal load of 1,696 tritium producing burnable absorber rods (TPBARs) with an average refueling of 41,000 kg low enriched uranium (LEU) on an 18 month cycle. After inputting an initial year, starting inventory of unobligated uranium and tritium production forecast, the model will compare and contrast the depletion rate of the LEU between the entered alternatives. This is an annual tritium production rate of approximately 0.059 grams of tritium per kilogram of LEU (g-T/kg-LEU). To date, the Nuclear Regulatory Commission (NRC) license has not been amended to accept a full load of TPBARs so the nominal tritium production has not yet been achieved. The alternatives currently loaded into the model include the three light water SMRs evaluated in TAMU senior projects including, mPower, Holtec and NuScale designs. Initial evaluations of tritium production in light water reactor (LWR) based SMRs using optimized loads TPBARs is on the order 0.02-0.06 grams of tritium per kilogram of LEU used. The TAMU students also chose to model tritium production in the GE-Hitachi SPRISM, a pooltype sodium fast reactor (SFR) utilizing a modified TPBAR type target. The team was unable to complete their project so no data is available. In order to include results from a fast reactor, the SRNL Technical Advisory Committee (TAC) ran a Monte Carlo N-Particle (MCNP) model of a basic SFR for comparison. A 600MWth core surrounded by a lithium blanket produced approximately 1,000 grams of tritium annually with a 13% enriched, 6 year core. This is similar results to a mid-1990’s study where the Fast Flux Test Facility (FFTF), a 400 MWth reactor at the Idaho National Laboratory (INL), could produce about 1,000 grams with an external lithium target. Normalized to the LWRs values, comparative tritium production for an SFR could be approximately 0.31 g-T/kg LEU.« less

Fate of Noble Metals during the Pyroprocessing of Spent Nuclear Fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

B.R. Westphal; D. Vaden; S.X. Li

During the pyroprocessing of spent nuclear fuel by electrochemical techniques, fission products are separated as the fuel is oxidized at the anode and refined uranium is deposited at the cathode. Those fission products that are oxidized into the molten salt electrolyte are considered active metals while those that do not react are considered noble metals. The primary noble metals encountered during pyroprocessing are molybdenum, zirconium, ruthenium, rhodium, palladium, and technetium. Pyroprocessing of spent fuel to date has involved two distinctly different electrorefiner designs, in particular the anode to cathode configuration. For one electrorefiner, the anode and cathode collector are horizontallymore » displaced such that uranium is transported across the electrolyte medium. As expected, the noble metal removal from the uranium during refining is very high, typically in excess of 99%. For the other electrorefiner, the anode and cathode collector are vertically collocated to maximize uranium throughput. This arrangement results in significantly less noble metals removal from the uranium during refining, typically no better than 20%. In addition to electrorefiner design, operating parameters can also influence the retention of noble metals, albeit at the cost of uranium recovery. Experiments performed to date have shown that as much as 100% of the noble metals can be retained by the cladding hulls while affecting the uranium recovery by only 6%. However, it is likely that commercial pyroprocessing of spent fuel will require the uranium recovery to be much closer to 100%. The above mentioned design and operational issues will likely be driven by the effects of noble metal contamination on fuel fabrication and performance. These effects will be presented in terms of thermal properties (expansion, conductivity, and fusion) and radioactivity considerations. Ultimately, the incorporation of minor amounts of noble metals from pyroprocessing into fast reactor metallic fuel will be shown to be of no consequence to reactor performance.« less

Magnesium fluoride reduction-vessel liners. Final report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Latham-Brown, C.E.

1986-03-26

The work described in this report details a program that demonstrated a method by which magnesium fluoride, the by-product of the reduction reaction of uranium tetrafluoride to uranium metal could be used to replace the present graphite used to line the reduction vessel. Utilization of magnesium fluoride (MgF2) as a reduction-vessel liner has the potential to decrease carbon contamination and thereby reduce DU derby rejects due to chemistry. Additionally, there would be the elimination of the cost of the graphite crucible liner and the associated disposal costs by replacement with the by-product of the reduction reaction, which is magnesium fluoride.more » The process would ultimately result in reduced manufacturing costs for derby metal and higher yield of finished penetrators. This was to be accomplished in such a manner as to produce uranium metal derbies which would be accommodated into the present Nuclear Metals-Carolina Metals penetrator production process with minimal changes in equipment and procedures.« less

Compact reaction cell for homogenizing and down-blending highly enriched uranium metal

DOEpatents

McLean, W. II; Miller, P.E.; Horton, J.A.

1995-05-02

The invention is a specialized reaction cell for converting uranium metal to uranium oxide. In a preferred form, the reaction cell comprises a reaction chamber with increasing diameter along its length (e.g. a cylindrical chamber having a diameter of about 2 inches in a lower portion and having a diameter of from about 4 to about 12 inches in an upper portion). Such dimensions are important to achieve the necessary conversion while at the same time affording criticality control and transportability of the cell and product. The reaction chamber further comprises an upper port and a lower port, the lower port allowing for the entry of reactant gases into the reaction chamber, the upper port allowing for the exit of gases from the reaction chamber. A diffuser plate is attached to the lower port of the reaction chamber and serves to shape the flow of gas into the reaction chamber. The reaction cell further comprises means for introducing gases into the reaction chamber and a heating means capable of heating the contents of the reaction chamber. The present invention also relates to a method for converting uranium metal to uranium oxide in the reaction cell of the present invention. The invention is useful for down-blending highly enriched uranium metal by the simultaneous conversion of highly enriched uranium metal and natural or depleted uranium metal to uranium oxide within the reaction cell. 4 figs.

Compact reaction cell for homogenizing and down-blanding highly enriched uranium metal

DOEpatents

McLean, II, William; Miller, Philip E.; Horton, James A.

1995-01-01

The invention is a specialized reaction cell for converting uranium metal to uranium oxide. In a preferred form, the reaction cell comprises a reaction chamber with increasing diameter along its length (e.g. a cylindrical chamber having a diameter of about 2 inches in a lower portion and having a diameter of from about 4 to about 12 inches in an upper portion). Such dimensions are important to achieve the necessary conversion while at the same time affording criticality control and transportability of the cell and product. The reaction chamber further comprises an upper port and a lower port, the lower port allowing for the entry of reactant gasses into the reaction chamber, the upper port allowing for the exit of gasses from the reaction chamber. A diffuser plate is attached to the lower port of the reaction chamber and serves to shape the flow of gas into the reaction chamber. The reaction cell further comprises means for introducing gasses into the reaction chamber and a heating means capable of heating the contents of the reaction chamber. The present invention also relates to a method for converting uranium metal to uranium oxide in the reaction cell of the present invention. The invention is useful for down-blending highly enriched uranium metal by the simultaneous conversion of highly enriched uranium metal and natural or depleted uranium metal to uranium oxide within the reaction cell.

Mortality (1968-2008) in a French cohort of uranium enrichment workers potentially exposed to rapidly soluble uranium compounds.

PubMed

Zhivin, Sergey; Guseva Canu, Irina; Samson, Eric; Laurent, Olivier; Grellier, James; Collomb, Philippe; Zablotska, Lydia B; Laurier, Dominique

2016-03-01

Until recently, enrichment of uranium for civil and military purposes in France was carried out by gaseous diffusion using rapidly soluble uranium compounds. We analysed the relationship between exposure to soluble uranium compounds and exposure to external γ-radiation and mortality in a cohort of 4688 French uranium enrichment workers who were employed between 1964 and 2006. Data on individual annual exposure to radiological and non-radiological hazards were collected for workers of the AREVA NC, CEA and Eurodif uranium enrichment plants from job-exposure matrixes and external dosimetry records, differentiating between natural, enriched and depleted uranium. Cause-specific mortality was compared with the French general population via standardised mortality ratios (SMR), and was analysed via Poisson regression using log-linear and linear excess relative risk models. Over the period of follow-up, 131 161 person-years at risk were accrued and 21% of the subjects had died. A strong healthy worker effect was observed: all causes SMR=0.69, 95% CI 0.65 to 0.74. SMR for pleural cancer was significantly increased (2.3, 95% CI 1.06 to 4.4), but was only based on nine cases. Internal uranium and external γ-radiation exposures were not significantly associated with any cause of mortality. This is the first study of French uranium enrichment workers. Although limited in statistical power, further follow-up of this cohort, estimation of internal uranium doses and pooling with similar cohorts should elucidate potential risks associated with exposure to soluble uranium compounds. Published by the BMJ Publishing Group Limited. For permission to use (where not already granted under a licence) please go to http://www.bmj.com/company/products-services/rights-and-licensing/

Vanadium-uranium extraction from Wyoming vanadiferoud silicates. Report of investigations/1983

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hayashi, M.; Nichols, I.L.; Huiatt, J.L.

1983-11-01

The Bureau of Mines conducted laboratory studies on low-grade vanadiferous silicates from the Pumpkin Buttes and Nine Mile Lake deposits of Wyoming to examine techniques for extracting vanadium and uranium. Recovery from low-grade sources such as these could contribute to future vanadium production and reduce reliance on vanadium imports.

11. VIEW OF DEPLETED URANIUM INGOT. THE METALS WERE PLACED ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

11. VIEW OF DEPLETED URANIUM INGOT. THE METALS WERE PLACED IN CRUCIBLES, LOADED INTO ONE OF EIGHT INDUCTION FURNACES AND MELTED IN A VACUUM ATMOSPHERE. (11/11/57) - Rocky Flats Plant, Non-Nuclear Production Facility, South of Cottonwood Avenue, west of Seventh Avenue & east of Building 460, Golden, Jefferson County, CO

Atomic Fuel, Understanding the Atom Series. Revised.

ERIC Educational Resources Information Center

Hogerton, John F.

This publication is part of the "Understanding the Atom" series. Complete sets of the series are available free to teachers, schools, and public librarians who can make them available for reference or use by groups. Among the topics discussed are: What Atomic Fuel Is; The Odyssey of Uranium; Production of Uranium; Fabrication of Reactor…

76 FR 41528 - Notice of Availability of Environmental Assessment and Finding of No Significant Impact for...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-07-14

...., Irigaray and Christensen Ranch Uranium In-Situ Recovery Projects (Willow Creek Project), Wyoming AGENCY... production operations and the recovery of uranium by in-situ recovery (ISR) at the Irigaray and Christensen Ranch facilities in Johnson and Campbell Counties, Wyoming (Willow Creek Project). The NRC has prepared...

Measurement of the Total Cross Section of Uranium-Uranium Collisions at √{sNN} = 192 . 8 GeV

NASA Astrophysics Data System (ADS)

Baltz, A. J.; Fischer, W.; Blaskiewicz, M.; Gassner, D.; Drees, K. A.; Luo, Y.; Minty, M.; Thieberger, P.; Wilinski, M.; Pshenichnov, I. A.

2014-03-01

The total cross section of Uranium-Uranium at √{sNN} = 192 . 8 GeV has been measured to be 515 +/-13stat +/-22sys barn, which agrees with the calculated theoretical value of 487.3 barn within experimental error. That this total cross section is more than an order of magnitude larger than the geometric ion-ion cross section is primarily due to Bound-Free Pair Production (BFPP) and Electro-Magnetic Dissociation (EMD). Nearly all beam losses were due to geometric, BFPP and EMD collisions. This allowed the determination of the total cross section from the measured beam loss rates and luminosity. The beam loss rate is calculated from a time-dependent measurement of the total beam intensity. The luminosity is measured via the detection of neutron pairs in time-coincidence in the Zero Degree Calorimeters. Apart from a general interest in verifying the calculations experimentally, an accurate prediction of the losses created in the heavy ion collisions is of practical interest for the LHC, where collision products have the potential to quench cryogenically cooled magnets.

Uranium Anodic Dissolution under Slightly Alkaline Conditions Progress Report Full-Scale Demonstration with DU Foil

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gelis, A.; Brown, M. A.; Wiedmeyer, S.

2014-02-18

Argonne National Laboratory (Argonne) is developing an alternative method for digesting irradiated low enriched uranium (LEU) foil targets to produce 99Mo in neutral/alkaline media. This method consists of the electrolytic dissolution of irradiated uranium foil in sodium bicarbonate solution, followed by precipitation of base-insoluble fission and activation products, and uranyl-carbonate species with CaO. The addition of CaO is vital for the effective anion exchange separation of 99MoO 4 2- from the fission products, since most of the interfering anions (e.g., CO 3 2-) are removed from the solution, while molybdate remains in solution. An anion exchange is used to retainmore » and to purify the 99Mo from the filtrate. The electrochemical dissolver has been designed and fabricated in 304 stainless-steel (SS), and tested for the dissolution of a full-size depleted uranium (DU) target, wrapped in Al foil. Future work will include testing with low-burn-up DU foil at Argonne and later with high-burn-up LEU foils at Oak Ridge National Laboratory.« less

Yellow Canary uranium deposits, Daggett County, Utah

USGS Publications Warehouse

Wilmarth, Verl Richard

1953-01-01

The Yellow Canary uranium deposit is on the west side of Red Creek Canyon in the northern part of the Uinta Mountains, Daggett County, Utah. Two claims have been developed by means of an adit, three opencuts, and several hundred feet of bulldozer trenches. No uranium ore has been produced from this deposit. The deposit is in the pre-Cambrian Red Creek quartzite. This formation is composed of intercalated beds of quartzite, hornblendite, garnet schist, staurolite schist, and quartz-mica schist and is intruded by dioritic dikes. A thick unit of highly fractured white quartzite near the top of the formation contains tyuyamunite as coatings on fracture surfaces. The tyuyamunite is associated with carnotite, volborthite, iron oxides, azurite, malachite, brochantite, and hyalite. The uranium and vanadium minerals are probably alteration products of primary minerals. The uranium content of 15 samples from this property ranged from 0.000 to 0.57 percent.

SULFIDE METHOD PLUTONIUM SEPARATION

DOEpatents

Duffield, R.B.

1958-08-12

A process is described for the recovery of plutonium from neutron irradiated uranium solutions. Such a solution is first treated with a soluble sullide, causing precipitation of the plutoniunn and uraniunn values present, along with those impurities which form insoluble sulfides. The precipitate is then treated with a solution of carbonate ions, which will dissolve the uranium and plutonium present while the fission product sulfides remain unaffected. After separation from the residue, this solution may then be treated by any of the usual methods, such as formation of a lanthanum fluoride precipitate, to effect separation of plutoniunn from uranium.

[The risks of out of area missions: depleted uranium].

PubMed

Ciprani, F; Moroni, M

2006-01-01

Depleted uranium (DU), a waste product of uranium enrichment, has several civilian and military applications. It was used as armor-piercing ammunition in international conflicts and was claimed to contribute to health problems, known as the Gulf War Syndrome and recently as the Balkan Syndrome. Leukaemia/Limphoma cases among UN soldiers in the Balkans have been related hypothetically to exposure to DU. The investigations published in the scientific literature give no support for this hypothesis. However future follow-up is necessary for evaluation of long-term risk.

PRODUCTION OF URANIUM TUBING

DOEpatents

Creutz, E.C.

1958-04-15

The manufacture of thin-walled uranium tubing by the hot-piercing techique is described. Uranium billets are preheated to a temperature above 780 d C. The heated billet is fed to a station where it is engaged on its external surface by three convex-surfaced rotating rollers which are set at an angle to the axis of the billet to produce a surface friction force in one direction to force the billet over a piercing mandrel. While being formed around the mandrel and before losing the desired shape, the tube thus formed is cooled by a water spray.

230Th-234U Model-Ages of Some Uranium Standard Reference Materials

DOE Office of Scientific and Technical Information (OSTI.GOV)

Williams, R W; Gaffney, A M; Kristo, M J

The 'age' of a sample of uranium is an important aspect of a nuclear forensic investigation and of the attribution of the material to its source. To the extent that the sample obeys the standard rules of radiochronometry, then the production ages of even very recent material can be determined using the {sup 230}Th-{sup 234}U chronometer. These standard rules may be summarized as (a) the daughter/parent ratio at time=zero must be known, and (b) there has been no daughter/parent fractionation since production. For most samples of uranium, the 'ages' determined using this chronometer are semantically 'model-ages' because (a) some assumptionmore » of the initial {sup 230}Th content in the sample is required and (b) closed-system behavior is assumed. The uranium standard reference materials originally prepared and distributed by the former US National Bureau of Standards and now distributed by New Brunswick Laboratory as certified reference materials (NBS SRM = NBL CRM) are good candidates for samples where both rules are met. The U isotopic standards have known purification and production dates, and closed-system behavior in the solid form (U{sub 3}O{sub 8}) may be assumed with confidence. We present here {sup 230}Th-{sup 234}U model-ages for several of these standards, determined by isotope dilution mass spectrometry using a multicollector ICP-MS, and compare these ages with their known production history.« less

Reactivity of formic acid (HCOOD and DCOOH) at uranium and UO 2.0 surfaces

NASA Astrophysics Data System (ADS)

Manner, William L.; Lloyd, Jane A.; Paffett, Mark T.

1999-10-01

Interactions of DCOOH and HCOOD with uranium and UO 2.0 surfaces have been examined using surface-specific techniques of thermal desorption mass spectroscopy (TDMS), X-ray photoelectron spectroscopy (XPS), and static secondary ion mass spectroscopy (SSIMS). On the clean uranium surface, formate is the predominant product following formic acid adsorption at 100 K. A wide range of products is observed after annealing to 200 K, including formate, hydroxyl, O ads, and H ads (D ads) groups. Adsorbed formate decomposes by 300 K increasing the concentration of the remaining surface products. Surface-adsorbed carbon following TDMS measurements remains as the carbide, as indicated from XPS and SSIMS measurements. The only gaseous species created in high yields from the clean surface upon annealing are H 2, HD, and D 2. On the oxide surface (UO 2.0), adsorbed formate groups are more stable toward dissociation in comparison with the clean uranium surface. Between 100 and 300 K the predominant species on the UO 2.0 surface are surface formate and hydroxyl groups. Hydroxyl groups react between 300 and 350 K to release water from the surface. Adsorbed formate groups decompose between 400 and 500 K to release CO and H 2CO (D 2CO) groups from the oxide surface. Carbon was not detected on the oxide surface by XPS or SSIMS after annealing to 500 K, indicating that all carbon-containing species either desorb in the form of CO-containing products or migrate into the surface.

Nuclear Fuel Reprocessing

DOE Office of Scientific and Technical Information (OSTI.GOV)

Harold F. McFarlane; Terry Todd

2013-11-01

Reprocessing is essential to closing nuclear fuel cycle. Natural uranium contains only 0.7 percent 235U, the fissile (see glossary for technical terms) isotope that produces most of the fission energy in a nuclear power plant. Prior to being used in commercial nuclear fuel, uranium is typically enriched to 3–5% in 235U. If the enrichment process discards depleted uranium at 0.2 percent 235U, it takes more than seven tonnes of uranium feed to produce one tonne of 4%-enriched uranium. Nuclear fuel discharged at the end of its economic lifetime contains less one percent 235U, but still more than the natural ore.more » Less than one percent of the uranium that enters the fuel cycle is actually used in a single pass through the reactor. The other naturally occurring isotope, 238U, directly contributes in a minor way to power generation. However, its main role is to transmute into plutoniumby neutron capture and subsequent radioactive decay of unstable uraniumand neptuniumisotopes. 239Pu and 241Pu are fissile isotopes that produce more than 40% of the fission energy in commercially deployed reactors. It is recovery of the plutonium (and to a lesser extent the uranium) for use in recycled nuclear fuel that has been the primary focus of commercial reprocessing. Uraniumtargets irradiated in special purpose reactors are also reprocessed to obtain the fission product 99Mo, the parent isotope of technetium, which is widely used inmedical procedures. Among the fission products, recovery of such expensive metals as platinum and rhodium is technically achievable, but not economically viable in current market and regulatory conditions. During the past 60 years, many different techniques for reprocessing used nuclear fuel have been proposed and tested in the laboratory. However, commercial reprocessing has been implemented along a single line of aqueous solvent extraction technology called plutonium uranium reduction extraction process (PUREX). Similarly, hundreds of types of reactor fuels have been irradiated for different purposes, but the vast majority of commercial fuel is uranium oxide clad in zirconium alloy tubing. As a result, commercial reprocessing plants have relatively narrow technical requirements for used nuclear that is accepted for processing.« less

Neutrinoless double beta decay with 82SeF6 and direct ion imaging

NASA Astrophysics Data System (ADS)

Nygren, D. R.; Jones, B. J. P.; López-March, N.; Mei, Y.; Psihas, F.; Renner, J.

2018-03-01

We present a new neutrinoless double beta decay concept: the high pressure selenium hexafluoride gas time projection chamber. A promising new detection technique is outlined which combines techniques pioneered in high pressure xenon gas, such as topological discrimination, with the high Q-value afforded by the double beta decay isotope 82Se. The lack of free electrons in SeF6 mandates the use of an ion TPC. The microphysics of ion production and drift, which have many nuances, are explored. Background estimates are presented, suggesting that such a detector may achieve background indices of better than 1 count per ton per year in the region of interest at the 100 kg scale, and still better at the ton-scale.

METHOD OF PREPARING UF$sub 6$

DOEpatents

Davidson, R.; Fried, S.

1959-10-27

A method is described of preparing uraniurn hexafluoride without the use of fluorine gas by reacting uraniurn tetrafluoride with oxygen gas under rigorously anhydrous conditions at 600 to 1300 deg K within a pre-fluorinated nickel vessel.

Estimation of weekly 99Mo production by AHR 200 kW

NASA Astrophysics Data System (ADS)

Siregar, I. H.; Suharyana; Khakim, A.; Siregar, D.; Frida, A. R.

2016-11-01

The estimation of weekly 99Mo production by AHR 200 kW fueled with Low Enriched Uranium Uranyl Nitrate solution has been simulated by using MCNPX computer code. We have employed the AHR design of Babcock & Wilcox Medical Isotope Production System with 9Be Reflector and Stainless steel vessel. We found that when the concentration of uranium in the fresh fuel was 108 gr U/L of UO2(NO3)2 fuel solution, the multiplication factor was 1.0517. The 99Mo concentration reached saturated at tenth day operation. The AHR can produce approximately 1.96×103 6-day-Ci weekly.

Pair production in low-energy collisions of uranium nuclei beyond the monopole approximation

NASA Astrophysics Data System (ADS)

Maltsev, I. A.; Shabaev, V. M.; Tupitsyn, I. I.; Kozhedub, Y. S.; Plunien, G.; Stöhlker, Th.

2017-10-01

A method for calculation of electron-positron pair production in low-energy heavy-ion collisions beyond the monopole approximation is presented. The method is based on the numerical solving of the time-dependent Dirac equation with the full two-center potential. The one-electron wave functions are expanded in the finite basis set constructed on the two-dimensional spatial grid. Employing the developed approach the probabilities of bound-free pair production are calculated for collisions of bare uranium nuclei at the energy near the Coulomb barrier. The obtained results are compared with the corresponding values calculated in the monopole approximation.

SEPARATION OF PLUTONIUM

DOEpatents

Maddock, A.G.; Smith, F.

1959-08-25

A method is described for separating plutonium from uranium and fission products by treating a nitrate solution of fission products, uranium, and hexavalent plutonium with a relatively water-insoluble fluoride to adsorb fission products on the fluoride, treating the residual solution with a reducing agent for plutonium to reduce its valence to four and less, treating the reduced plutonium solution with a relatively insoluble fluoride to adsorb the plutonium on the fluoride, removing the solution, and subsequently treating the fluoride with its adsorbed plutonium with a concentrated aqueous solution of at least one of a group consisting of aluminum nitrate, ferric nitrate, and manganous nitrate to remove the plutonium from the fluoride.

Restructuring the Uranium Mining Industry in Romania: Actual Situation and Prospects

DOE Office of Scientific and Technical Information (OSTI.GOV)

Georgescu, P.D.; Petrescu, S.T.; Iuhas, T.F.

2002-07-01

Uranium prospecting in Romania has started some 50 years ago, when a bilateral agreement between Romania and the former Soviet Union had been concluded and a joint Romanian-Soviet enterprise was created. The production started in 1952 by the opening of some deposits from western Transylvania (Bihor and Ciudanovita). From 1962 the production has continued only with Romanian participation on the ore deposit Avram Iancu and from 1985 on the deposits from Eastern Carpathians (Crucea and Botusana). Starting with 1978 the extracted ores have been completely processed in the Uranium Ore Processing Plant from Feldioara, Brasov. Complying with the initial stipulationsmore » of the Nuclear National Program launched at the beginning of the 1980's, the construction of a nuclear power station in Cernavoda has started in Romania, using natural uranium and heavy water (CANDU type), having five units of 650 MW installed capacity. After 1989 this initial Nuclear National Program was revised and the construction of the first unit (number 1) was finalized and put in operation in 1996. In 2001 the works at the unit number 2 were resumed, having the year 2005 as the scheduled activating date. The future of the other 3 units, being in different construction phases, hasn't been clearly decided. Taking into consideration the exhaustion degree of some ore deposits and from the prospect of exploiting other ore deposits, the uranium industry will be subject of an ample restructuring process. This process includes workings of modernization of the mines in operation and of the processing plant, increasing the profitableness, lowering of the production costs by closing out and ecological rehabilitation of some areas affected by mining works and even new openings of some uraniferous exploitations. This paper presents the actual situation and the prospects of uranium mining industry on the base of some new technical and economical strategic concepts in accordance with the actual Romanian Program for Nuclear Energetics. (authors)« less

Occupational safety data and casualty rates for the uranium fuel cycle. [Glossaries

DOE Office of Scientific and Technical Information (OSTI.GOV)

O'Donnell, F.R.; Hoy, H.C.

1981-10-01

Occupational casualty (injuries, illnesses, fatalities, and lost workdays) and production data are presented and used to calculate occupational casualty incidence rates for technologies that make up the uranium fuel cycle, including: mining, milling, conversion, and enrichment of uranium; fabrication of reactor fuel; transportation of uranium and fuel elements; generation of electric power; and transmission of electric power. Each technology is treated in a separate chapter. All data sources are referenced. All steps used to calculate normalized occupational casualty incidence rates from the data are presented. Rates given include fatalities, serious cases, and lost workdays per 100 man-years worked, per 10/supmore » 12/ Btu of energy output, and per other appropriate units of output.« less

Biogenic formation and growth of uraninite (UO₂).

PubMed

Lee, Seung Yeop; Baik, Min Hoon; Choi, Jong Won

2010-11-15

Biogenic UO₂ (uraninite) nanocrystals may be formed as a product of a microbial reduction process in uranium-enriched environments near the Earth's surface. We investigated the size, nanometer-scale structure, and aggregation state of UO₂ formed by iron-reducing bacterium, Shewanella putrefaciens CN32, from a uranium-rich solution. Characterization of biogenic UO₂ precipitates by high-resolution transmission electron microscopy (HRTEM) revealed that the UO₂ nanoparticles formed were highly aggregated by organic polymers. Nearly all of the nanocrystals were networked in more or less 100 nm diameter spherical aggregates that displayed some concentric UO₂ accumulation with heterogeneity. Interestingly, pure UO₂ nanocrystals were piled on one another at several positions via UO₂-UO₂ interactions, which seem to be intimately related to a specific step in the process of growing large single crystals. In the process, calcium that was easily complexed with aqueous uranium(VI) appeared not to be combined with bioreduced uranium(IV), probably due to its lower binding energy. However, when phosphate was added to the system, calcium was found to be easily associated with uranium(IV), forming a new uranium phase, ningyoite. These results will extend the limited knowledge of microbial uraniferous mineralization and may provide new insights into the fate of aqueous uranium complexes.

Adsorbent Alkali Conditioning for Uranium Adsorption from Seawater. Adsorbent Performance and Technology Cost Evaluation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tsouris, Costas; Mayes, Richard T.; Janke, Christopher James

The Fuel Resources program of the Fuel Cycle Research and Development program of the Office of Nuclear Energy (NE) is focused on identifying and implementing actions to assure that nuclear fuel resources are available in the United States. An immense source of uranium is seawater, which contains an estimated amount of 4.5 billion tonnes of dissolved uranium. This unconventional resource can provide a price cap and ensure centuries of uranium supply for future nuclear energy production. NE initiated a multidisciplinary program with participants from national laboratories, universities, and research institutes to enable technical breakthroughs related to uranium recovery from seawater.more » The goal is to develop advanced adsorbents to reduce the seawater uranium recovery technology cost and uncertainties. Under this program, Oak Ridge National Laboratory (ORNL) has developed a new amidoxime-based adsorbent of high surface area, which tripled the uranium capacity of leading Japanese adsorbents. Parallel efforts have been focused on the optimization of the physicochemical and operating parameters used during the preparation of the adsorbent for deployment. A set of parameters that need to be optimized are related to the conditioning of the adsorbent with alkali solution, which is necessary prior to adsorbent deployment. Previous work indicated that alkali-conditioning parameters significantly affect the adsorbent performance. Initiated in 2014, this study had as a goal to determine optimal parameters such as base type and concentration, temperature, and duration of conditioning that maximize the uranium adsorption performance of amidoxime functionalized adsorbent, while keeping the cost of uranium production low. After base-treatment at various conditions, samples of adsorbent developed at ORNL were tested in this study with batch simulated seawater solution of 8-ppm uranium concentration, batch seawater spiked with uranium nitrate at 75-100 ppb uranium, and continuous-flow natural seawater at the Pacific Northwest National Laboratory (PNNL). Fourier Transform Infrared (FTIR) spectroscopy, Nuclear Magnetic Resonance (NMR) spectroscopy, Scanning Electron Microscopy (SEM), and elemental analysis were used to characterize the adsorbent at different stages of adsorbent preparation and treatment. The study can be divided into two parts: (A) investigation of optimal parameters for KOH adsorbent conditioning and (B) investigation of other possible agents for alkali conditioning, including cost analysis on the basis of uranium production. In the first part of the study, tests with simulated seawater containing 8 ppm uranium showed that the uranium adsorption capacity increased with an increase in the KOH concentration and conditioning time and temperature at each of the KOH concentrations used. FTIR and solid state NMR studies indicated that KOH conditioning converts the amidoxime functional groups into more hydrophilic carboxylate. The longer the KOH conditioning time, up to three hours, the higher was the loading capacity from the simulated seawater solution which is composed of only uranyl, sodium, chloride, and carbonate ions. Marine testing with natural seawater, on the other hand, showed that the uranium adsorption capacity of the adsorbent increased with KOH conditioning temperature, and gradually decreased with increasing KOH conditioning time from one hour to three hours at 80 C. This behavior is due to the conversion of amidoxime to carboxylate. The carboxylate groups are needed to increase the hydrophilicity of the adsorbent; however, conversion of a significant amount of amidoxime to carboxylate leads to loss in selectivity toward uranyl ions. Thus, there is an optimum KOH conditioning time for each temperature at which an optimum ratio between amidoxime and carboxylate is reached. For the case of base conditioning with 0.44 M KOH at 80 C, the optimal conditioning time is 1 hour, with respect to the highest uranium loading capacity from natural seawater. Uptake of other metal ions such as V, Fe, and Cu follows the same trend as that of uranium. Also, the uptake of Ca, Mg, and Zn ions increased with increasing KOH conditioning time, probably due to formation of more carboxylates, which leads to conversion of uranium-selective binding sites to less selective sites. In the second part of the study, inorganic based reagents such as sodium hydroxide (NaOH), sodium carbonate (Na 2CO 3), cesium hydroxide (CsOH), as well as organic based reagents such as ammonium hydroxide (AOH), tetramethylammonium hydroxide (TMAOH), tetraethylammonium hydroxide (TEAOH), triethylmethylammonium hydroxide (TEMAOH), tetrapropylammonium hydroxide (TPAOH) and tetrabutylammonium hydroxide (TBAOH), in addition to KOH, were used for alkaline conditioning. NaOH has emerged as a better reagent for alkaline conditioning of amidoxime-based adsorbent because of higher uranium uptake capacity, higher uranium uptake selectivity ...« less

High-temperature electrically conductive ceramic composite and method for making same

DOEpatents

Beck, David E.; Gooch, Jack G.; Holcombe, Jr., Cressie E.; Masters, David R.

1983-01-01

The present invention relates to a metal-oxide ceramic composition useful in induction heating applications for treating uranium and uranium alloys. The ceramic composition is electrically conductive at room temperature and is nonreactive with molten uranium. The composition is prepared from a particulate admixture of 20 to 50 vol. % niobium and zirconium oxide which may be stabilized with an addition of a further oxide such as magnesium oxide, calcium oxide, or yttria. The composition is prepared by blending the powders, pressing or casting the blend into the desired product configuration, and then sintering the casting or compact in an inert atmosphere. In the casting operation, calcium aluminate is preferably added to the admixture in place of a like quantity of zirconia for providing a cement to help maintain the integrity of the sintered product.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lyon, B.F.

The United States Department of Energy (DOE) currently manages the UF, Cylinder Project. The project was formed to maintain and safely manage depleted uranium hexafluoride (UF{sub 6}) stored in approximately 50,000 carbon steel cylinders. The cylinders located at three DOE sites: the K-25 site at Oak Ridge, Tennessee (K-25); the Paducah Gaseous Diffusion Plant in Paducah, Kentucky (PGDP), and the Portsmouth Gaseous Diffusion Plant (PORTS) in Portsmouth, Ohio. The System Requirements Document (SRD) (LMES 1997a) delineates the requirements of the project. The appropriate actions needed to fulfill these requirements are then specified within the System Engineering Management Plan (SEMP) (LMESmore » 1997b). The report presented herein documents activities that in whole or in part satisfy specific requirements and actions stated in the UF{sub 6} Cylinder Project SRD and SEMP with respect to forecasting cylinder conditions. The wall thickness projections made in this report are based on the assumption that the corrosion trends noted will continue. Some activities planned may substantially reduce the rate of corrosion, in which case the results presented here are conservative. The results presented here are intended to supercede and enlarge the scope of those presented previously (Lyon 1995,1996, 1997). In particular, projections are made for thin-walled cylinders (nominal initial thickness 312.5 mils) and thick-walled cylinders (nominal initial thickness 625 mils). In addition, a preliminary analysis is conducted for the minimum thickness at the head/skirt interface for skirted cylinders.« less

A Radiation-Triggered Surveillance System for UF6 Cylinder Monitoring

DOE Office of Scientific and Technical Information (OSTI.GOV)

Curtis, Michael M.; Myjak, Mitchell J.

This report provides background information and representative scenarios for testing a prototype radiation-triggered surveillance system at an operating facility that handles uranium hexafluoride (UF 6) cylinders. The safeguards objective is to trigger cameras using radiation, or radiation and motion, rather than motion alone, to reduce significantly the number of image files generated by a motion-triggered system. The authors recommend the use of radiation-triggered surveillance at all facilities where cylinder paths are heavily traversed by personnel. The International Atomic Energy Agency (IAEA) has begun using surveillance cameras in the feed and withdrawal areas of gas centrifuge enrichment plants (GCEPs). The camerasmore » generate imagery using elapsed time or motion, but this creates problems in areas occupied 24/7 by personnel. Either motion-or-interval-based triggering generates thousands of review files over the course of a month. Since inspectors must review the files to verify operator material-flow-declarations, a plethora of files significantly extends the review process. The primary advantage of radiation-triggered surveillance is the opportunity to obtain full-time cylinder throughput verification versus what presently amounts to part-time verification. Cost savings should be substantial, as the IAEA presently uses frequent unannounced inspections to verify cylinder-throughput declarations. The use of radiation-triggered surveillance allows the IAEA to implement less frequent unannounced inspections for the purpose of flow verification, but its principal advantage is significantly shorter and more effective inspector video reviews.« less

Process for converting magnesium fluoride to calcium fluoride

DOEpatents

Kreuzmann, A.B.; Palmer, D.A.

1984-12-21

This invention is a process for the conversion of magnesium fluoride to calcium fluoride whereby magnesium fluoride is decomposed by heating in the presence of calcium carbonate, calcium oxide or calcium hydroxide. Magnesium fluoride is a by-product of the reduction of uranium tetrafluoride to form uranium metal and has no known commercial use, thus its production creates a significant storage problem. The advantage of this invention is that the quality of calcium fluoride produced is sufficient to be used in the industrial manufacture of anhydrous hydrogen fluoride, steel mill flux or ceramic applications.

SEPARATION OF RUTHENIUM FROM AQUEOUS SOLUTIONS

DOEpatents

Callis, C.F.; Moore, R.L.

1959-09-01

>The separation of ruthenium from aqueous solutions containing uranium plutonium, ruthenium, and fission products is described. The separation is accomplished by providing a nitric acid solution of plutonium, uranium, ruthenium, and fission products, oxidizing plutonium to the hexavalent state with sodium dichromate, contacting the solution with a water-immiscible organic solvent, such as hexone, to extract plutonyl, uranyl, ruthenium, and fission products, reducing with sodium ferrite the plutonyl in the solvent phase to trivalent plutonium, reextracting from the solvent phase the trivalent plutonium, ruthenium, and some fission products with an aqueous solution containing a salting out agent, introducing ozone into the aqueous acid solution to oxidize plutonium to the hexavalent state and ruthenium to ruthenium tetraoxide, and volatizing off the ruthenium tetraoxide.

Reconnaissance of uranium and copper deposits in parts of New Mexico, Colorado, Utah, Idaho, and Wyoming

USGS Publications Warehouse

Gott, Garland B.; Erickson, Ralph L.

1952-01-01

Because of the common association of uranium and copper in several of the commercial uranium deposits in the Colorado Plateau Province, a reconnaissance was made of several known deposits of copper disseminated through sandstone to determine whether they might be a source of uranium. In order to obtain more information regarding the relationship between copper, uranium and carbonaceous materials, some of the uraniferious asphaltrite deposits in the Shinarump conglomerate along the west flank of the San Rafael Swell were also investigated briefly. During this reconnaissance 18 deposits were examined in New Mexico, eight in Utah, two in Idaho, and one each in Wyoming and Colorado. No uranium deposits of commercial grade are associated with the copper deposits that were examined. The uraniferous asphaltites in the Shinarump conglomerate of Triassic age on the west flank of the San Rafael Swell, however, are promising from the standpoint of commercial uranium production. Spectrographic analyses of crude oil, asphalt, and bituminous shales show a rather consistent suite of trace metals including vanadium, nickel, copper, cobalt, chromium, lead zinc, and molybdenum. The similarity of the metal assemblage, including uranium of the San Rafael Swell asphaltites, to the metal assemblage in crude oil and other bituminous materials suggests that these metals were concentrated in the asphaltites from petroleum. However, the hypothesis that uranium minerals were already present before the hydrocarbons were introduced and that some sort of replacement or uranium minerals by carbon compounds was effected after the petroleum migrated into the uranium deposit should not be disregarded. The widespread association of uranium with asphaltic material suggests that it also may have been concentrated by some agency connected with the formation of petroleum. The problem of the association of uranium and other trace metals with hydrocarbons should be studied further both in the field and in the laboratory.

Nuclear Forensics: Measurements of Uranium Oxides Using Time-of-Flight Secondary Ion Mass Spectrometry (TOF-SIMS)

DTIC Science & Technology

2010-03-01

Isotope Ratio Analysis of Actinides , Fission Products, and Geolocators by High- efficiency Multi-collector Thermal Ionization Mass Spectrometry...Information, 1999. Hou, Xiaolin, and Per Roos. “ Critical Comparison of radiometric and Mass Spectrometric Methods for the Determination of...NUCLEAR FORENSICS: MEASUREMENTS OF URANIUM OXIDES USING TIME-OF-FLIGHT SECONDARY ION MASS

Critical review of analytical techniques for safeguarding the thorium-uranium fuel cycle

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hakkila, E.A.

1978-10-01

Conventional analytical methods applicable to the determination of thorium, uranium, and plutonium in feed, product, and waste streams from reprocessing thorium-based nuclear reactor fuels are reviewed. Separations methods of interest for these analyses are discussed. Recommendations concerning the applicability of various techniques to reprocessing samples are included. 15 tables, 218 references.

Surface degradation of uranium tetrafluoride

DOE PAGES

Tobin, J. G.; Duffin, A. M.; Yu, S. -W.; ...

2017-05-01

A detailed analysis of a single crystal of uranium tetrafluoride has been carried out. The techniques include x-ray absorption spectroscopy, as well as x-ray photoelectron spectroscopy and x-ray emission spectroscopy. Evidence will be presented for the presence of a uranyl species, possibly UO 2F 2, as a product of, or participant in the surface degradation.

Gas exchange-wind speed relation measured with sulfur hexafluoride on a lake

NASA Technical Reports Server (NTRS)

Wanninkhof, R.; Broecker, W. S.; Ledwell, J. R.

1985-01-01

Gas-exchange processes control the uptake and release of various gases in natural systems such as oceans, rivers, and lakes. Not much is known about the effect of wind speed on gas exchange in such systems. In the experiment described here, sulfur hexafluoride was dissolved in lake water, and the rate of escape of the gas with wind speed (at wind speeds up to 6 meters per second) was determined over a 1-month period. A sharp change in the wind speed dependence of the gas-exchange coefficient was found at wind speeds of about 2.4 meters per second, in agreement with the results of wind-tunnel studies. However the gas-exchange coefficients at wind speeds above 3 meters per second were smaller than those observed in wind tunnels and are in agreement with earlier lake and ocean results.

A sulfur hexafluoride sensor using quantum cascade and CO2 laser-based photoacoustic spectroscopy.

PubMed

Rocha, Mila; Sthel, Marcelo; Lima, Guilherme; da Silva, Marcelo; Schramm, Delson; Miklós, András; Vargas, Helion

2010-01-01

The increase in greenhouse gas emissions is a serious environmental problem and has stimulated the scientific community to pay attention to the need for detection and monitoring of gases released into the atmosphere. In this regard, the development of sensitive and selective gas sensors has been the subject of several research programs. An important greenhouse gas is sulphur hexafluoride, an almost non-reactive gas widely employed in industrial processes worldwide. Indeed it is estimated that it has a radiative forcing of 0.52 W/m(2). This work compares two photoacoustic spectrometers, one coupled to a CO(2) laser and another one coupled to a Quantum Cascade (QC) laser, for the detection of SF(6). The laser photoacoustic spectrometers described in this work have been developed for gas detection at small concentrations. Detection limits of 20 ppbv for CO(2) laser and 50 ppbv for quantum cascade laser were obtained.

Uranium Mining and Norm in North America-Some Perspectives on Occupational Radiation Exposure.

PubMed

Brown, Steven H; Chambers, Douglas B

2017-07-01

All soils and rocks contain naturally occurring radioactive materials (NORM). Many ores and raw materials contain relatively elevated levels of natural radionuclides, and processing such materials can further increase the concentrations of naturally occurring radionuclides. In the U.S., these materials are sometimes referred to as technologically-enhanced naturally occurring radioactive materials (TENORM). Examples of NORM minerals include uranium ores, monazite (a source of rare earth minerals), and phosphate rock used to produce phosphate fertilizer. The processing of these materials has the potential to result in above-background radiation exposure to workers. Following a brief review of the sources and potential for worker exposure from NORM in these varied industries, this paper will then present an overview of uranium mining and recovery in North America, including discussion on the mining methods currently being used for both conventional (underground, open pit) and in situ leach (ISL), also referred to as In Situ Recovery (ISR), and the production of NORM materials and wastes associated with these uranium recovery methods. The radiological composition of the NORM products and wastes produced and recent data on radiological exposures received by workers in the North American uranium recovery industry are then described. The paper also identifies the responsible government agencies in the U.S. and Canada assigned the authority to regulate and control occupational exposure from these NORM materials.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mehta, Vrajesh; Maillot, Fabien; Wang, Zheming

Uranyl phosphate solids are often found with uranium ores, and their low solubility makes them promising target phases for in situ remediation of uranium-contaminated subsurface environments. The products and solubility of uranium(VI) precipitated with phosphate can be affected by the pH, dissolved inorganic carbon (DIC) concentration, and co-solute composition (e.g. Na+/Ca2+) of the groundwater. Batch experiments were performed to study the effect of these parameters on the products and extent of uranium precipitation induced by phosphate addition. In the absence of co-solute cations, chernikovite [H3O(UO2)(PO4)•3H2O] precipitated despite uranyl orthophosphate [(UO2)3(PO4)2•4H2O] being thermodynamically more favorable under certain conditions. As determined usingmore » X-ray diffraction, electron microscopy, and laser induced fluorescence spectroscopy, the presence of Na+ or Ca2+ as a co-solute led to the precipitation of sodium autunite ([Na2(UO2)2(PO4)2] and autunite [Ca(UO2)2(PO4)2]), which are structurally similar to chernikovite. In the presence of sodium, the dissolved U(VI) concentrations were generally in agreement with equilibrium predictions of sodium autunite solubility. However, in the calcium-containing systems, the observed concentrations were below the predicted solubility of autunite, suggesting the possibility of uranium adsorption to or incorporation in a calcium phosphate precipitate in addition to the precipitation of autunite.« less

Nuclear forensic analysis of uranium oxide powders interdicted in Victoria, Australia

DOE PAGES

Kristo, Michael Joseph; Keegan, Elizabeth; Colella, Michael; ...

2015-04-13

Nuclear forensic analysis was conducted on two uranium samples confiscated during a police investigation in Victoria, Australia. The first sample, designated NSR-F-270409-1, was a depleted uranium powder of moderate purity (~1000 μg/g total elemental impurities). The chemical form of the uranium was a compound similar to K 2(UO 2) 3O 4·4H 2O. While aliquoting NSR-F-270409-1 for analysis, the body and head of a Tineid moth was discovered in the sample. The second sample, designated NSR-F-270409-2, was also a depleted uranium powder. It was of reasonably high purity (~380 μg/g total elemental impurities). The chemical form of the uranium was primarilymore » UO 3·2H 2O, with minor phases of U 3O 8 and UO 2. While aliquoting NSR-F-270409-2 for analysis, a metal staple of unknown origin was discovered in the sample. The presence of 236U and 232U in both samples indicates that the uranium feed stocks for these samples experienced a neutron flux at some point in their history. The reactor burn-up calculated from the isotopic composition of the uranium is consistent with that of spent fuel from natural uranium (NU) fueled Pu production. These nuclear forensic conclusions allow us to categorically exclude Australia as the origin of the material and greatly reduce the number of candidate sources.« less

Uranium-series constraints on radionuclide transport and groundwater flow at the Nopal I uranium deposit, Sierra Pena Blanca, Mexico.

PubMed

Goldstein, Steven J; Abdel-Fattah, Amr I; Murrell, Michael T; Dobson, Patrick F; Norman, Deborah E; Amato, Ronald S; Nunn, Andrew J

2010-03-01

Uranium-series data for groundwater samples from the Nopal I uranium ore deposit were obtained to place constraints on radionuclide transport and hydrologic processes for a nuclear waste repository located in fractured, unsaturated volcanic tuff. Decreasing uranium concentrations for wells drilled in 2003 are consistent with a simple physical mixing model that indicates that groundwater velocities are low ( approximately 10 m/y). Uranium isotopic constraints, well productivities, and radon systematics also suggest limited groundwater mixing and slow flow in the saturated zone. Uranium isotopic systematics for seepage water collected in the mine adit show a spatial dependence which is consistent with longer water-rock interaction times and higher uranium dissolution inputs at the front adit where the deposit is located. Uranium-series disequilibria measurements for mostly unsaturated zone samples indicate that (230)Th/(238)U activity ratios range from 0.005 to 0.48 and (226)Ra/(238)U activity ratios range from 0.006 to 113. (239)Pu/(238)U mass ratios for the saturated zone are <2 x 10(-14), and Pu mobility in the saturated zone is >1000 times lower than the U mobility. Saturated zone mobility decreases in the order (238)U approximately (226)Ra > (230)Th approximately (239)Pu. Radium and thorium appear to have higher mobility in the unsaturated zone based on U-series data from fractures and seepage water near the deposit.

Uranium redox transition pathways in acetate-amended sediments

USGS Publications Warehouse

Bargar, John R.; Williams, Kenneth H.; Campbell, Kate M.; Long, Philip E.; Stubbs, Joanne E.; Suvorova, Elenal I.; Lezama-Pacheco, Juan S.; Alessi, Daniel S.; Stylo, Malgorzata; Webb, Samuel M.; Davis, James A.; Giammar, Daniel E.; Blue, Lisa Y.; Bernier-Latmani, Rizlan

2013-01-01

Redox transitions of uranium [from U(VI) to U(IV)] in low-temperature sediments govern the mobility of uranium in the environment and the accumulation of uranium in ore bodies, and inform our understanding of Earth’s geochemical history. The molecular-scale mechanistic pathways of these transitions determine the U(IV) products formed, thus influencing uranium isotope fractionation, reoxidation, and transport in sediments. Studies that improve our understanding of these pathways have the potential to substantially advance process understanding across a number of earth sciences disciplines. Detailed mechanistic information regarding uranium redox transitions in field sediments is largely nonexistent, owing to the difficulty of directly observing molecular-scale processes in the subsurface and the compositional/physical complexity of subsurface systems. Here, we present results from an in situ study of uranium redox transitions occurring in aquifer sediments under sulfate-reducing conditions. Based on molecular-scale spectroscopic, pore-scale geochemical, and macroscale aqueous evidence, we propose a biotic–abiotic transition pathway in which biomass-hosted mackinawite (FeS) is an electron source to reduce U(VI) to U(IV), which subsequently reacts with biomass to produce monomeric U(IV) species. A species resembling nanoscale uraninite is also present, implying the operation of at least two redox transition pathways. The presence of multiple pathways in low-temperature sediments unifies apparently contrasting prior observations and helps to explain sustained uranium reduction under disparate biogeochemical conditions. These findings have direct implications for our understanding of uranium bioremediation, ore formation, and global geochemical processes.

Spent nuclear fuel recycling with plasma reduction and etching

DOEpatents

Kim, Yong Ho

2012-06-05

A method of extracting uranium from spent nuclear fuel (SNF) particles is disclosed. Spent nuclear fuel (SNF) (containing oxides of uranium, oxides of fission products (FP) and oxides of transuranic (TRU) elements (including plutonium)) are subjected to a hydrogen plasma and a fluorine plasma. The hydrogen plasma reduces the uranium and plutonium oxides from their oxide state. The fluorine plasma etches the SNF metals to form UF₆ and PuF₄. During subjection of the SNF particles to the fluorine plasma, the temperature is maintained in the range of 1200-2000 deg K to: a) allow any PuF₆ (gas) that is formed to decompose back to PuF₄ (solid), and b) to maintain stability of the UF₆. Uranium (in the form of gaseous UF₆) is easily extracted and separated from the plutonium (in the form of solid PuF₄). The use of plasmas instead of high temperature reactors or flames mitigates the high temperature corrosive atmosphere and the production of PuF₆ (as a final product). Use of plasmas provide faster reaction rates, greater control over the individual electron and ion temperatures, and allow the use of CF₄ or NF₃ as the fluorine sources instead of F₂ or HF.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kristo, Michael Joseph; Keegan, Elizabeth; Colella, Michael

Nuclear forensic analysis was conducted on two uranium samples confiscated during a police investigation in Victoria, Australia. The first sample, designated NSR-F-270409-1, was a depleted uranium powder of moderate purity (~1000 μg/g total elemental impurities). The chemical form of the uranium was a compound similar to K 2(UO 2) 3O 4·4H 2O. While aliquoting NSR-F-270409-1 for analysis, the body and head of a Tineid moth was discovered in the sample. The second sample, designated NSR-F-270409-2, was also a depleted uranium powder. It was of reasonably high purity (~380 μg/g total elemental impurities). The chemical form of the uranium was primarilymore » UO 3·2H 2O, with minor phases of U 3O 8 and UO 2. While aliquoting NSR-F-270409-2 for analysis, a metal staple of unknown origin was discovered in the sample. The presence of 236U and 232U in both samples indicates that the uranium feed stocks for these samples experienced a neutron flux at some point in their history. The reactor burn-up calculated from the isotopic composition of the uranium is consistent with that of spent fuel from natural uranium (NU) fueled Pu production. These nuclear forensic conclusions allow us to categorically exclude Australia as the origin of the material and greatly reduce the number of candidate sources.« less

Method for cleaning bomb-reduced uranium derbies

DOEpatents

Banker, John G.; Wigginton, Hubert L.; Beck, David E.; Holcombe, Cressie E.

1981-01-01

The concentration of carbon in uranium metal ingots induction cast from derbies prepared by the bomb-reduction of uranium tetrafluoride in the presence of magnesium is effectively reduced to less than 100 ppm by removing residual magnesium fluoride from the surface of the derbies prior to casting. This magnesium fluoride is removed from the derbies by immersing them in an alkali metal salt bath which reacts with and decomposes the magnesium fluoride. A water quenching operation followed by a warm nitric acid bath and a water rinse removes the residual salt and reaction products from the derbies.

Method for cleaning bomb-reduced uranium derbies

DOEpatents

Banker, J.G.; Wigginton, H.L.; Beck, D.E.; Holcombe, C.E.

The concentration of carbon in uranium metal ingots induction cast from derbies prepared by the bomb-reduction of uranium tetrafluoride in the presence of magnesium is effectively reduced to less than 100 ppM by removing residual magnesium fluoride from the surface of the derbies prior to casting. This magnesium fluoride is removed from the derbies by immersing them in an alkali metal salt bath which reacts with and decomposes the magnesium fluoride. A water quenching operation followed by a warm nitric acid bath and a water rinse removes the residual salt and reaction products from the derbies.

Uranium migration in spark plasma sintered W/UO2 CERMETS

NASA Astrophysics Data System (ADS)

Tucker, Dennis S.; Wu, Yaqiao; Burns, Jatuporn

2018-03-01

W/UO2 CERMET samples were sintered in a Spark Plasma Sintering (SPS) furnace at various temperature under vacuum and pressure. High Resolution Transmission Electron Microscopy (HRTEM) with Energy Dispersive Spectroscopy (EDS) was performed on the samples to determine interface structures and uranium diffusion from the UO2 particles into the tungsten matrix. Local Electrode Atom Probe (LEAP) was also performed to determine stoichiometry of the UO2 particles. It was seen that uranium diffused approximately 10-15 nm into the tungsten matrix. This is explained in terms of production of oxygen vacancies and Fick's law of diffusion.

Monitoring genotoxic exposure in uranium mines.

PubMed Central

Srám, R J; Dobiás, L; Rössner, P; Veselá, D; Veselý, D; Rakusová, R; Rericha, V

1993-01-01

Recent data from deep uranium mines in Czechoslovakia indicated that mines are exposed to other mutagenic factors in addition to radon daughter products. Mycotoxins were identified as a possible source of mutagens in these mines. Mycotoxins were examined in 38 samples from mines and in throat swabs taken from 116 miners and 78 controls. The following mycotoxins were identified from mines samples: aflatoxins B1 and G1, citrinin, citreoviridin, mycophenolic acid, and sterigmatocystin. Some mold strains isolated from mines and throat swabs were investigated for mutagenic activity by the SOS chromotest and Salmonella assay with strains TA100 and TA98. Mutagenicity was observed, especially with metabolic activation in vitro. These data suggest that mycotoxins produced by molds in uranium mines are a new genotoxic factor for uranium miners. PMID:8143610

HEU Holdup Measurements in 321-M B and Spare U-Al Casting Furnaces

DOE Office of Scientific and Technical Information (OSTI.GOV)

Salaymeh, S.R.

The Analytical Development Section of Savannah River Technology Center (SRTC) was requested by the Facilities Decontamination Division (FDD) to determine the holdup of enriched uranium in the 321-M facility as part of an overall deactivation project of the facility. The 321-M facility was used to fabricate enriched uranium fuel assemblies, lithium-aluminum target tubes, neptunium assemblies, and miscellaneous components for the production reactors. This report covers holdup measurements in two uranium aluminum alloy (U-Al) casting furnaces. Our results indicate an upper limit of 235U content for the B and Spare furnaces of 51 and 67 g respectively. This report discusses themore » methodology, non-destructive assay (NDA) measurements, and results of the uranium holdup on the two furnaces.« less

SEPARATION OF PLUTONIUM VALUES FROM URANIUM AND FISSION PRODUCT VALUES

DOEpatents

Maddock, A.G.; Booth, A.H.

1960-09-13

Separation of plutonium present in small amounts from neutron irradiated uranium by making use of the phenomenon of chemisorption is described. Plutonium in the tetravalent state is chemically absorbed on a fluoride in solid form. The steps for the separation comprise dissolving the irradiated uranium in nitric acid, oxidizing the plutonium in the resulting solution to the hexavalent state, adding to the solution a soluble calcium salt which by the common ion effect inhibits dissolution of the fluoride by the solution, passing the solution through a bed or column of subdivided calcium fluoride which has been sintered to about 8OO deg C to remove the chemisorbable fission products, reducing the plutonium in the solution thus obtained to the tetravalent state, and again passing the solution through a similar bed or column of calcium fluoride to selectively absorb the plutonium, which may then be recovered by treating the calcium fluoride with a solution of ammonium oxalate.

Distillation of cadmium from uranium plutonium cadmium alloy

NASA Astrophysics Data System (ADS)

Kato, Tetsuya; Iizuka, Masatoshi; Inoue, Tadashi; Iwai, Takashi; Arai, Yasuo

2005-04-01

Uranium-plutonium alloy was prepared by distillation of cadmium from U-Pu-Cd ternary alloy. The initial ternary alloy contained 2.9 wt% U and 8.7 wt% Pu other than Cd, which were recovered by molten salt electrolysis with liquid Cd cathode. The distillation experiments were conducted in 10 g scale of the initial alloy using a small-scale distillation furnace equipped with an evaporator and a condenser in a vacuum vessel. After distillation at 1073 K, the weight of the residue was in good agreement with that of the loaded actinides, where the content of Cd decreased to less than 0.05 wt%. The uranium-plutonium alloy product was recovered without adhering to the yttria crucible. The cross section of the product was observed using electron probe micro-analyzer and it was found to consist of a dense material. Almost all of the evaporated Cd was recovered in the condenser and so enclosed well in the apparatus.

Energy resources available to the United States, 1985 to 2000.

PubMed

Hayes, E T

1979-01-19

Energy and the gross national product have grown hand in hand at 3 to 3(1/2) percent a year for almost 40 years. Our energy growth is slowing down and will sentially level off in the 1990's. Our production of oil and gas passed a peak in the early 1970's, and there is no resource base to justify predictions of increased yields. Coal is the only fossil fuel capable of increased production. There are serious doubts that our uranium resources can support a large light-water reactor program. Finding rates for petroleum, natural gas, and uranium are less than half of what they were 20 years ago.

METHOD OF SEPARATING FISSION PRODUCTS FROM FUSED BISMUTH-CONTAINING URANIUM

DOEpatents

Wiswall, R.H.

1958-06-24

A process is described for removing metal selectively from liquid metal compositions. The method effects separation of flssion product metals selectively from dilute solution in fused bismuth, which contains uraniunn in solution without removal of more than 1% of the uranium. The process comprises contacting the fused bismuth with a fused salt composition consisting of sodium, potassium and lithium chlorides, adding to fused bismuth and molten salt a quantity of bismuth chloride which is stoichiometrically required to convert the flssion product metals to be removed to their chlorides which are more stable in the fused salt than in the molten metal and are, therefore, preferentially taken up in the fused salt phase.

40 CFR 98.300 - Definition of the source category.

Code of Federal Regulations, 2011 CFR

2011-07-01

... insulated with or containing sulfur hexafluoride (SF6) or perfluorocarbons (PFCs) used within an electric...-pressure and hermetically sealed-pressure switchgear and gas-insulated lines containing SF6 or PFCs. (4...) Other containers of SF6 or PFC. ...

40 CFR 98.300 - Definition of the source category.

Code of Federal Regulations, 2014 CFR

2014-07-01

... insulated with or containing sulfur hexafluoride (SF6) or perfluorocarbons (PFCs) used within an electric...-pressure and hermetically sealed-pressure switchgear and gas-insulated lines containing SF6 or PFCs. (4...) Other containers of SF6 or PFC. ...

40 CFR 98.300 - Definition of the source category.

Code of Federal Regulations, 2013 CFR

2013-07-01

... insulated with or containing sulfur hexafluoride (SF6) or perfluorocarbons (PFCs) used within an electric...-pressure and hermetically sealed-pressure switchgear and gas-insulated lines containing SF6 or PFCs. (4...) Other containers of SF6 or PFC. ...

40 CFR 98.300 - Definition of the source category.

Code of Federal Regulations, 2012 CFR

2012-07-01

... insulated with or containing sulfur hexafluoride (SF6) or perfluorocarbons (PFCs) used within an electric...-pressure and hermetically sealed-pressure switchgear and gas-insulated lines containing SF6 or PFCs. (4...) Other containers of SF6 or PFC. ...

Translocation of uranium from water to foodstuff while cooking.

PubMed

Krishnapriya, K C; Baksi, Ananya; Chaudhari, Swathi; Gupta, Soujit Sen; Pradeep, T

2015-10-30

The present work report the unusual uranium uptake by foodstuff, especially those rich in carbohydrates like rice when they are cooked in water, contaminated with uranium. The major staple diet in South Asia, rice, was chosen to study its interaction with UO2(2+), the active uranium species in water, using inductively coupled plasma mass spectrometry. Highest uptake limit was checked by cooking rice at very high uranium concentration and it was found to be good scavenger of uranium. To gain insight into the mechanism of uptake, direct interaction of UO2(2+) with monosaccharides was also studied, using electrospray ionization mass spectrometry taking mannose as a model. The studies have been done with dissolved uranium salt, uranyl nitrate hexahydrate (UO2(NO3)2·6H2O), as well as the leachate of a stable oxide of uranium, UO2(s), both of which exist as UO2(2+) in water. Among the eight different rice varieties investigated, Karnataka Ponni showed the maximum uranium uptake whereas unpolished Basmati rice showed the minimum. Interaction with other foodstuffs (potato, carrot, peas, kidney beans and lentils) with and without NaCl affected the extent of chemical interaction but was not consistent with the carbohydrate content. Uranium interaction with D-mannose monitored through ESI-MS, under optimized instrumental parameters, identified the peaks corresponding to uranyl adduct with mannose monomer, dimer and trimer and the species were confirmed by MS/MS studies. The product ion mass spectra showed peaks illustrating water loss from the parent ion as the collision energy was increased, an evidence for the strong interaction of uranium with mannose. This study would constitute the essential background for understanding interaction of uranium with various foods. Extension of this work would involve identification of foodstuff as green heavy metal scavengers. Copyright © 2015. Published by Elsevier B.V.

Hypertension and hematologic parameters in a community near a uranium processing facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wagner, Sara E., E-mail: swagner@uga.edu; Burch, James B.; South Carolina Statewide Cancer Prevention and Control Program, Columbia, SC

Background: Environmental uranium exposure originating as a byproduct of uranium processing can impact human health. The Fernald Feed Materials Production Center functioned as a uranium processing facility from 1951 to 1989, and potential health effects among residents living near this plant were investigated via the Fernald Medical Monitoring Program (FMMP). Methods: Data from 8216 adult FMMP participants were used to test the hypothesis that elevated uranium exposure was associated with indicators of hypertension or changes in hematologic parameters at entry into the program. A cumulative uranium exposure estimate, developed by FMMP investigators, was used to classify exposure. Systolic and diastolicmore » blood pressure and physician diagnoses were used to assess hypertension; and red blood cells, platelets, and white blood cell differential counts were used to characterize hematology. The relationship between uranium exposure and hypertension or hematologic parameters was evaluated using generalized linear models and quantile regression for continuous outcomes, and logistic regression or ordinal logistic regression for categorical outcomes, after adjustment for potential confounding factors. Results: Of 8216 adult FMMP participants 4187 (51%) had low cumulative uranium exposure, 1273 (15%) had moderate exposure, and 2756 (34%) were in the high (>0.50 Sievert) cumulative lifetime uranium exposure category. Participants with elevated uranium exposure had decreased white blood cell and lymphocyte counts and increased eosinophil counts. Female participants with higher uranium exposures had elevated systolic blood pressure compared to women with lower exposures. However, no exposure-related changes were observed in diastolic blood pressure or hypertension diagnoses among female or male participants. Conclusions: Results from this investigation suggest that residents in the vicinity of the Fernald plant with elevated exposure to uranium primarily via inhalation exhibited decreases in white blood cell counts, and small, though statistically significant, gender-specific alterations in systolic blood pressure at entry into the FMMP.« less

Exploration for uranium-vanadium deposits by the U.S. Geological Survey in the Club Mesa area, Uravan district, Montrose County, Colorado

USGS Publications Warehouse

Boardman, R.L.; Litsey, L.R.; Bowers, H.E.

1958-01-01

Club Mesa is one of the most productive areas for uranium-vanadium ore in southwestern Colorago. The average grade of this ore has ranged from about 0.25 to 1.50 percent U3O8 and 1.5 to 5.0 percent V2O5.

IRON COATED URANIUM AND ITS PRODUCTION

DOEpatents

Gray, A.G.

1960-03-15

A method of applying a protective coating to a metallic uranium article is given. The method comprises etching the surface of the article with an etchant solution containlng chloride ions, such as a solution of phosphoric acid and hydrochloric acid, cleaning the etched surface, electroplating iron thereon from a ferrous ammonium sulfate electroplating bath, and soldering an aluminum sheath to the resultant iron layer.

Uranium mining in France

DOE Office of Scientific and Technical Information (OSTI.GOV)

White, L.

Since the onset of the first ''oil shock'' in 1974, France has pursued a policy of steadily increasing energy independence based on nuclear power for generation of electricity. In 1973, nuclear reactors supplied only 8% of France's electrical power. A strong development effort lifted the nuclear share to 23% in 1980, to 66% in 1985, and the plan is to raise the total to 75% by 1990. In 1976, Cogema (Compagnie Generale des Matieres Nucleaires) was organized from the production division of France's Commissariat a l'Energie Atomique (CEA) to handle fuel supply and spent fuel reprocessing for the expanding industrymore » (see subsequent article on Cogema). In parallel with growth of the French nuclear power, Cogema has become a world leader in all aspects of the fuel cycle, providing services not only domestically but internationally as well. As a uranium mining company, Cogema has steadily developed domestic and foreign sources of supply, and over the years it has maintained the world's strongest uranium exploration effort throughout the ups and downs of the market. As a result, the company has become the world's leading uranium supplier, with about 20% of total production contributed either by its domestic mining divisions or overseas subsidiaries.« less

URANIUM RECOVERY PROCESS

DOEpatents

Hyman, H.H.; Dreher, J.L.

1959-07-01

The recovery of uranium from the acidic aqueous metal waste solutions resulting from the bismuth phosphate carrier precipitation of plutonium from solutions of neutron irradiated uranium is described. The waste solutions consist of phosphoric acid, sulfuric acid, and uranium as a uranyl salt, together with salts of the fission products normally associated with neutron irradiated uranium. Generally, the process of the invention involves the partial neutralization of the waste solution with sodium hydroxide, followed by conversion of the solution to a pH 11 by mixing therewith sufficient sodium carbonate. The resultant carbonate-complexed waste is contacted with a titanated silica gel and the adsorbent separated from the aqueous medium. The aqueous solution is then mixed with sufficient acetic acid to bring the pH of the aqueous medium to between 4 and 5, whereby sodium uranyl acetate is precipitated. The precipitate is dissolved in nitric acid and the resulting solution preferably provided with salting out agents. Uranyl nitrate is recovered from the solution by extraction with an ether such as diethyl ether.

Anodic behavior of uranium in AlCl3-1-ethyl-3-methyl-imidazolium chloride ionic liquid

NASA Astrophysics Data System (ADS)

Jiang, Yidong; Luo, Lizhu; Wang, Shaofei; Bin, Ren; Zhang, Guikai; Wang, Xiaolin

2018-01-01

The oxidation state of metals unambiguously affects its anodic behavior in ionic liquid. We systematically investigated the anodic behavior of uranium with different surface oxidation states by electrochemical measurements, spectroscopic methods and surface analysis techniques. In the anodic process, metal uranium can be oxidized to U3+. The corresponding products accumulated on the metal/ILs interface will form a viscous layer. The anodic behavior of uranium is also strongly dependent upon the surface oxide states including thickness and homogeneity of the oxide film. With an increase in the thickness of oxide film, it will be breached at potentials in excess of a critical value. A uniform oxide on uranium surface can be breached evenly, and then the underlying metal starts to dissolve forming a viscous layer which can facilitate uniformly stripping of oxide, thus giving an oxide-free surface. Otherwise, a nonuniform oxide can result in a severe pitted surface with residue oxygen.

Multiple recycle of REMIX fuel at VVER-1000 operation in closed fuel cycle

NASA Astrophysics Data System (ADS)

Alekseev, P. N.; Bobrov, E. A.; Chibinyaev, A. V.; Teplov, P. S.; Dudnikov, A. A.

2015-12-01

The basic features of loading the VVER-1000 core with a new variant of REMIX fuel (REgenerated MIXture of U-Pu oxides) are considered during its multiple recycle in a closed nuclear fuel cycle. The fuel composition is produced on the basis of the uranium-plutonium regenerate extracted at processing the spent nuclear fuel (SNF) from a VVER-1000, depleted uranium, and the fissionable material: 235U as a part of highly enriched uranium (HEU) from warheads superfluous for defense purposes or 233U accumulated in thorium blankets of fusion (electronuclear) neutron sources or fast reactors. Production of such a fuel assumes no use of natural uranium in addition. When converting a part of the VVER-1000 reactors to the closed fuel cycle based on the REMIX technology, the consumption of natural uranium decreases considerably, and there is no substantial degradation of the isotopic composition of plutonium or change in the reactor-safety characteristics at the passage from recycle to recycle.

The effect of hydrogen peroxide on uranium oxide films on 316L stainless steel

NASA Astrophysics Data System (ADS)

Wilbraham, Richard J.; Boxall, Colin; Goddard, David T.; Taylor, Robin J.; Woodbury, Simon E.

2015-09-01

For the first time the effect of hydrogen peroxide on the dissolution of electrodeposited uranium oxide films on 316L stainless steel planchets (acting as simulant uranium-contaminated metal surfaces) has been studied. Analysis of the H2O2-mediated film dissolution processes via open circuit potentiometry, alpha counting and SEM/EDX imaging has shown that in near-neutral solutions of pH 6.1 and at [H2O2] ⩽ 100 μmol dm-3 the electrodeposited uranium oxide layer is freely dissolving, the associated rate of film dissolution being significantly increased over leaching of similar films in pH 6.1 peroxide-free water. At H2O2 concentrations between 1 mmol dm-3 and 0.1 mol dm-3, formation of an insoluble studtite product layer occurs at the surface of the uranium oxide film. In analogy to corrosion processes on common metal substrates such as steel, the studtite layer effectively passivates the underlying uranium oxide layer against subsequent dissolution. Finally, at [H2O2] > 0.1 mol dm-3 the uranium oxide film, again in analogy to common corrosion processes, behaves as if in a transpassive state and begins to dissolve. This transition from passive to transpassive behaviour in the effect of peroxide concentration on UO2 films has not hitherto been observed or explored, either in terms of corrosion processes or otherwise. Through consideration of thermodynamic solubility product and complex formation constant data, we attribute the transition to the formation of soluble uranyl-peroxide complexes under mildly alkaline, high [H2O2] conditions - a conclusion that has implications for the design of both acid minimal, metal ion oxidant-free decontamination strategies with low secondary waste arisings, and single step processes for spent nuclear fuel dissolution such as the Carbonate-based Oxidative Leaching (COL) process.

Biomineralization of U(VI) phosphate promoted by microbially-mediated phytate hydrolysis in contaminated soils

NASA Astrophysics Data System (ADS)

Salome, Kathleen R.; Beazley, Melanie J.; Webb, Samuel M.; Sobecky, Patricia A.; Taillefert, Martial

2017-01-01

The bioreduction of uranium may immobilize a significant fraction of this toxic contaminant in reduced environments at circumneutral pH. In oxic and low pH environments, however, the low solubility of U(VI)-phosphate minerals also makes them good candidates for the immobilization of U(VI) in the solid phase. As inorganic phosphate is generally scarce in soils, the biomineralization of U(VI)-phosphate minerals via microbially-mediated organophosphate hydrolysis may represent the main immobilization process of uranium in these environments. In this study, contaminated sediments were incubated aerobically in two pH conditions to examine whether phytate, a naturally-occurring and abundant organophosphate in soils, could represent a potential phosphorous source to promote U(VI)-phosphate biomineralization by natural microbial communities. While phytate hydrolysis was not evident at pH 7.0, nearly complete hydrolysis was observed both with and without electron donor at pH 5.5, suggesting indigenous microorganisms express acidic phytases in these sediments. While the rate of hydrolysis of phytate generally increased in the presence of uranium, the net rate of inorganic phosphate production in solution was decreased and inositol phosphate intermediates were generated in contrast to similar incubations conducted without uranium. These findings suggest uranium stress enhanced the phytate-metabolism of the microbial community, while simultaneously inhibiting phosphatase production and/or activity by the indigenous population. Finally, phytate hydrolysis drastically decreased uranium solubility, likely due to formation of ternary sorption complexes, U(VI)-phytate precipitates, and U(VI)-phosphate minerals. Overall, the results of this study provide evidence for the ability of natural microbial communities to liberate phosphate from phytate in acidic sediments, possibly as a detoxification mechanism, and demonstrate the potential utility of phytate-promoted uranium immobilization in subsurface environments. These processes should be investigated in more detail with pure cultures isolated from these sediments.

Evaluation of N,N-dialkylamides as promising process extractants

NASA Astrophysics Data System (ADS)

Pathak, P. N.; Prabhu, D. R.; Kanekar, A. S.; Manchanda, V. K.

2010-03-01

Studies carried out at BARC, India on the development of new extractants for reprocessing of spent fuel suggested that while straight chain N,N-dihexyloctanamide (DHOA) is promising alternative to TBP for the reprocessing of irradiated uranium based fuels, branched chain N,N-di(2-ethylhexyl)isobutyramide (D2EHIBA) is suitable for the selective recovery of 233U from irradiated Th. In advanced fuel cycle scenarios, the coprocessing of U/Pu stream appears attractive particularly with respect to development of proliferation resistant technologies. DHOA extracted Pu(IV) more efficiently than TBP, both at trace-level concentration as well as under uranium/plutonium loading conditions. Uranium extraction behavior of DHOA was however, similar to that of TBP during the extraction cycle. Stripping behavior of U and Pu (without any reductant) was better for DHOA than that of TBP. It was observed during batch studies that whereas 99% Pu is stripped in four stages in case of DHOA, only 89% Pu is stripped in case of TBP under identical experimental conditions. DHOA offered better fission product decontamination than that of TBP. GANEX (Group ActiNide EXtraction) and ARTIST (Amide-based Radio-resources Treatment with Interim Storage of Transuranics) processes proposed for actinide partitioning use branched chain amides for the selective extraction of uranium from spent fuel feed solutions. The branched-alkyl monoamide (BAMA) proposed to be used in ARTIST process is N,N-di-(2-ethylhexyl)butyramide (D2EHBA). In this context, the extraction behavior of U(VI) and Pu(IV) were compared using D2EHIBA, TBP, and D2EHBA under similar concentration of nitric acid (0.5 — 6M) and of uranium (0-50g/L). These studies suggested that D2EHIBA is a promising extractant for selective extraction of uranium over plutonium in process streams. Similarly, D2EHIBA offered distinctly better decontamination of 233U over Th and fission products under THOREX feed conditions. The possibility of simultaneous stripping and precipitation of thorium (as oxalate) from loaded organic phase was explored using 0.05M oxalic acid. Ammonium diuranate (ADU) precipitation was performed on the oxalate supernatant for the recovery of uranium. Quantitative recovery (>99.9%) of Th as well as of U was achieved. Radiolytic studies suggested that irradiated DHOA and D2EHIBA behaved better with respect to fission product decontamination as compared to that of TBP.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Waldhoff, Stephanie T.; Anthoff, David; Rose, Steven K.

We use FUND 3.8 to estimate the social cost of four greenhouse gases: carbon dioxide, methane, nitrous oxide, and sulphur hexafluoride emissions. The damage potential for each gas—the ratio of the social cost of the non-carbon dioxide greenhouse gas to the social cost of carbon dioxide—is also estimated. The damage potentials are compared to several metrics, focusing in particular on the global warming potentials, which are frequently used to measure the trade-off between gases in the form of carbon dioxide equivalents. We find that damage potentials could be significantly higher than global warming potentials. This finding implies that previous papersmore » have underestimated the relative importance of reducing non-carbon dioxide greenhouse gas emissions from an economic damage perspective. We show results for a range of sensitivity analyses: carbon dioxide fertilization on agriculture productivity, terrestrial feedbacks, climate sensitivity, discounting, equity weighting, and socioeconomic and emissions scenarios. The sensitivity of the results to carbon dioxide fertilization is a primary focus as it is an important element of climate change that has not been considered in much of the previous literature. We estimate that carbon dioxide fertilization has a large positive impact that reduces the social cost of carbon dioxide with a much smaller effect on the other greenhouse gases. As a result, our estimates of the damage potentials of methane and nitrous oxide are much higher compared to estimates that ignore carbon dioxide fertilization. As a result, our base estimates of the damage potential for methane and nitrous oxide that include carbon dioxide fertilization are twice their respective global warming potentials. Our base estimate of the damage potential of sulphur hexafluoride is similar to the one previous estimate, both almost three times the global warming potential.« less

Uranium mineralization and unconformities: how do they correlate? - A look beyond the classic unconformity-type deposit model?

NASA Astrophysics Data System (ADS)

Markwitz, Vanessa; Porwal, Alok; Campbell McCuaig, T.; Kreuzer, Oliver P.

2010-05-01

Uranium deposits are usually classified based on the characteristics of their host rocks and geological environments (Dahlkamp, 1993; OECD/NEA Red Book and IAEA, 2000; Cuney, 2009). The traditional unconformity-related deposit types are the most economical deposits in the world, with the highest grades amongst all uranium deposit types. In order to predict undiscovered uranium deposits, there is a need to understand the spatial association of uranium mineralization with structures and unconformities. Hydrothermal uranium deposits develop by uranium enriched fluids from source rocks, transported along permeable pathways to their depositional environment. Unconformities are not only separating competent from incompetent sequences, but provide the physico-chemical gradient in the depositional environment. They acted as important fluid flow pathways for uranium to migrate not only for surface-derived oxygenated fluids, but also for high oxidized metamorphic and magmatic fluids, dominated by their geological environment in which the unconformities occur. We have carried out comprehensive empirical spatial analyses of various types of uranium deposits in Australia, and first results indicate that there is a strong spatial correlation between unconformities and uranium deposits, not only for traditional unconformity-related deposits but also for other styles. As a start we analysed uranium deposits in Queensland and in particular Proterozoic metasomatic-related deposits in the Mount Isa Inlier and Late Carboniferous to Early Permian volcanic-hosted uranium occurrences in Georgetown and Charters Towers Regions show strong spatial associations with contemporary and older unconformities. The Georgetown Inlier in northern Queensland consists of a diverse range of rocks, including Proterozoic and early Palaeozoic metamorphic rocks and granites and late Palaeozoic volcanic rocks and related granites. Uranium-molybdenum (+/- fluorine) mineralization in the Georgetown inlier varies from strata- to structure-bound and occurs above regional unconformities. The Proterozoic basins in the Mount Isa Inlier rest unconformably on Palaeoproterozoic basement accompanied by volcanic and igneous rocks, which were deformed and metamorphosed in the Mesoproterozoic. Uranium occurrences in the Western Succession of Mount Isa are either hosted in clastic metasediments or mafic volcanics that belong to the Palaeoproterozoic Eastern Creek Volcanics. Uranium and vanadium mineralization occur in metasomatised and hematite-magnetite-carbonate alteration zones, bounded by major faults and regional unconformities. The results of this study highlight the importance of unconformities in uranium minerals systems as possible fluid pathways and/or surfaces of physico-chemical contrast that could have facilitated the precipitation of uranium, not only in classical unconformity style uranium deposits but in several other styles of uranium mineralization as well. References Cuney, M., 2009. The extreme diversity of uranium deposits. Mineralium Deposita, 44, 3-9. Dahlkamp, F. J., 1993. Uranium ore deposits. Springer, Berlin, p 460. OECD / NEA Red Book & IAEA, 2000. Uranium 1999: Resources, Production and Demand. OECD Nuclear Energy Agency and International Atomic Energy Agency, Paris.

Monitoring genotoxic exposure in uranium mines

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sram, R.J.; Vesela, D.; Vesely, D.

1993-10-01

Recent data from deep uranium mines in Czechoslovakia indicated that miners are exposed to other mutagenic factors in addition to radon daughter products. Mycotoxins were identified as a possible source of mutagens in these mines. Mycotoxins were examined in 38 samples from mines and in throat swabs taken from 116 miners and 78 controls. The following mycotoxins were identified from mines samples: aflatoxins B{sub 1} and G1, citrinin, citreoviridin, mycophenolic acid, and sterigmatocystin. Some mold strains isolated from mines and throat swabs were investigated for mutagenic activity by the SOS chromotest and Salmonella assay with strains TA100 and TA98. Mutagenicitymore » was observed, especially with metabolic activation in citro. These data suggest that mycotoxins produced by molds in uranium mines are a new genotoxic factor im uranium miners. 17 refs., 4 tabs.« less

Uptake of Uranium from Seawater by Amidoxime-Based Polymeric Adsorbent: Field Experiments, Modeling, and Updated Economic Assessment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kim, Jungseung; Tsouris, Constantinos; Oyola, Yatsandra

2014-04-09

Uranium recovery from seawater has been investigated for several decades for the purpose of securing nuclear fuel for energy production. In this study, field column experiments have been performed at the Marine Sciences Laboratory of the Pacific Northwest National Laboratory (PNNL) using a laboratory-proven, amidoxime-based polymeric adsorbent developed at the Oak Ridge National Laboratory (ORNL). The adsorbent was packed either in in-line filters or in flow-through columns. The maximum amount of uranium uptake from seawater was 3.3 mg of U/g of adsorbent after 8 weeks of contact between the adsorbent and seawater. This uranium adsorption amount was about 3 timesmore » higher than the maximum amount achieved in this study by a leading adsorbent developed at the Japan Atomic Energy Agency (JAEA).« less

Electrochemical and Infrared Absorption Spectroscopy Detection of SF₆ Decomposition Products.

PubMed

Dong, Ming; Zhang, Chongxing; Ren, Ming; Albarracín, Ricardo; Ye, Rixin

2017-11-15

Sulfur hexafluoride (SF₆) gas-insulated electrical equipment is widely used in high-voltage (HV) and extra-high-voltage (EHV) power systems. Partial discharge (PD) and local heating can occur in the electrical equipment because of insulation faults, which results in SF₆ decomposition and ultimately generates several types of decomposition products. These SF₆ decomposition products can be qualitatively and quantitatively detected with relevant detection methods, and such detection contributes to diagnosing the internal faults and evaluating the security risks of the equipment. At present, multiple detection methods exist for analyzing the SF₆ decomposition products, and electrochemical sensing (ES) and infrared (IR) spectroscopy are well suited for application in online detection. In this study, the combination of ES with IR spectroscopy is used to detect SF₆ gas decomposition. First, the characteristics of these two detection methods are studied, and the data analysis matrix is established. Then, a qualitative and quantitative analysis ES-IR model is established by adopting a two-step approach. A SF₆ decomposition detector is designed and manufactured by combining an electrochemical sensor and IR spectroscopy technology. The detector is used to detect SF₆ gas decomposition and is verified to reliably and accurately detect the gas components and concentrations.

Uranium in Holocene valley-fill sediments, and uranium, radon, and helium in waters, Lake Tahoe-Carson Range area, Nevada and California, U.S.A.

USGS Publications Warehouse

Otton, J.K.; Zielinski, R.A.; Been, J.M.

1989-01-01

Uraniferous Holocene sediments occur in the Carson Range of Nevada and California, U.S.A., between Lake Tahoe and Carson Valley. The hosts for the uranium include peat and interbedded organic-rich sand, silt, and mud that underly valley floors, fens, and marshes along stream valleys between the crest of the range and the edge of Lake Tahoe. The known uranium accumulations extend along the Carson Range from the area just southeast of South Lake Tahoe northward to the area just east of Carson City; however, they almost certainly continue beyond the study area to the north, west, and south. Due to the young age of the accumulations, uranium in them is in gross disequilibrium with its highly radioactive daughter products. These accumulations have thus escaped discovery with radiation detection equipment in the past. The uranium content of these sediments approaches 0.6 percent; however, the average is in the range of 300-500 ppm. Waters associated with these sediments locally contain as much as 177 ppb uranium. Modest levels of helium and radon also occur in these waters. Uraniferous waters are clearly entering the private and public water supply systems in some parts of the study area; however, it is not known how much uranium is reaching users of these water supplies. Many of the waters sampled in the study area exceed the published health effects guidance level of the Environmental Protection Agency. Regulatory standards for uranium in waters have not been published, however. Much uranium is stored in the sediments along these stream valleys. Estimates for a marsh and a fen along one drainage are 24,000 and 15,000 kg, respectively. The potential effects of man-induced environmental changes on the uranium are uncertain. Laboratory studies of uraniferous sediment rich in organic matter may allow us to evaluate the potential of liberating uranium from such sediments and creating transient increases in the level of uranium moving in water in the natural environment. ?? 1989 Springer-Verlag New York Inc.

PROCESS FOR THE CONCENTRATION OF ORES CONTAINING GOLD AND URANIUM

DOEpatents

Gaudin, A.M.; Dasher, J.

1958-06-10

ABS>A process is described for concentrating certain low grade uranium and gold bearing ores, in which the gangue is mainly quartz. The production of the concentrate is accomplished by subjecting the crushed ore to a froth floatation process using a fatty acid as a collector in conjunction with a potassium amyl xanthate collector. Pine oil is used as the frothing agent.

RUM JUNGLE PROJECT

DOE Office of Scientific and Technical Information (OSTI.GOV)

Australia. Atomic Energy Commission

1963-01-01

A report is presented on the uranium mining and treatment industry established at Rum Jungle and its contribution to the development of the Northern Territory. The Combined Development Agency contract for uranium procurement (terminated in 1963) and some of its results are described. A description of Rum Jungle and its geology and mineralization is given. Mining and treatment of ore are discussed, and some production statistics are given. (D.L.C.)

In-situ, time resolved monitoring of uranium in BFS:OPC grout. Part 1: Corrosion in water vapour.

PubMed

Stitt, C A; Paraskevoulakos, C; Banos, A; Harker, N J; Hallam, K R; Davenport, A; Street, S; Scott, T B

2017-08-11

Uranium encapsulated in grout was exposed to water vapour for extended periods of time. Through synchrotron x-ray powder diffraction and tomography measurements, uranium dioxide was determined the dominant corrosion product over a 50-week time period. The oxide growth rate initiated rapidly, with rates comparable to the U + H 2 O reaction. Over time, the reaction rate decreased and eventually plateaued to a rate similar to the U + H 2 O + O 2 reaction. This behaviour was not attributed to oxygen ingress, but instead the decreasing permeability of the grout, limiting oxidising species access to the metal surface.

THORIUM OXALATE-URANYL ACETATE COUPLED PROCEDURE FOR THE SEPARATION OF RADIOACTIVE MATERIALS

DOEpatents

Gofman, J.W.

1959-08-11

The recovery of fission products from neutronirradiated uranium is described. The neutron-irradiated uranium is dissolved in acid and thorium oxalate is precipitated in ihe solution formed, whereby the fission products are carried on the thorium oxalate. The separated thorium oxalate precipitate is then dissolved in an aqueous oxalate solution and the solution formed is acidified, limiting ihe excess acidity to a maximum of 2 N, whereby thorium oxalate precipitates and carries lanthanum-rareearth- and alkaline-earth-metal fission products while the zirconium-fission-product remains in solution. This precipitate, too, is dissolved in an aqaeous oxalate solution at elevated temperature, and lanthanum-rare-earth ions are added to the solution whereby lanthanum-rare-earth oxalate forms and the lanthanum-rare-earth-type and alkalineearth-metal-type fission products are carried on the oxalate. The precipitate is separated from the solution.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bakel, Allen J.; Conner, Cliff; Quigley, Kevin

One of the missions of the Reduced Enrichment for Research and Test Reactors (RERTR) program (and now the National Nuclear Security Administrations Material Management and Minimization program) is to facilitate the use of low enriched uranium (LEU) targets for 99Mo production. The conversion from highly enriched uranium (HEU) to LEU targets will require five to six times more uranium to produce an equivalent amount of 99Mo. The work discussed here addresses the technical challenges encountered in the treatment of uranyl nitrate hexahydrate (UNH)/nitric acid solutions remaining after the dissolution of LEU targets. Specifically, the focus of this work is themore » calcination of the uranium waste from 99Mo production using LEU foil targets and the Modified Cintichem Process. Work with our calciner system showed that high furnace temperature, a large vent tube, and a mechanical shield are beneficial for calciner operation. One- and two-step direct calcination processes were evaluated. The high-temperature one-step process led to contamination of the calciner system. The two-step direct calcination process operated stably and resulted in a relatively large amount of material in the calciner cup. Chemically assisted calcination using peroxide was rejected for further work due to the difficulty in handling the products. Chemically assisted calcination using formic acid was rejected due to unstable operation. Chemically assisted calcination using oxalic acid was recommended, although a better understanding of its chemistry is needed. Overall, this work showed that the two-step direct calcination and the in-cup oxalic acid processes are the best approaches for the treatment of the UNH/nitric acid waste solutions remaining from dissolution of LEU targets for 99Mo production.« less

Diversification in the Supply Chain of (99)Mo Ensures a Future for (99m)Tc.

PubMed

Cutler, Cathy S; Schwarz, Sally W

2014-07-01

The uncertain availability of (99m)Tc has become a concern for nuclear medicine departments across the globe. An issue for the United States is that currently it is dependent on a supply of (99m)Tc (from (99)Mo) that is derived solely by production outside the United States. Since the United States uses half the world's (99)Mo production, the U.S. (99)Mo supply chain would be greatly enhanced if a producer were located within the United States. The fragility of the old (99)Mo supply chain is being addressed as new facilities are constructed and new processes are developed to produce (99)Mo without highly enriched uranium. The conversion to low-enriched uranium is necessary to minimize the potential misuse of highly enriched uranium in the world for nonpeaceful means. New production facilities, new methods for the production of (99)Mo, and a new generator elution system for the supply of (99m)Tc are currently being pursued. The progress made in all these areas will be discussed, as they all highlight the need to embrace diversity to ensure that we have a robust and reliable supply of (99m)Tc in the future. © 2014 by the Society of Nuclear Medicine and Molecular Imaging, Inc.

Mechanistic approach for nitride fuel evolution and fission product release under irradiation

NASA Astrophysics Data System (ADS)

Dolgodvorov, A. P.; Ozrin, V. D.

2017-01-01

A model for describing uranium-plutonium mixed nitride fuel pellet burning was developed. Except fission products generating, the model includes impurities of oxygen and carbon. Nitrogen behaviour in nitride fuel was analysed and the nitrogen chemical potential in solid solution with uranium-plutonium nitride was constructed. The chemical program module was tested with the help of thermodynamic equilibrium phase distribution calculation. Results were compared with analogous data in literature, quite good agreement was achieved, especially for uranium sesquinitride, metallic species and some oxides. Calculation of a process of nitride fuel burning was also conducted. Used mechanistic approaches for fission product evolution give the opportunity to find fission gas release fractions and also volumes of intergranular secondary phases. Calculations present that the most massive secondary phases are the oxide and metallic phases. Oxide phase contain approximately 1 % wt of substance over all time of burning with slightly increasing of content. Metallic phase has considerable rising of mass and by the last stage of burning it contains about 0.6 % wt of substance. Intermetallic phase has less increasing rate than metallic phase and include from 0.1 to 0.2 % wt over all time of burning. The highest element fractions of released gaseous fission products correspond to caesium and iodide.

Improving flow patterns and spillage characteristics of a box-type commercial kitchen hood.

PubMed

Huang, Rong Fung; Chen, Jia-Kun; Han, Meng-Ji; Priyambodo, Yusuf

2014-01-01

A conventional box-type commercial kitchen hood and its improved version (termed the "IQV commercial kitchen hood") were studied using the laser-assisted smoke flow visualization technique and tracer-gas (sulfur hexafluoride) detection methods. The laser-assisted smoke flow visualization technique qualitatively revealed the flow field of the hood and the areas apt for leakages of hood containment. The tracer-gas concentration detection method measured the quantitative leakage levels of the hood containment. The oil mists that were generated in the conventional box-type commercial kitchen hood leaked significantly into the environment from the areas near the front edges of ceiling and side walls. Around these areas, the boundary-layer separation occurred, inducing highly unsteady and turbulent recirculating flow, and leading to spillages of hood containment due to inappropriate aerodynamic design at the front edges of the ceiling and side walls. The tracer-gas concentration measurements on the conventional box-type commercial kitchen hood showed that the sulfur hexafluoride concentrations detected at the hood face attained very large values on an order of magnitude about 10(3)-10(4) ppb. By combining the backward-offset narrow suction slot, deflection plates, and quarter-circular arcs at the hood entrance, the IQV commercial kitchen hood presented a flow field containing four backward-inclined cyclone flow structures. The oil mists generated by cooking were coherently confined in these upward-rising cyclone flow structures and finally exhausted through the narrow suction slot. The tracer-gas concentration measurements on the IQV commercial kitchen hood showed that the order of magnitude of the sulfur hexafluoride concentrations detected at the hood face is negligibly small--only about 10(0) ppb across the whole hood face.

Depleted and natural uranium: chemistry and toxicological effects.

PubMed

Craft, Elena; Abu-Qare, Aquel; Flaherty, Meghan; Garofolo, Melissa; Rincavage, Heather; Abou-Donia, Mohamed

2004-01-01

Depleted uranium (DU) is a by-product from the chemical enrichment of naturally occurring uranium. Natural uranium is comprised of three radioactive isotopes: (238)U, (235)U, and (234)U. This enrichment process reduces the radioactivity of DU to roughly 30% of that of natural uranium. Nonmilitary uses of DU include counterweights in airplanes, shields against radiation in medical radiotherapy units and transport of radioactive isotopes. DU has also been used during wartime in heavy tank armor, armor-piercing bullets, and missiles, due to its desirable chemical properties coupled with its decreased radioactivity. DU weapons are used unreservedly by the armed forces. Chemically and toxicologically, DU behaves similarly to natural uranium metal. Although the effects of DU on human health are not easily discerned, they may be produced by both its chemical and radiological properties. DU can be toxic to many bodily systems, as presented in this review. Most importantly, normal functioning of the kidney, brain, liver, and heart can be affected by DU exposure. Numerous other systems can also be affected by DU exposure, and these are also reviewed. Despite the prevalence of DU usage in many applications, limited data exist regarding the toxicological consequences on human health. This review focuses on the chemistry, pharmacokinetics, and toxicological effects of depleted and natural uranium on several systems in the mammalian body. A section on risk assessment concludes the review.