Porous membrane electrochemical cell for uranium and transuranic recovery from molten salt electrolyte

DOEpatents

Willit, James L [Ratavia, IL

2007-09-11

An improved process and device for the recovery of the minor actinides and the transuranic elements (TRU's) from a molten salt electrolyte. The process involves placing the device, an electrically non-conducting barrier between an anode salt and a cathode salt. The porous barrier allows uranium to diffuse between the anode and cathode, yet slows the diffusion of uranium ions so as to cause depletion of uranium ions in the catholyte. This allows for the eventual preferential deposition of transuranics present in spent nuclear fuel such as Np, Pu, Am, Cm. The device also comprises an uranium oxidation anode. The oxidation anode is solid uranium metal in the form of spent nuclear fuel. The spent fuel is placed in a ferric metal anode basket which serves as the electrical lead or contact between the molten electrolyte and the anodic uranium metal.

Porous membrane electrochemical cell for uranium and transuranic recovery from molten salt electrolyte

DOEpatents

Willit, James L [Batavia, IL

2010-09-21

An improved process and device for the recovery of the minor actinides and the transuranic elements (TRU's) from a molten salt electrolyte. The process involves placing the device, an electrically non-conducting barrier between an anode salt and a cathode salt. The porous barrier allows uranium to diffuse between the anode and cathode, yet slows the diffusion of uranium ions so as to cause depletion of uranium ions in the catholyte. This allows for the eventual preferential deposition of transuranics present in spent nuclear fuel such as Np, Pu, Am, Cm. The device also comprises an uranium oxidation anode. The oxidation anode is solid uranium metal in the form of spent nuclear fuel. The spent fuel is placed in a ferric metal anode basket which serves as the electrical lead or contact between the molten electrolyte and the anodic uranium metal.

URANIUM RECOVERY PROCESS

DOEpatents

Yeager, J.H.

1958-08-12

In the prior art processing of uranium ores, the ore is flrst digested with nitric acid and filtered, and the uranium values are then extracted tom the filtrate by contacting with an organic solvent. The insoluble residue has been processed separately in order to recover any uranium which it might contain. The improvement consists in contacting a slurry, composed of both solution and residue, with the organic solvent prior to filtration. Tbe result is that uranium values contained in the residue are extracted along with the uranium values contained th the solution in one step.

RECOVERY OF URANIUM VALUES FROM RESIDUES

DOEpatents

Schaap, W.B.

1959-08-18

A process is described for the recovery of uranium from insoluble oxide residues resistant to repeated leaching with mineral acids. The residue is treated with gaseous hydrogen fluoride, then with hydrogen and again with hydrogen fluoride, preferably at 500 to 700 deg C, prior to the mineral acid leaching.

URANIUM LEACHING AND RECOVERY PROCESS

DOEpatents

McClaine, L.A.

1959-08-18

A process is described for recovering uranium from carbonate leach solutions by precipitating uranium as a mixed oxidation state compound. Uranium is recovered by adding a quadrivalent uranium carbon;te solution to the carbonate solution, adjusting the pH to 13 or greater, and precipitating the uranium as a filterable mixed oxidation state compound. In the event vanadium occurs with the uranium, the vanadium is unaffected by the uranium precipitation step and remains in the carbonate solution. The uranium-free solution is electrolyzed in the cathode compartment of a mercury cathode diaphragm cell to reduce and precipitate the vanadium.

RECOVERY OF URANIUM FROM CARBONATE LEACH LIQUORS

DOEpatents

Wilson, H.F.

1958-07-01

An improved process is described for the recovery of uranium from vanadifrous ores. In the prior art such ores have been digested with alkali carbonate solutions at a pH of less than 10 and then contacted with a strong base anion exchange resin to separate uranium from vanadium. It has been found that if the exchamge resin feed solution has its pH adjusted to the range 10.8 to 11.8, that vanadium adsorption on the resin is markedly decreased and the separation of uranium from the vanadium is thereby improved.

RECOVERY OF URANIUM VALUES FROM URANIUM BEARING RAW MATERIALS

DOEpatents

Michal, E.J.; Porter, R.R.

1959-06-16

Uranium leaching from ground uranium-bearing raw materials using MnO/sub 2/ in H/sub 2/SO/sub 4/ is described. The MnO/sub 2/ oxidizes U to the leachable hexavalent state. The MnO/sub 2/ does not replace Fe normally added, because the Fe complexes P and catalyzes the MnO/sub 2/ reaction. Three examples of continuous processes are given, but batch operation is also possible. The use of MnO/sub 2/ makes possible recovery of very low U values. (T.R.H.)

RECOVERY OF URANIUM AND THORIUM FROM AQUEOUS SOLUTIONS

DOEpatents

Calkins, G.D.

1958-06-10

>A process is described for the recovery of uranium and thorium from monazite sand, which is frequently processed by treating it with a hot sodium hydroxide solution whereby a precipitate forms consisting mainly of oxides or hydroxides of the rare earths, thorium and uranium. The precipitate is dissolved in mineral acid, and the acid solution is then neutralized to a pH value of between 5.2 and 6.2 whereby both the uranium and thorium precipitate as the hydroxides, while substantially all the rare earth metal values present remain in the solution. The uranium and thoriunn can then be separated by dissolving the precipitate in a solution containing a mixture of alkali carbonate and alkali bicarbonate: and contacting the carbonate solution with a strong-base anion exchange resin whereby the uranium values are adsorbed on the resin while the thorium remains in solution.

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.

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.

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.

Method of separating and recovering uranium and related cations from spent Purex-type systems

DOEpatents

Mailen, J.C.; Tallent, O.K.

1987-02-25

A process for separating uranium and related cations from a spent Purex-type solvent extraction system which contains degradation complexes of tributylphosphate wherein the system is subjected to an ion-exchange process prior to a sodium carbonate scrubbing step. A further embodiment comprises recovery of the separated uranium and related cations. 5 figs.

CATALYZED OXIDATION OF URANIUM IN CARBONATE SOLUTIONS

DOEpatents

Clifford, W.E.

1962-05-29

A process is given wherein carbonate solutions are employed to leach uranium from ores and the like containing lower valent uranium species by utilizing catalytic amounts of copper in the presence of ammonia therein and simultaneously supplying an oxidizing agent thereto. The catalysis accelerates rate of dissolution and increases recovery of uranium from the ore. (AEC)

Recovery of uranium from seawater by immobilized tannin

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

Sakaguchi, T.; Nakajima, A.

1987-06-01

Tannin compounds having multiple adjacent hydroxy groups have an extremely high affinity for uranium. To prevent the leaching of tannins into water and to improve the adsorbing characteristics of these compounds, the authors tried to immobilize tannins. The immobilized tannin has the most favorable features for uranium recovery; high selective adsorption ability to uranium, rapid adsorption rate, and applicability in both column and batch systems. The immobilized tannin can recover uranium from natural seawater with high efficiency. About 2530 ..mu..g uranium is adsorbed per gram of this adsorbent within 22 h. Depending on the concentration in seawater, an enrichment ofmore » up to 766,000-fold within the adsorbent is possible. Almost all uranium adsorbed is easily desorbed with a very dilute acid. Thus, the immobilized tannin can be used repeatedly in the adsorption-desorption process.« 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.

Recovery of Uranium from Wet Phosphoric Acid by Solvent Extraction Processes

DOE PAGES

Beltrami, Denis; Cote, Gérard; Mokhtari, Hamid; ...

2014-11-17

Between 1951 and 1991, we developed about 17 processes to recover uranium from wet phosphoric acid (WPA), but the viability of these processes was subject to the variation of the uranium price market. Nowadays, uranium from WPA appears to be attractive due to the increase of the global uranium demand resulting from the emergence of developing countries. Moreover, the increasing demand provides impetus for a new look at the applicable technology with a view to improvements as well as altogether new approaches. This paper gives an overview on extraction processes developed in the past to recover uranium from wet phosphoricmore » acid (WPA) as well as the physicochemistry involved in these processes. Recent advances concerning the development of new extraction systems are also reported and discussed.« 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.

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.

RECOVERY OF URANIUM FROM PITCHBLENDE

DOEpatents

Ruehle, A.E.

1958-06-24

The decontamination of uranium from molybdenum is described. When acid solutions containing uranyl nitrate are contacted with ether for the purpose of extracting the uranium values, complex molybdenum compounds are coextracted with the uranium and also again back-extracted from the ether with the uranium. This invention provides a process for extracting uranium in which coextraction of molybdenum is avoided. It has been found that polyhydric alcohols form complexes with molybdenum which are preferentially water-soluble are taken up by the ether extractant to only a very minor degree. The preferred embodiment of the process uses mannitol, sorbitol or a mixture of the two as the complexing agent.

SLURRY SOLVENT EXTRACTION PROCESS FOR THE RECOVERY OF METALS FROM SOLID MATERIALS

DOEpatents

Grinstead, R.R.

1959-01-20

A solvent extraction process is described for recovering uranium from low grade uranium bearing minerals such as carnotit or shale. The finely communited ore is made up as an aqueous slurry containing the necessary amount of acid to solubilize the uranium and simultaneously or subsequently contacted with an organic solvent extractant such as the alkyl ortho-, or pyro phosphoric acids, alkyl phosphites or alkyl phosphonates in combination with a diluent such as kerosene or carbon tetrachlorids. The extractant phase is separated from the slurry and treated by any suitable process to recover the uranium therefrom. One method for recovering the uranium comprises treating the extract with aqueous HF containing a reducing agent such as ferrous sulfate, which reduces the uranium and causes it to be precipitated as uranium tetrafluoride.

RECOVERY OF URANIUM FROM ZIRCONIUM-URANIUM NUCLEAR FUELS

DOEpatents

Gens, T.A.

1962-07-10

An improvement was made in a process of recovering uranium from a uranium-zirconium composition which was hydrochlorinated with gsseous hydrogen chloride at a temperature of from 350 to 800 deg C resulting in volatilization of the zirconium, as zirconium tetrachloride, and the formation of a uranium containing nitric acid insoluble residue. The improvement consists of reacting the nitric acid insoluble hydrochlorination residue with gaseous carbon tetrachloride at a temperature in the range 550 to 600 deg C, and thereafter recovering the resulting uranium chloride vapors. (AEC)

URANIUM RECOVERY PROCESS

DOEpatents

Kaufman, D.

1958-04-15

A process of recovering uranium from very low-grade ore residues is described. These low-grade uraniumcontaining hydroxide precipitates, which also contain hydrated silica and iron and aluminum hydroxides, are subjected to multiple leachings with aqueous solutions of sodium carbonate at a pH of at least 9. This leaching serves to selectively extract the uranium from the precipitate, but to leave the greater part of the silica, iron, and aluminum with the residue. The uranium is then separated from the leach liquor by the addition of an acid in sufficient amount to destroy the carbonate followed by the addition of ammonia to precipitate uranium as ammonium diuranate.

RECOVERY OF URANIUM BY AROMATIC DITHIOCARBAMATE COMPLEXING

DOEpatents

Neville, O.K.

1959-08-11

A selective complexing organic solvent extraction process is presented for the separation of uranium values from an aqueous nitric acid solution of neutron irradiated thorium. The process comprises contacting the solution with an organic aromatic dithiccarbamaie and recovering the resulting urancdithiccarbamate complex with an organic solvent such as ethyl acetate.

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.

Removal and recovery of uranium(VI) by waste digested activated sludge in fed-batch stirred tank reactor.

PubMed

Jain, Rohan; Peräniemi, Sirpa; Jordan, Norbert; Vogel, Manja; Weiss, Stephan; Foerstendorf, Harald; Lakaniemi, Aino-Maija

2018-05-24

This study demonstrated the removal and recovery of uranium(VI) in a fed-batch stirred tank reactor (STR) using waste digested activated sludge (WDAS). The batch adsorption experiments showed that WDAS can adsorb 200 (±9.0) mg of uranium(VI) per g of WDAS. The maximum adsorption of uranium(VI) was achieved even at an acidic initial pH of 2.7 which increased to a pH of 4.0 in the equilibrium state. Desorption of uranium(VI) from WDAS was successfully demonstrated from the release of more than 95% of uranium(VI) using both acidic (0.5 M HCl) and alkaline (1.0 M Na 2 CO 3 ) eluents. Due to the fast kinetics of uranium(VI) adsorption onto WDAS, the fed-batch STR was successfully operated at a mixing time of 15 min. Twelve consecutive uranium(VI) adsorption steps with an average adsorption efficiency of 91.5% required only two desorption steps to elute more than 95% of uranium(VI) from WDAS. Uranium(VI) was shown to interact predominantly with the phosphoryl and carboxyl groups of the WDAS, as revealed by in situ infrared spectroscopy and time-resolved laser-induced fluorescence spectroscopy studies. This study provides a proof-of-concept of the use of fed-batch STR process based on WDAS for the removal and recovery of uranium(VI). Copyright © 2018 Elsevier Ltd. All rights reserved.

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.

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.

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.

76 FR 41308 - Strata Energy, Inc., Ross In Situ Recovery Uranium Project, Crook County, WY; Notice of Materials...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-07-13

..., Inc., Ross In Situ Recovery Uranium Project, Crook County, WY; Notice of Materials License Application...-4737, or by e-mail to [email protected] . The Ross In Situ Recovery Uranium Project License... source and byproduct materials license at its Ross In Situ Recovery Uranium Project site located in Crook...

URANIUM PRECIPITATION PROCESS

DOEpatents

Thunaes, A.; Brown, E.A.; Smith, H.W.; Simard, R.

1957-12-01

A method for the recovery of uranium from sulfuric acid solutions is described. In the present process, sulfuric acid is added to the uranium bearing solution to bring the pH to between 1 and 1.8, preferably to about 1.4, and aluminum metal is then used as a reducing agent to convert hexavalent uranium to the tetravalent state. As the reaction proceeds, the pH rises amd a selective precipitation of uranium occurs resulting in a high grade precipitate. This process is an improvement over the process using metallic iron, in that metallic aluminum reacts less readily than metallic iron with sulfuric acid, thus avoiding consumption of the reducing agent and a raising of the pH without accomplishing the desired reduction of the hexavalent uranium in the solution. Another disadvantage to the use of iron is that positive ferric ions will precipitate with negative phosphate and arsenate ions at the pH range employed.

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.

Inherently safe in situ uranium recovery

DOEpatents

Krumhansl, James L; Brady, Patrick V

2014-04-29

An in situ recovery of uranium operation involves circulating reactive fluids through an underground uranium deposit. These fluids contain chemicals that dissolve the uranium ore. Uranium is recovered from the fluids after they are pumped back to the surface. Chemicals used to accomplish this include complexing agents that are organic, readily degradable, and/or have a predictable lifetime in an aquifer. Efficiency is increased through development of organic agents targeted to complexing tetravalent uranium rather than hexavalent uranium. The operation provides for in situ immobilization of some oxy-anion pollutants under oxidizing conditions as well as reducing conditions. The operation also artificially reestablishes reducing conditions on the aquifer after uranium recovery is completed. With the ability to have the impacted aquifer reliably remediated, the uranium recovery operation can be considered inherently safe.

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.

Recovery of rare earth elements (REEs) from uranium containing solutions using biosorption

NASA Astrophysics Data System (ADS)

Botalov, Maxim; Bunkov, Grigory; Rychkov, Vladimir; Kirillov, Evgeny; Kirillov, Sergey; Kuchta, Kerstin; Kücüker, Mehmet Ali; Atamaniuk, Iryna

2017-09-01

Batch testing and determination of appropriate biosorbent and experimental procedures for recovery of REEs from artificial solution as well as the efficiency of the process for recovery of REEs from artificial solution via biosorption are given in present research work.

76 FR 72006 - Draft Interim Staff Guidance: Evaluations of Uranium Recovery Facility Surveys of Radon and Radon...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-11-21

.... Discussion Uranium recovery facility licensees, including in-situ recovery facilities and conventional... Recovery Facility Surveys of Radon and Radon Progeny in Air and Demonstrations of Compliance AGENCY... Staff Guidance, ``Evaluations of Uranium Recovery Facility Surveys of Radon and Radon Progeny in Air and...

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.

Surface Functionalized Nanostructured Ceramic Sorbents for the Effective Collection and Recovery of Uranium from Seawater

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

Chouyyok, Wilaiwan; Pittman, Jonathan W.; Warner, Marvin G.

2016-05-02

The ability to collect uranium from seawater offers the potential for a nearly limitless fuel supply for nuclear energy. We evaluated the use of functionalized nanostructured sorbents for the collection and recovery of uranium from seawater. Extraction of trace minerals from seawater and brines is challenging due to the high ionic strength of seawater, low mineral concentrations, and fouling of surfaces over time. We demonstrate that rationally assembled sorbent materials that integrate high affinity surface chemistry and high surface area nanostructures into an application relevant micro/macro structure enables collection performance that far exceeds typical sorbent materials. High surface area nanostructuredmore » silica with surface chemistries composed of phosphonic acid, phosphonates, 3,4 hydroxypyridinone, and EDTA showed superior performance for uranium collection. A few phosphorous-based commercial resins, specifically Diphonix and Ln Resin, also performed well. We demonstrate an effective and environmentally benign method of stripping the uranium from the high affinity sorbents using inexpensive nontoxic carbonate solutions. The cyclic use of preferred sorbents and acidic reconditioning of materials was shown to improve performance. Composite thin films composed of the nanostructured sorbents and a porous polymer binder are shown to have excellent kinetics and good capacity while providing an effective processing configuration for trace mineral recovery from solutions. Initial work using the composite thin films shows significant improvements in processing capacity over the previously reported sorbent materials.« less

FY-15 Progress Report on Cleanup of irradiated SHINE Target Solutions Containing 140g-U/L Uranyl Sulfate

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

Bennett, Megan E.; Bowers, Delbert L.; Vandegrift, George F.

2015-09-01

During FY 2012 and 2013, a process was developed to convert the SHINE Target Solution (STS) of irradiated uranyl sulfate (140 g U/L) to uranyl nitrate. This process is necessary so that the uranium solution can be processed by the UREX (Uranium Extraction) separation process, which will remove impurities from the uranium so that it can be recycled. The uranyl sulfate solution must contain <0.02 M SO 4 2- so that the uranium will be extractable into the UREXsolvent. In addition, it is desired that the barium content be below 0.0007 M, as this is the limit in the Resourcemore » Conservation and Recovery Act (RCRA).« less

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...

78 FR 19330 - Supplemental Environmental Impact Statement for the Ross In-Situ Uranium Recovery Project in...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-03-29

... Ross In-Situ Uranium Recovery Project in Crook County, Wyoming AGENCY: Nuclear Regulatory Commission... Commission (NRC) for a new source materials license for the proposed Ross In-Situ Uranium Recovery (ISR... SEIS is Supplement 5 to NUREG-1910, ``Generic Environmental Impact Statement for In-Situ Leach Uranium...

Materials. Section 1 of Symposium on the peaceful uses of atomic energy in Australia, 1958, held in Sydney, in June 1958

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

None

The environments of the known uranium occurences in South Australia arc described, and the relation of uranium mineralization with sodic granitic rocks is emphasized. The problems in designing equipment for radiometric prospecting are reviewed. The fabrication and properties of BeO, UO/sub 2/, ThO/sub 2/, and mixed oxides are discussed. The use of pulsing in a uranium extraction pilot plant ion exchange column is described. The wetting of metals by liquid metals is reviewed with emphasis on liquid sodium. The geological nature, extent, and future prospects of minerals with atomic energy applications, occurring in New South Wales are outlined. The developmentmore » of a process for uranium recovery from Mary Kathleen ores is described. Techniques and processes involved in locating, mining, and concentrating davidite-type ores at Radium Hill, South Australia are described. The uranium deposits of the Northern Territory, Australia, are classified and described. The flotation behavior of the simple oxide minerals, uraninite and the colloform variety is discussed. The Port Pirie Treatment Plant for uranium recovery from refractory Radium Hill concentrates is described. The plant utilizes the sulfuric acid-ion exchange process. The uranium deposits of Queensland are described. the details of the production of uranium ore concentrates at Rum jungle near Darwin, Australia, are given. A brief account of the use of neutron diffraction analysis in crystallography is given, and the neutron spectrometers installed on the High Flux Australian Research Reactor are described. (T.R.H.)« less

PROCESS FOR THE RECOVERY AND PURIFICATION OF URANIUM DEPOSITS

DOEpatents

Carter, J.M.; Kamen, M.D.

1958-10-14

A process is presented for recovering uranium values from UCl/sub 4/ deposits formed on calutrons. Such deposits are removed from the calutron parts by an aqueous wash solution which then contains the uranium values in addition to the following impurities: Ni, Cu, Fe, and Cr. This impurity bearing wash solution is treated with an oxidizing agent, and the oxidized solution is then treated with ammonia in order to precipitate the uranium as ammonium diuranate. The metal impurities of iron and chromium, which form insoluble hydroxides, are precipitated along with the uranium values. The precipitate is separated from the solution, dissolved in acid, and the solution again treated with ammonia and ammonium carbonate, which results in the precipitation of the metal impurities as hydroxides while the uranium values remain in solution.

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.

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

METHOD OF RECOVERING URANIUM COMPOUNDS

DOEpatents

Poirier, R.H.

1957-10-29

S>The recovery of uranium compounds which have been adsorbed on anion exchange resins is discussed. The uranium and thorium-containing residues from monazite processed by alkali hydroxide are separated from solution, and leached with an alkali metal carbonate solution, whereby the uranium and thorium hydrorides are dissolved. The carbonate solution is then passed over an anion exchange resin causing the uranium to be adsorbed while the thorium remains in solution. The uranium may be recovered by contacting the uranium-holding resin with an aqueous ammonium carbonate solution whereby the uranium values are eluted from the resin and then heating the eluate whereby carbon dioxide and ammonia are given off, the pH value of the solution is lowered, and the uranium is precipitated.

Natural thorium resources and recovery: Options and impacts

USGS Publications Warehouse

Ault, Timothy; Van Gosen, Bradley S.; Krahn, Steven; Croff, Allen

2016-01-01

This paper reviews the front end of the thorium fuel cycle, including the extent and variety of thorium deposits, the potential sources of thorium production, and the physical and chemical technologies required to isolate and purify thorium. Thorium is frequently found within rare earth element–bearing minerals that exist in diverse types of mineral deposits, often in conjunction with other minerals mined for their commercial value. It may be possible to recover substantial quantities of thorium as a by-product from active titanium, uranium, tin, iron, and rare earth mines. Incremental physical and chemical processing is required to obtain a purified thorium product from thorium minerals, but documented experience with these processes is extensive, and incorporating thorium recovery should not be overly challenging. The anticipated environmental impacts of by-product thorium recovery are small relative to those of uranium recovery since existing mining infrastructure utilization avoids the opening and operation of new mines and thorium recovery removes radionuclides from the mining tailings.

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.

Nuclear Weapons. National Nuclear Security Administration’s Plans for Its Uranium Processing Facility Should Better Reflect Funding Estimates and Technology Readiness

DTIC Science & Technology

2010-11-01

metal. Recovery extraction centrifugal contactors A process that uses solvent to extract uranium for purposes of purification. Agile machining A...extraction centrifugal contactors 5 6 Yes 6 No Agile machining 5 5 No 6 No Chip management 5 6 Yes 6 No Special casting 3 6 Yes 6 No Source: GAO

ALKYL PYROPHOSPHATE METAL SOLVENT EXTRACTANTS AND PROCESS

DOEpatents

Long, R.L.

1958-09-30

A process is presented for the recovery of uranium from aqueous mineral acidic solutions by solvent extraction. The extractant is a synmmetrical dialkyl pyrophosphate in which the alkyl substituents have a chain length of from 4 to 17 carbon atoms. Mentioned as a preferred extractant is dioctyl pyrophosphate. The uranium is precipitated irom the organic extractant phase with an agent such as HF, fluoride salts. alcohol, or ammonia.

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

78 FR 17450 - Notice of Issuance of Materials License Renewal, Operating License SUA-1341, Uranium One USA, Inc...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-03-21

... License Renewal, Operating License SUA-1341, Uranium One USA, Inc., Willow Creek Uranium In Situ Recovery.... SUA- 1341 to Uranium One USA, Inc. (Uranium One) for its Willow Creek Uranium In Situ Recovery (ISR) Project in Johnson and Campbell Counties, Wyoming. ADDRESSES: Please refer to Docket ID NRC-2009-0036 when...

PROCESS FOR RECOVERY OF URANIUM VALUES FROM IMPURE SOLUTIONS THEREOF

DOEpatents

Kilner, S.B.

1959-11-01

A process is presented for the recovery of uraninm values from impure solutions which are obtained, for example, by washing residual uranium salt or uranium metal deposits from stainless steel surfaces using an aqueous or certain acidic aqueous solutions. The solutions include uranyl and oxidized iron, chromium, nickel, and copper ions and may contain manganese, zinc, and silver ions. In accordance with one procedure. the uranyl ions are reduced to the uranous state, and the impurity ions are complexed with cyanide under acidic conditions. The solution is then treated with ammonium hydroxide or alkali metal hydroxide to precipitate uranous hydroxide away from the complexed impurity ions in the solution. Alternatively, an excess of alkali metal cyanide is added to the reduced solution until the solution becomes sufficiently alkaline for the uranons hydroxide to precipitate. An essential feature in operating the process is in maintaining the pH of the solution sufficiently acid during the complexing operation to prevent the precipitation of the impurity metal hydroxides.

Recovery of tritium from tritiated molecules

DOEpatents

Swansiger, W.A.

1984-10-17

This invention relates to the recovery of tritium from various tritiated molecules by reaction with uranium. More particularly, the invention relates to the recovery of tritium from tritiated molecules by reaction with uranium wherein the reaction is conducted in a reactor which permits the reaction to occur as a moving front reaction from the point where the tritium enters the reactor charged with uranium down the reactor until the uranium is exhausted.

Looking North at Uranium recovery Recycle Tanks in Red Room ...

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

Looking North at Uranium recovery Recycle Tanks in Red Room in Recycle Recovery Building - Hematite Fuel Fabrication Facility, Recycle Recovery Building, 3300 State Road P, Festus, Jefferson County, MO

U.sup.+4 generation in HTER

DOEpatents

Miller, William E [Naperville, IL; Gay, Eddie C [Park Forest, IL; Tomczuk, Zygmunt [Homer Glen, IL

2006-03-14

A improved device and process for recycling spent nuclear fuels, in particular uranium metal, that facilitates the refinement and recovery of uranium metal from spent metallic nuclear fuels. The electrorefiner device comprises two anodes in predetermined spatial relation to a cathode. The anodese have separate current and voltage controls. A much higher voltage than normal for the electrorefining process is applied to the second anode, thereby facilitating oxidization of uranium (III), U.sup.+, to uranium (IV), U.sup.+4. The current path from the second anode to the cathode is physically shorter than the similar current path from the second anode to the spent nuclear fuel contained in a first anode shaped as a basket. The resulting U.sup.+4 oxidizes and solubilizes rough uranium deposited on the surface of the cathode. A softer uranium metal surface is left on the cathode and is more readily removed by a scraper.

75 FR 71677 - Reimbursement for Costs of Remedial Action at Active Uranium and Thorium Processing Sites

Federal Register 2010, 2011, 2012, 2013, 2014

2010-11-24

... fiscal year (FY) 2011. SUMMARY: This Notice announces the Department of Energy (DOE) acceptance of claims... the American Recovery and Reinvestment Act of 2009 (Recovery Act). In addition, Congress provided $10... approximately $24.3 million of Recovery Act funds available for reimbursement in FY 2011, as well as the $10...

76 FR 50728 - Science Advisory Board Staff Office; Notification of Public Teleconferences of the Science...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-08-16

... advisory report related to uranium and thorium in-situ leach recovery and post-closure stability monitoring... . Technical Contact: Technical background information pertaining to the Uranium In-Situ leach recovery--Post... entitled ``Considerations Related to Post-Closure Monitoring of Uranium In-Situ Leach/In-Situ Recovery (ISL...

77 FR 70486 - Supplemental Environmental Impact Statement for Proposed Dewey-Burdock In-Situ Uranium Recovery...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-11-26

... Proposed Dewey- Burdock In-Situ Uranium Recovery Project in Custer and Fall River Counties, SD AGENCY... draft Supplemental Environmental Impact Statement (Draft SEIS) for the Dewey-Burdock In-Situ Uranium... NUREG-1910, ``Generic Environmental Impact Statement for In-Situ Leach Uranium Milling Facilities,'' May...

PROCESSING OF MONAZITE SAND

DOEpatents

Calkins, G.D.; Bohlmann, E.G.

1957-12-01

A process for the recovery of thorium, uranium, and rare earths from monazite sands is presented. The sands are first digested and dissolved in concentrated NaOH, and the solution is then diluted causing precipitation of uranium, thorium and rare earth hydroxides. The precipitate is collected and dissolved in HCl, and the pH of this solution is adjusted to about 6, precipitating the hydroxides of thorium and uranium but leaving the rare earths in solution. The rare earths are then separated from the solution by precipitation at a still higher pH. The thorium and uranium containing precipitate is redissolved in HNO/sub 3/ and the two elements are separated by extraction into tributyl phosphate and back extraction with a weakly acidic solution to remove the thorium.

77 FR 3460 - Reimbursement for Costs of Remedial Action at Active Uranium and Thorium Processing Sites

Federal Register 2010, 2011, 2012, 2013, 2014

2012-01-24

... Processing Sites AGENCY: Department of Energy. ACTION: Notice of the acceptance of Title X claims during... under Title X of the Energy Policy Act of 1992. In FY 2009, Congress appropriated $70 million for Title X in the American Recovery and Reinvestment Act of 2009 (Recovery Act). Also in FY 2009, Congress...

Raoultella sp. SM1, a novel iron-reducing and uranium-precipitating strain.

PubMed

Sklodowska, Aleksandra; Mielnicki, Sebastian; Drewniak, Lukasz

2018-03-01

The main aim of this study was the characterisation of novel Raoutella isolate, an iron-reducing and uranium-precipitating strain, originating from microbial mats occurring in the sediments of a closed down uranium mine in Kowary (SW Poland). Characterisation was done in the context of its potential role in the functioning of these mats and the possibility to use them in uranium removal/recovery processes. In our experiment, we observed the biological precipitation of iron and uranium's secondary minerals containing oxygen, potassium, sodium and phosphor, which were identified as ningyoite-like minerals. The isolated strain, Raoultella sp. SM1, was also able to dissimilatory reduce iron (III) and uranium (VI) in the presence of citrate as an electron donor. Our studies allowed us to characterise a new strain which may be used as a model microorganism in the study of Fe and U respiratory processes and which may be useful in the bioremediation of uranium-contaminated waters and sediments. During this process, uranium may be immobilised in ningyoite-like minerals and can then be recovered in nano/micro-particle form, which may be easily transformed to uraninite. Copyright © 2017 Elsevier Ltd. All rights reserved.

Experiments in anodic film effects during electrorefining of scrap U-10Mo fuels in support of modeling efforts

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

Van Kleeck, M.; Chemical Sciences and Engineering Division, Argonne National Laboratory, Argonne, IL 60439; Willit, J.

A monolithic uranium molybdenum alloy clad in zirconium has been proposed as a low enriched uranium (LEU) fuel option for research and test reactors, as part of the Reduced Enrichment for Research and Test Reactors program. Scrap from the fuel's manufacture will contain a significant portion of recoverable LEU. Pyroprocessing has been identified as an option to perform this recovery. A model of a pyroprocessing recovery procedure has been developed to assist in refining the LEU recovery process and designing the facility. Corrosion theory and a two mechanism transport model were implemented on a Mat-Lab platform to perform the modeling.more » In developing this model, improved anodic behavior prediction became necessary since a dense uranium-rich salt film was observed at the anode surface during electrorefining experiments. Experiments were conducted on uranium metal to determine the film's character and the conditions under which it forms. The electro-refiner salt used in all the experiments was eutectic LiCl/KCl containing UCl{sub 3}. The anodic film material was analyzed with ICP-OES to determine its composition. Both cyclic voltammetry and potentiodynamic scans were conducted at operating temperatures between 475 and 575 C. degrees to interrogate the electrochemical behavior of the uranium. The results show that an anodic film was produced on the uranium electrode. The film initially passivated the surface of the uranium on the working electrode. At high over potentials after a trans-passive region, the current observed was nearly equal to the current observed at the initial active level. Analytical results support the presence of K{sub 2}UCl{sub 6} at the uranium surface, within the error of the analytical method.« less

Recovery and recycling of uranium from rejected coated particles for compact high temperature reactors

NASA Astrophysics Data System (ADS)

Pai, Rajesh V.; Mollick, P. K.; Kumar, Ashok; Banerjee, J.; Radhakrishna, J.; Chakravartty, J. K.

2016-05-01

UO2 microspheres prepared by internal gelation technique were coated with pyrolytic carbon and silicon carbide using CVD technique. The particles which were not meeting the specifications were rejected. The rejected/failed UO2 based coated particles prepared by CVD technique was used for oxidation and recovery and recycling. The oxidation behaviour of sintered UO2 microspheres coated with different layers of carbon and SiC was studied by thermal techniques to develop a method for recycling and recovery of uranium from the failed/rejected coated particles. It was observed that the complete removal of outer carbon from the spheres is difficult. The crushing of microspheres enabled easier accessibility of oxygen and oxidation of carbon and uranium at 800-1000 °C. With the optimized process of multiple crushing using die & plunger and sieving the broken coated layers, we could recycle around fifty percent of the UO2 microspheres which could be directly recoated. The rest of the particles were recycled using a wet recycling method.

Removing oxygen from a solvent extractant in an uranium recovery process

DOEpatents

Hurst, Fred J.; Brown, Gilbert M.; Posey, Franz A.

1984-01-01

An improvement in effecting uranium recovery from phosphoric acid solutions is provided by sparging dissolved oxygen contained in solutions and solvents used in a reductive stripping stage with an effective volume of a nonoxidizing gas before the introduction of the solutions and solvents into the stage. Effective volumes of nonoxidizing gases, selected from the group consisting of argon, carbon dioxide, carbon monoxide, helium, hydrogen, nitrogen, sulfur dioxide, and mixtures thereof, displace oxygen from the solutions and solvents thereby reduce deleterious effects of oxygen such as excessive consumption of elemental or ferrous and accumulation of complex iron phosphates or cruds.

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.

Field Testing of Downgradient Uranium Mobility at an In-Situ Recovery Uranium Mine

NASA Astrophysics Data System (ADS)

Reimus, P. W.; Clay, J. T.; Rearick, M.; Perkins, G.; Brown, S. T.; Basu, A.; Chamberlain, K.

2015-12-01

In-situ recovery (ISR) mining of uranium involves the injection of O2 and CO2 (or NaHCO3) into saturated roll-front deposits to oxidize and solubilize the uranium, which is then removed by ion exchange at the surface and processed into U3O8. While ISR is economical and environmentally-friendly relative to conventional mining, one of the challenges of extracting uranium by this process is that it leaves behind a geochemically-altered aquifer that is exceedingly difficult to restore to pre-mining geochemical conditions, a regulatory objective. In this research, we evaluated the ability of the aquifer downgradient of an ISR mining area to attenuate the transport of uranium and other problem constituents that are mobilized by the mining process. Such an evaluation can help inform both regulators and the mining industry as to how much restoration of the mined ore zone is necessary to achieve regulatory compliance at various distances downgradient of the mining zone even if complete restoration of the ore zone proves to be difficult or impossible. Three single-well push-pull tests and one cross-well test were conducted in which water from an unrestored, previously-mined ore zone was injected into an unmined ore zone that served as a geochemical proxy for the downgradient aquifer. In all tests, non-reactive tracers were injected with the previously-mined ore zone water to allow the transport of uranium and other constituents to be compared to that of the nonreactive species. In the single-well tests, it was shown that the recovery of uranium relative to the nonreactive tracers ranged from 12-25%, suggesting significant attenuation capacity of the aquifer. In the cross-well test, selenate, molybdate and metavanadate were injected with the unrestored water to provide information on the transport of these potentially-problematic anionic constituents. In addition to the species-specific transport information, this test provided valuable constraints on redox conditions within the system, as redox couples involving these species collectively bracket the predicted transition redox potential for the U(VI)/U(IV) couple. Reduction should provide much longer-lasting immobilization of constituents than adsorption, especially given the inherent reducing characteristics of roll-front systems.

76 FR 36918 - Science Advisory Board Staff Office; Notification of a Public Teleconference and Meeting of the...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-06-23

... of EPA's Draft Technical Report Pertaining to Uranium and Thorium In-Situ Leach Recovery and Post... Related to Post-Closure Monitoring of Uranium In-Situ Leach/In-Situ Recovery (ISL/ISR) Sites.'' DATES: The... pertaining to Uranium In-Situ Leach Recovery--Post-Closure Stability Monitoring can be found at http://www...

Resource potential for commodities in addition to Uranium in sandstone-hosted deposits: Chapter 13

USGS Publications Warehouse

Breit, George N.

2016-01-01

Sandstone-hosted deposits mined primarily for their uranium content also have been a source of vanadium and modest amounts of copper. Processing of these ores has also recovered small amounts of molybdenum, rhenium, rare earth elements, scandium, and selenium. These deposits share a generally common origin, but variations in the source of metals, composition of ore-forming solutions, and geologic history result in complex variability in deposit composition. This heterogeneity is evident regionally within the same host rock, as well as within districts. Future recovery of elements associated with uranium in these deposits will be strongly dependent on mining and ore-processing methods.

PLUTONIUM RECOVERY FROM NEUTRON-BOMBARDED URANIUM FUEL

DOEpatents

Moore, R.H.

1964-03-24

A process of recovering plutonium from fuel by dissolution in molten KAlCl/sub 4/ double salt is described. Molten lithium chloride plus stannous chloride is added to reduce plutonium tetrachloride to the trichloride, which is dissolved in a lithium chloride phase while the uranium, as the tetrachloride, is dissolved in a double-salt phase. Separation of the two phases is discussed. (AEC)

75 FR 62153 - Notice of the Nuclear Regulatory Commission Issuance of Materials License SUA-1596 for Uranium...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-10-07

... Commission Issuance of Materials License SUA-1596 for Uranium One Americas, Inc. Moore Ranch In Situ Recovery..., Inc. (Uranium One) for its Moore Ranch uranium in situ recovery (ISR) facility in Campbell County... discussed in detail were the applicant's proposal as described in its license application to conduct in situ...

Biotechnology for the extractive metals industries

NASA Astrophysics Data System (ADS)

Brierley, James A.

1990-01-01

Biotechnology is an alternative process for the extraction of metals, the beneficiation of ores, and the recovery of metals from aqueous systems. Currently, microbial-based processes are used for leaching copper and uranium, enhancing the recovery of gold from refractory ores, and treating industrial wastewater to recover metal values. Future developments, emanating from fundamental and applied research and advances through genetic engineering, are expected to increase the use and efficiency of these biotechnological processes.

Reactive transport modeling at uranium in situ recovery sites: uncertainties in uranium sorption on iron hydroxides

USGS Publications Warehouse

Johnson, Raymond H.; Tutu, Hlanganani; Brown, Adrian; Figueroa, Linda; Wolkersdorfer, Christian

2013-01-01

Geochemical changes that can occur down gradient from uranium in situ recovery (ISR) sites are important for various stakeholders to understand when evaluating potential effects on surrounding groundwater quality. If down gradient solid-phase material consists of sandstone with iron hydroxide coatings (no pyrite or organic carbon), sorption of uranium on iron hydroxides can control uranium mobility. Using one-dimensional reactive transport models with PHREEQC, two different geochemical databases, and various geochemical parameters, the uncertainties in uranium sorption on iron hydroxides are evaluated, because these oxidized zones create a greater risk for future uranium transport than fully reduced zones where uranium generally precipitates.

The Resin-in-pulp Process and Its Application to Ores from Brosses "BRS 10"; LE PROCEDE "RESIN IN PULP" ET SON APPLICATION AUX MINERAIS DES BROSSES "BRS 10"

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

Kremer, M.

1959-03-01

The resin-in-pulp process is a technical variant of the recovery process of uranium in dilute solution by means of ion exchange resins. An anion resin, XE 123, of a welldefined grain size is placed in direct contact with the pulp produced by sulfuric acid attack on ore with a low uranium content. This process is of particular value in the treatment of pulps that cannot be filtered or decanted, such as those obtained with ore from Brosses. The preparation of the pulp, the elution of the uranium, and its fixation, as well as the various factors encountered in these operations,more » are discussed. (auth)« less

Processing of irradiated, enriched uranium fuels at the Savannah River Plant

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

Hyder, M L; Perkins, W C; Thompson, M C

Uranium fuels containing /sup 235/U at enrichments from 1.1% to 94% are processed and recovered, along with neptunium and plutonium byproducts. The fuels to be processed are dissolved in nitric acid. Aluminum-clad fuels are disssolved using a mercury catalyst to give a solution rich in aluminum. Fuels clad in more resistant materials are dissolved in an electrolytic dissolver. The resulting solutions are subjected to head-end treatment, including clarification and adjustment of acid and uranium concentration before being fed to solvent extraction. Uranium, neptunium, and plutonium are separated from fission products and from one another by multistage countercurrent solvent extraction withmore » dilute tri-n-butyl phosphate in kerosene. Nitric acid is used as the salting agent in addition to aluminum or other metal nitrates present in the feed solution. Nuclear safety is maintained through conservative process design and the use of monitoring devices as secondary controls. The enriched uranium is recovered as a dilute solution and shipped off-site for further processing. Neptunium is concentrated and sent to HB-Line for recovery from solution. The relatively small quantities of plutonium present are normally discarded in aqueous waste, unless the content of /sup 238/Pu is high enough to make its recovery desirable. Most of the /sup 238/Pu can be recovered by batch extraction of the waste solution, purified by counter-current solvent extraction, and converted to oxide in HB-Line. By modifying the flowsheet, /sup 239/Pu can be recovered from low-enriched uranium in the extraction cycle; neptunium is then not recovered. The solvent is subjected to an alkaline wash before reuse to remove degraded solvent and fission products. The aqueous waste is concentrated and partially deacidified by evaporation before being neutralized and sent to the waste tanks; nitric acid from the overheads is recovered for reuse.« less

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.

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.

COORDINATION COMPOUND-SOLVENT EXTRACTION PROCESS FOR URANIUM RECOVERY

DOEpatents

Reas, W.H.

1959-03-10

A method is presented for the separation of uranium from aqueous solutions containing a uranyl salt and thorium. Thc separation is effected by adding to such solutions an organic complexing agent, and then contacting the solution with an organic solvent in which the organic complexing agent is soluble. By use of the proper complexing agent in the proper concentrations uranium will be complexed and subsequently removed in the organic solvent phase, while the thorium remains in the aqueous phase. Mentioned as suitable organic complexing agents are antipyrine, bromoantipyrine, and pyramidon.

PLUTONIUM RECOVERY FROM NEUTRON-BOMBARDED URANIUM FUEL

DOEpatents

Moore, R.H.

1962-04-10

A process of recovering plutonium from neutronbombarded uranium fuel by dissolving the fuel in equimolar aluminum chloride-potassium chloride; heating the mass to above 700 deg C for decomposition of plutonium tetrachloride to the trichloride; extracting the plutonium trichloride into a molten salt containing from 40 to 60 mole % of lithium chloride, from 15 to 40 mole % of sodium chloride, and from 0 to 40 mole % of potassium chloride or calcium chloride; and separating the layer of equimolar chlorides containing the uranium from the layer formed of the plutonium-containing salt is described. (AEC)

PROCESS OF RECOVERING URANIUM

DOEpatents

Price, T.D.; Jeung, N.M.

1958-06-17

An improved precipitation method is described for the recovery of uranium from aqueous solutions. After removal of all but small amounts of Ni or Cu, and after complexing any iron present, the uranium is separated as the peroxide by adding H/sub 2/O/sub 2/. The improvement lies in the fact that the addition of H/sub 2/O/sub 2/ and consequent precipitation are carried out at a temperature below the freezing; point of the solution, so that minute crystals of solvent are present as seed crystals for the precipitation.

Method for oxygen reduction in a uranium-recovery process. [US DOE patent application

DOEpatents

Hurst, F.J.; Brown, G.M.; Posey, F.A.

1981-11-04

An improvement in effecting uranium recovery from phosphoric acid solutions is provided by sparging dissolved oxygen contained in solutions and solvents used in a reductive stripping stage with an effective volume of a nonoxidizing gas before the introduction of the solutions and solvents into the stage. Effective volumes of nonoxidizing gases, selected from the group consisting of argon, carbon dioxide, carbon monoxide, helium, hydrogen, nitrogen, sulfur dioxide, and mixtures thereof, displace oxygen from the solutions and solvents thereby reduce deleterious effects of oxygen such as excessive consumption of elemental or ferrous iron and accumulation of complex iron phosphates or cruds.

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

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

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

Youker, Amanda J.; Krebs, John F.; Quigley, Kevin J.

With funding from the National Nuclear Security Administrations Material Management and Minimization Office, Argonne National Laboratory (Argonne) is providing technical assistance to help accelerate the U.S. production of Mo-99 using a non-highly enriched uranium (non-HEU) source. A potential Mo-99 production pathway is by accelerator-initiated fissioning in a subcritical uranyl sulfate solution containing low enriched uranium (LEU). As part of the Argonne development effort, we are undertaking the AMORE (Argonne Molybdenum Research Experiment) project, which is essentially a pilot facility for all phases of Mo-99 production, recovery, and purification. Production of Mo-99 and other fission products in the subcritical target solutionmore » is initiated by putting an electron beam on a depleted uranium (DU) target; the fast neutrons produced in the DU target are thermalized and lead to fissioning of U-235. At the end of irradiation, Mo is recovered from the target solution and separated from uranium and most of the fission products by using a titania column. The Mo is stripped from the column with an alkaline solution. After acidification of the Mo product solution from the recovery column, the Mo is concentrated (and further purified) in a second titania column. The strip solution from the concentration column is then purified with the LEU Modified Cintichem process. A full description of the process can be found elsewhere [1–3]. The initial commissioning steps for the AMORE project include performing a Mo-99 spike test with pH 1 sulfuric acid in the target vessel without a beam on the target to demonstrate the initial Mo separation-and-recovery process, followed by the concentration column process. All glovebox operations were tested with cold solutions prior to performing the Mo-99 spike tests. Two Mo-99 spike tests with pH 1 sulfuric acid have been performed to date. Figure 1 shows the flow diagram for the remotely operated Mo-recovery system for the AMORE project. There are two separate pumps and flow paths for the acid and base operations. The system contains three sample ladders with eight sample loops per ladder for target mixing; column loading, including acid and water washes; and column stripping, including the final water wash.« less

RECOVERY OF VALUABLE MATERIAL FROM GRAPHITE BODIES

DOEpatents

Fromm, L.W. Jr.

1959-09-01

An electrolytic process for recovering uranium from a graphite fuel element is described. The uraniumcontaining graphite body is disposed as the anode of a cell containing a nitric acid electrolyte and a 5 amp/cm/sup 2/ current passed to induce a progressive disintegration of the graphite body. The dissolved uranium is quickly and easily separated from the resulting graphite particles by simple mechanical means, such as centrifugation, filtration, and decontamination.

JPRS Report, Science & Technology, Japan

DTIC Science & Technology

1987-11-12

Change (4) Future Direction Anyway, it has become almost clear that the effect of power recovery cannot be expected from the insulation of...process spent fuels in greater safety and to recover the uranium or plutonium from spent fuels for effective reapplication. In 1974, the PNC began...constructed to serve as a pilot plant that could be used to establish reprocessing technology for the next practical stage. 32 As for enriched uranium

75 FR 467 - Notice of Opportunity for Hearing, License Application Request of Powertech (USA) Inc. Dewey...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-01-05

... Hearing, License Application Request of Powertech (USA) Inc. Dewey-Burdock In Situ Uranium Recovery... Regulatory Commission (NRC) for the Dewey-Burdock In Situ Recovery Facility in Fall River and Custer Counties, South Dakota. The Dewey-Burdock facility would involve the recovery of uranium by in situ recovery (ISR...

75 FR 8936 - Environmental Impacts Statements; Notice of Availability

Federal Register 2010, 2011, 2012, 2013, 2014

2010-02-26

... Notices EIS No. 20090421, Draft EIS, NRC, WY, Moore Ranch In-Situ Uranium Recovery (ISR) Project, Proposal to Construct, Operate, Conduct Aquifer Restoration, and Decommission an In-Situ Recovery (ISR... No. 20090423, Draft EIS, NRC, WY, Nichols Ranch In-Situ Uranium Recovery (ISR) Project, Proposal to...

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.

The role of uranium-arene bonding in H2O reduction catalysis

NASA Astrophysics Data System (ADS)

Halter, Dominik P.; Heinemann, Frank W.; Maron, Laurent; Meyer, Karsten

2018-03-01

The reactivity of uranium compounds towards small molecules typically occurs through stoichiometric rather than catalytic processes. Examples of uranium catalysts reacting with water are particularly scarce, because stable uranyl groups form that preclude the recovery of the uranium compound. Recently, however, an arene-anchored, electron-rich uranium complex has been shown to facilitate the electrocatalytic formation of H2 from H2O. Here, we present the precise role of uranium-arene δ bonding in intermediates of the catalytic cycle, as well as details of the atypical two-electron oxidative addition of H2O to the trivalent uranium catalyst. Both aspects were explored by synthesizing mid- and high-valent uranium-oxo intermediates and by performing comparative studies with a structurally related complex that cannot engage in δ bonding. The redox activity of the arene anchor and a covalent δ-bonding interaction with the uranium ion during H2 formation were supported by density functional theory analysis. Detailed insight into this catalytic system may inspire the design of ligands for new uranium catalysts.

Preparation of High Purity, High Molecular-Weight Chitin from Ionic Liquids for Use as an Adsorbate for the Extraction of Uranium from Seawater

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

Rogers, Robin

Ensuring a domestic supply of uranium is a key issue facing the wider implementation of nuclear power. Uranium is mostly mined in Kazakhstan, Australia, and Canada, and there are few high-grade uranium reserves left worldwide. Therefore, one of the most appealing potential sources of uranium is the vast quantity dissolved in the oceans (estimated to be 4.4 billion tons worldwide). There have been research efforts centered on finding a means to extract uranium from seawater for decades, but so far none have resulted in an economically viable product, due in part to the fact that the materials that have beenmore » successfully demonstrated to date are too costly (in terms of money and energy) to produce on the necessary scale. Ionic Liquids (salts which melt below 100{degrees}C) can completely dissolve raw crustacean shells, leading to recovery of a high purity, high molecular weight chitin powder and to fibers and films which can be spun directly from the extract solution suggesting that continuous processing might be feasible. The work proposed here will utilize the unprecedented control this makes possible over the chitin fiber a) to prepare electrospun nanofibers of very high surface area and in specific architectures, b) to modify the fiber surfaces chemically with selective extractant capacity, and c) to demonstrate their utility in the direct extraction and recovery of uranium from seawater. This approach will 1) provide direct extraction of chitin from shellfish waste thus saving energy over the current industrial process for obtaining chitin; 2) allow continuous processing of nanofibers for very high surface area fibers in an economical operation; 3) provide a unique high molecular weight chitin not available from the current industrial process, leading to stronger, more durable fibers; and 4) allow easy chemical modification of the large surface areas of the fibers for appending uranyl selective functionality providing selectivity and ease of stripping. The resulting sorbent should prove economically feasible, as well as providing an overall net energy gain.« less

A study on recovery of uranium in the anode basket residues delivered from the pyrochemical process of used nuclear fuel

NASA Astrophysics Data System (ADS)

Eun, H. C.; Kim, T. J.; Jang, J. H.; Kim, G. Y.; Park, S. B.; Yoon, D. S.; Kim, S. H.; Paek, S. W.; Lee, S. J.

2018-04-01

In this study, the chlorination of uranium oxide (UO2) using ammonium chloride and zirconium as chemical agents was conducted to recover the uranium in the anode basket residues from the pyrochemical process of used nuclear fuel. The chlorination of UO2 was predicted using thermodynamic equilibrium calculations. The experimental conditions for the chlorination were determined using a chlorination test with cerium oxide (CeO2). In the chlorination test, it was confirmed that UO2 was chlorinated into UCl3 at 320 °C, some UO2 remained without changes in the chemical form, and ZrO2, Zr2O, and ZrCl2 were generated as byproducts.

PROCESS FOR RECOVERY OF URANIUM AND VANADIUM FROM CARBONATE SOLUTIONS BY REDUCTION-PRECIPITATION

DOEpatents

Ellis, D.A.; Lindblom, R.O.

1957-09-24

A process employing carbonate leaching of ores and an advantageous methcd of recovering the uranium and vanadium from the leach solution is described. The uranium and vanadium can be precipitated from carbonate leach solutions by reaction with sodium amalgam leaving the leach solution in such a condition that it is economical to replenish for recycling. Such a carbonate leach solution is treated with a dilute sodium amalgam having a sodium concentration within a range of about 0.01 to 0.5% of sodium. Efficiency of the treatment is dependent on at least three additional factors, intimacy of contact of the amalgam with the leach solution, rate of addition of the amalgam and exclusion of oxygen (air).

77 FR 25193 - Notice of Availability of the Draft Environmental Impact Statement for the Lost Creek Uranium In...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-04-27

... (EIS) for the Lost Creek Uranium In Situ Recovery (ISR) Project and by this notice is announcing the...-166318] Notice of Availability of the Draft Environmental Impact Statement for the Lost Creek Uranium In Situ Recovery Project, Wyoming AGENCY: Bureau of Land Management, Interior. [[Page 25194

Extraction of uranium from tailings by sulfuric acid leaching with oxidants

NASA Astrophysics Data System (ADS)

Huang, Jing; Li, Mi; Zhang, Xiaowen; Huang, Chunmei; Wu, Xiaoyan

2017-06-01

Recovery of uranium have been performed by leaching uranium-containing tailings in sulfuric acid system with the assistance of HF, HClO4, H2O2 and MnO2. The effect of reagent dosage, sulfuric acid concentration, Liquid/solid ratio, reaction temperature and particle size on the leaching of uranium were investigated. The results show that addiction of HF, HClO4, H2O2 and MnO2 significantly increased the extraction of uranium under 1M sulphuric acid condition and under the optimum reaction conditions a dissolution fraction of 85% by HClO4, 90% by HF, 95% by H2O2 can be reached respectively. The variation of technological mineralogy properites of tailings during leaching process show that the assistants can break gangue effectively. These observations suggest that optimum oxidants could potentially influence the extraction of uranium from tailings even under dilute acid condition.

Looking Northwest at Uranium Dryers Along North Side of Green ...

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

Looking Northwest at Uranium Dryers Along North Side of Green Room in Recycle Recovery Building - Hematite Fuel Fabrication Facility, Recycle Recovery Building, 3300 State Road P, Festus, Jefferson County, MO

75 FR 71702 - Science Advisory Board Staff Office; Request for Nominations of Experts for Review of EPA's Draft...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-11-24

... Nominations of Experts for Review of EPA's Draft Technical Report Pertaining to Uranium and Thorium In-Situ... expectation is that In-Situ Leach Recovery (ISL/ISR) operations will be the most common type of new uranium... pertaining to Uranium In-Situ Leach Recovery--Post-Closure Stability Monitoring can be found at the following...

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...

Dissolution of uranium oxides from simulated environmental swipes using ammonium bifluoride

DOE PAGES

Meyers, Lisa A.; Yoshida, Thomas M.; Chamberlin, Rebecca M.; ...

2016-11-01

We developed an analytical chemistry method to quantitatively recover microgram quanties of solid uranium oxides from swipe media using ammonium bifluoride (ABF, NH 4HF 2) solution. Recovery of uranium from surrogate swipe media (filter paper) was demonstrated at initial uranium loading levels between 3 and 20 µg filter -1. Moreover, the optimal conditions for extracting U 3O 8 and UO 2 are using 1 % ABF solution and incubating at 80 °C for one hour. The average uranium recoveries are 100 % for U 3O 8, and 90 % for UO 2. Finally, with this method, uranium concentration as lowmore » as 3 µg filter -1 can be recovered for analysis.« less

Accumulation of uranium by immobilized persimmon tannin

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

Sakaguchi, Takashi; Nakajima, Akira

1994-01-01

We have discovered that the extracted juice of unripe astringent persimmon fruit, designated as kakishibu or shibuol, has an extremely high affinity for uranium. To develop efficient adsorbents for uranium, we tried to immobilize kakishibu (persimmon tannin) with various aldehydes and mineral acids. Persimmon tannin immobilized with glutaraldehyde can accumulate 1.71 g (14 mEq U) of uranium per gram of the adsorbent. The uranium accumulating capacity of this adsorbent is several times greater than that of commercially available chelating resins (2-3 mEq/g). Immobilized persimmon tannin has the most favorable features for uranium recovery; high selective adsorption ability, rapid adsorption rate,more » and applicability in both column and batch systems. The uranium retained on immobilized persimmon tannin can be quantitatively and easily eluted with a very dilute acid, and the adsorbent can thus be easily recycled in the adsorption-desorption process. Immobilized persimmon tannin also has a high affinity for thorium. 23 refs., 13 figs., 7 tabs.« less

Ferric ion as a scavenging agent in a solvent extraction process

DOEpatents

Bruns, Lester E.; Martin, Earl C.

1976-01-01

Ferric ions are added into the aqueous feed of a plutonium scrap recovery process that employs a tributyl phosphate extractant. Radiolytic degradation products of tributyl phosphate such as dibutyl phosphate form a solid precipitate with iron and are removed from the extraction stages via the waste stream. Consequently, the solvent extraction characteristics are improved, particularly in respect to minimizing the formation of nonstrippable plutonium complexes in the stripping stages. The method is expected to be also applicable to the partitioning of plutonium and uranium in a scrap recovery process.

BLENDING LOW ENRICHED URANIUM WITH DEPLETED URANIUM TO CREATE A SOURCE MATERIAL ORE THAT CAN BE PROCESSED FOR THE RECOVERY OF YELLOWCAKE AT A CONVENTIONAL URANIUM MILL

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

Schutt, Stephen M.; Hochstein, Ron F.; Frydenlund, David C.

2003-02-27

Throughout the United States Department of Energy (DOE) complex, there are a number of streams of low enriched uranium (LEU) that contain various trace contaminants. These surplus nuclear materials require processing in order to meet commercial fuel cycle specifications. To date, they have not been designated as waste for disposal at the DOE's Nevada Test Site (NTS). Currently, with no commercial outlet available, the DOE is evaluating treatment and disposal as the ultimate disposition path for these materials. This paper will describe an innovative program that will provide a solution to DOE that will allow disposition of these materials atmore » a cost that will be competitive with treatment and disposal at the NTS, while at the same time recycling the material to recover a valuable energy resource (yellowcake) for reintroduction into the commercial nuclear fuel cycle. International Uranium (USA) Corporation (IUSA) and Nuclear Fuel Services, Inc. (NFS) have entered into a commercial relationship to pursue the development of this program. The program involves the design of a process and construction of a plant at NFS' site in Erwin, Tennessee, for the blending of contaminated LEU with depleted uranium (DU) to produce a uranium source material ore (USM Ore{trademark}). The USM Ore{trademark} will then be further processed at IUC's White Mesa Mill, located near Blanding, Utah, to produce conventional yellowcake, which can be delivered to conversion facilities, in the same manner as yellowcake that is produced from natural ores or other alternate feed materials. The primary source of feed for the business will be the significant sources of trace contaminated materials within the DOE complex. NFS has developed a dry blending process (DRYSM Process) to blend the surplus LEU material with DU at its Part 70 licensed facility, to produce USM Ore{trademark} with a U235 content within the range of U235 concentrations for source material. By reducing the U235 content to source material levels in this manner, the material will be suitable for processing at a conventional uranium mill under its existing Part 40 license to remove contaminants and enable the product to re-enter the commercial fuel cycle. The tailings from processing the USM Ore{trademark} at the mill will be permanently disposed of in the mill's tailings impoundment as 11e.(2) byproduct material. Blending LEU with DU to make a uranium source material ore that can be returned to the nuclear fuel cycle for processing to produce yellowcake, has never been accomplished before. This program will allow DOE to disposition its surplus LEU and DU in a cost effective manner, and at the same time provide for the recovery of valuable energy resources that would be lost through processing and disposal of the materials. This paper will discuss the nature of the surplus LEU and DU materials, the manner in which the LEU will be blended with DU to form a uranium source material ore, and the legal means by which this blending can be accomplished at a facility licensed under 10 CFR Part 70 to produce ore that can be processed at a conventional uranium mill licensed under 10 CFR Part 40.« less

Materials for the Recovery of Uranium from Seawater

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

Abney, Carter W.; Mayes, Richard T.; Saito, Tomonori

More than 1000× uranium exists in the oceans than exists in terrestrial ores. With nuclear power generation expected to increase over the coming decades, access to this unconventional reserve is a matter of energy security. With origins in the mid-1950’s, materials have been developed for the selective recovery of seawater uranium for more than six decades, with a renewed interest in particular since 2010. This review comprehensively surveys materials developed from 2000 – 2016 for recovery of seawater uranium, in particular including recent developments in inorganic materials, polymer adsorbents and related research pertaining to amidoxime, and nanostructured materials such asmore » metal-organic frameworks, porous-organic polymers, and mesoporous carbons. In conclusion, challenges of performing reliable and reproducible uranium adsorption studies are also discussed, as well as the standardization of parameters necessary to ensure valid comparisons between different adsorbents.« less

Materials for the Recovery of Uranium from Seawater

DOE PAGES

Abney, Carter W.; Mayes, Richard T.; Saito, Tomonori; ...

2017-11-22

More than 1000× uranium exists in the oceans than exists in terrestrial ores. With nuclear power generation expected to increase over the coming decades, access to this unconventional reserve is a matter of energy security. With origins in the mid-1950’s, materials have been developed for the selective recovery of seawater uranium for more than six decades, with a renewed interest in particular since 2010. This review comprehensively surveys materials developed from 2000 – 2016 for recovery of seawater uranium, in particular including recent developments in inorganic materials, polymer adsorbents and related research pertaining to amidoxime, and nanostructured materials such asmore » metal-organic frameworks, porous-organic polymers, and mesoporous carbons. In conclusion, challenges of performing reliable and reproducible uranium adsorption studies are also discussed, as well as the standardization of parameters necessary to ensure valid comparisons between different adsorbents.« less

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)

Symposium on the reprocessing of irradiated fuels. Book 2, Session IV

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

None

1958-12-31

Book two of this conference has a single-focused session IV entitled Nonaqueous Processing, with 8 papers. The session deals with fluoride volatility processes and pyrometallurgical or pyrochemical processes. The latter involves either an oxide drossing or molten metal extraction or fused salt extraction technique and results in only partial decontamination. Fluoride volatility processes appear to be especially favorable for recovery of enriched uranium and decontamination factors of 10/sup 7/ to 10/sup 8/ would be achieved by simpler means than those employed in solvent extraction. Data from lab research on the BrF/sub 3/ process and the ClF/sub 3/ process are givenmore » and discussed and pilot plant experience is described, all in connection with natural uranium or slightly enriched uranium processing. Fluoride volatility processes for enriched or high alloy fuels are described step by step. The economic and engineering considerations of both types of nonaqueous processing are treated separately and as fully as present knowledge allows. A comprehensive review of the chemistry of pyrometallurgical processes is included.« less

Column bioleaching of uranium embedded in granite porphyry by a mesophilic acidophilic consortium.

PubMed

Qiu, Guanzhou; Li, Qian; Yu, Runlan; Sun, Zhanxue; Liu, Yajie; Chen, Miao; Yin, Huaqun; Zhang, Yage; Liang, Yili; Xu, Lingling; Sun, Limin; Liu, Xueduan

2011-04-01

A mesophilic acidophilic consortium was enriched from acid mine drainage samples collected from several uranium mines in China. The performance of the consortium in column bioleaching of low-grade uranium embedded in granite porphyry was investigated. The influences of several chemical parameters on uranium extraction in column reactor were also investigated. A uranium recovery of 96.82% was achieved in 97 days column leaching process including 33 days acid pre-leaching stage and 64 days bioleaching stage. It was reflected that indirect leaching mechanism took precedence over direct. Furthermore, the bacterial community structure was analyzed by using Amplified Ribosomal DNA Restriction Analysis. The results showed that microorganisms on the residual surface were more diverse than that in the solution. Acidithiobacillus ferrooxidans was the dominant species in the solution and Leptospirillum ferriphilum on the residual surface. Copyright © 2011 Elsevier Ltd. All rights reserved.

NUCLEAR REACTOR FUEL SYSTEMS

DOEpatents

Thamer, B.J.; Bidwell, R.M.; Hammond, R.P.

1959-09-15

Homogeneous reactor fuel solutions are reported which provide automatic recombination of radiolytic gases and exhibit large thermal expansion characteristics, thereby providing stability at high temperatures and enabling reactor operation without the necessity of apparatus to recombine gases formed by the radiolytic dissociation of water in the fuel and without the necessity of liquid fuel handling outside the reactor vessel except for recovery processes. The fuels consist of phosphoric acid and water solutions of enriched uranium, wherein the uranium is in either the hexavalent or tetravalent state.

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)

Microbial communities associated with uranium in-situ recovery mining process are related to acid mine drainage assemblages.

PubMed

Coral, Thomas; Descostes, Michaël; De Boissezon, Hélène; Bernier-Latmani, Rizlan; de Alencastro, Luiz Felippe; Rossi, Pierre

2018-07-01

A large fraction (47%) of the world's uranium is mined by a technique called "In Situ Recovery" (ISR). This mining technique involves the injection of a leaching fluid (acidic or alkaline) into a uranium-bearing aquifer and the pumping of the resulting solution through cation exchange columns for the recovery of dissolved uranium. The present study reports the in-depth alterations brought to autochthonous microbial communities during acidic ISR activities. Water samples were collected from a uranium roll-front deposit that is part of an ISR mine in operation (Tortkuduk, Kazakhstan). Water samples were obtained at a depth of ca 500 m below ground level from several zones of the Uyuk aquifer following the natural redox zonation inherited from the roll front deposit, including the native mineralized orebody and both upstream and downstream adjacent locations. Samples were collected equally from both the entrance and the exit of the uranium concentration plant. Next-generation sequencing data showed that the redox gradient shaped the community structures, within the anaerobic, reduced, and oligotrophic habitats of the native aquifer zones. Acid injection induced drastic changes in the structures of these communities, with a large decrease in both cell numbers and diversity. Communities present in the acidified (pH values < 2) mining areas exhibited similarities to those present in acid mine drainage, with the dominance of Sulfobacillus sp., Leptospirillum sp. and Acidithiobacillus sp., as well as the archaean Ferroplasma sp. Communities located up- and downstream of the mineralized zone under ISR and affected by acidic fluids were blended with additional facultative anaerobic and acidophilic microorganisms. These mixed biomes may be suitable communities for the natural attenuation of ISR mining-affected subsurface through the reduction of metals and sulfate. Assessing the effect of acidification on the microbial community is critical to evaluating the potential for natural attenuation or active bioremediation strategies. Copyright © 2018 Elsevier B.V. All rights reserved.

Field Evaluation of the Restorative Capacity of the Aquifer Downgradient of a Uranium In-Situ Recovery Mining Site

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

Reimus, Paul William

A two-part field study was conducted in Smith Ranch-Highland in-situ recovery (ISR) near Douglas, Wyoming, to evaluate the restorative capacity of the aquifer downgradient (i.e., hydrologically downstream) of a Uranium ISR mining site with respect to the transport of uranium and other potential contaminants in groundwater after mining has ceased. The study was partially conducted by checking the Uranium content and the alkalinity of separate wells, some wells had been restored and others had not. A map and in-depth procedures of the study are included.

Nanostructured Metal Oxide Sorbents for the Collection and Recovery of Uranium from Seawater

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

Chouyyok, Wilaiwan; Warner, Cynthia L.; Mackie, Katherine E.

2016-02-07

The ability to collect uranium from seawater offers the potential for a long-term green fuel supply for nuclear energy. However, extraction of uranium, and other trace minerals, is challenging due to the high ionic strength and low mineral concentrations in seawater. Herein we evaluate the use of nanostructured metal oxide sorbents for the collection and recovery of uranium from seawater. Chemical affinity, chemical adsorption capacity and kinetics of preferred sorbent materials were evaluated. High surface area manganese and iron oxide nanomaterials showed excellent performance for uranium collection from seawater. Inexpensive nontoxic carbonate solutions were demonstrated to be an effective andmore » environmental benign method of stripping the uranium from the metal oxide sorbents. Various formats for the utilization of the nanostructured metals oxide sorbent materials are discussed including traditional and nontraditional methods such as magnetic separation. Keywords: Uranium, nano, manganese, iron, sorbent, seawater, magnetic, separations, nuclear energy« less

17. VIEW OF HYDRIDING SYSTEM IN BUILDING 881. THE HYDRIDING ...

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

17. VIEW OF HYDRIDING SYSTEM IN BUILDING 881. THE HYDRIDING SYSTEM WAS PART OF THE FAST ENRICHED URANIUM RECOVERY PROCESS. (11/11/59) - Rocky Flats Plant, General Manufacturing, Support, Records-Central Computing, Southern portion of Plant, Golden, Jefferson County, CO

Ross In Situ Uranium Recovery Project NESHAP Subpart W Construction Approval

EPA Pesticide Factsheets

On May 5, 2015, EPA issued a Construction Approval under the National Emission Standards for Hazardous Air Pollutants (NESHAPs) at 40 CFR Part 61, subpart W, to Strata Energy, Inc., for their Ross In Situ Recovery (ISR) Uranium Project in Crook County, WY.

Investigations Into the Reusability of Amidoxime-Based Polymeric Uranium Adsorbents

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

Kuo, Li-Jung; Gill, Gary A.; Strivens, Jonathan E.

Significant advancements in amidoxime-based polymeric adsorbents to extract uranium from seawater are achieved in recent years. The success of uranium adsorbent development can help provide a sustainable supply of fuel for nuclear reactors. To bring down the production cost of this new technology, in addition to the development of novel adsorbents with high uranium capacity and manufacture cost, the development of adsorbent re-using technique is critical because it can further reduce the cost of the adsorbent manufacture. In our last report, the use of high concentrations of bicarbonate solution (3M KHCO3) was identified as a cost-effective, environmental friendly method tomore » strip uranium from amidoxime-based polymeric adsorbents. This study aims to further improve the method for high recovery of uranium capacity in re-uses and to evaluate the performance of adsorbents after multiple re-use cycles. Adsorption of dissolved organic matter (DOM) on the uranium adsorbents during seawater exposure can hinder the uranium adsorption and slow down the adsorption rate. An additional NaOH rinse (0.5 M NaOH, room temperature) was applied after the 3 M KHCO3 elution to remove natural organic matter from adsorbents. The combination of 3 M KHCO3 elution and 0.5 M NaOH rinse significantly improves the recovery of uranium adsorption capacity in the re-used adsorbents. In the first re-use, most ORNL adsorbents tested achieve ~100% recovery by using 3 M KHCO3 elution + 0.5 M NaOH rinse approach, in comparison to 54% recovery when only 3 M KHCO3 elution was applied. A significant drop in capacity was observed when the adsorbents went through more than one re-use. FTIR spectra revealed that degradation of amidoxime ligands occurs during seawater exposure, and is more significant the longer the exposure time. Significantly elevated ratios of Ca/U and Mg/U in re-used adsorbents support the decrease in abundance of amidoxime ligands and increase carboxylate group from FT-IR analysis. The impact of the length of seawater exposure cycle in adsorbent re-use was evaluated by comparing the adsorption capacity for a common adsorbent formulation (ORNL AI8 formulation) under different exposure cycle (28 days and 42 days). Adsorbents with a 28 days seawater exposure cycle had higher recovery of uranium capacity than adsorbent with 42 days of seawater exposure. Under different cumulative seawater exposure time, the adsorbent with 28 days seawater exposure cycle also had less amidoxime ligands degradation than the adsorbent with 42 days seawater exposure cycle. These observations support the negative impact of prolonged seawater exposure on amidoxime ligands stability. Recovery of uranium capacity in re-uses also varies across different adsorbent formulations. Among three different ORNL adsorbents tested (AI8, AF8, AF1-DMSO), AI8 had the best recovery in each re-use, followed by AF8 and then AF1-DMSO. This demonstrates that continuing efforts on developing new adsorbents with high capacity and stability is critical. The overall performance of adsorbents in multiple re-use cycles can be evaluated by calculation total harvestable uranium, the summation of adsorbed uranium from each seawater exposure cycle. In this assessment, the ORNL AI8 braid with 28 days seawater exposure cycle can reach total harvestable uranium 10g Uranium/kg adsorbent in ~140 days; while the same type of braid but with 42 days seawater exposure cycle reach the same level in ~170 days. Notably, the performance of total harvestable uranium also varies among different adsorbent formulations (AI8 > AF1-DMSO > AF8). Short seawater exposure cycle is associated with high re-use frequency. The development of low-cost offshore adsorbent deployment/extraction is essential for high frequency reuse operation. This study also highlights the importance to examine the re-use performance of newly developed uranium adsorbents for selection of optimal adsorbents for ocean deployment.« less

Highly Efficient Interception and Precipitation of Uranium(VI) from Aqueous Solution by Iron-Electrocoagulation Combined with Cooperative Chelation by Organic Ligands.

PubMed

Li, Peng; Zhun, Bao; Wang, Xuegang; Liao, PingPing; Wang, Guanghui; Wang, Lizhang; Guo, Yadan; Zhang, Weimin

2017-12-19

A new strategy combining iron-electrocoagulation and organic ligands (OGLs) cooperative chelation was proposed to screen and precipitate low concentrations (0-18.52 μmol/L) of uranium contaminant in aqueous solution. We hypothesized that OGLs with amino, hydroxyl, and carboxyl groups hydrophobically/hydrophilically would realize precuring of uranyl ion at pH < 3.0, and the following iron-electrocoagulation would achieve faster and more efficient uranium precipitation. Experimentally, the strategy demonstrated highly efficient uranium(VI) precipitation efficiency, especially with hydrophilic macromolecular OGLs. The uranium removal efficiency at optimized experimental condition reached 99.65%. The decrease of zeta potential and the lattice enwrapping between U-OGLs chelates and flocculation precursor were ascribed to the enhanced uranium precipitation activity. Uranium was precipitated as oxides of U(VI) or higher valences that were easily captured in aggregated micelles under low operation current potential. The actual uranium tailing wastewater was treated, and a satisfied uranium removal efficiency of 99.02% was discovered. After elution of the precipitated flocs, a concentrated uranium solution (up to 106.52 μmol/L) with very few other metallic impurities was obtained. Therefore, the proposed strategy could remove uranium and concentrate it concurrently. This work could provide new insights into the purification and recovery of uranium from aqueous solutions in a cost-effective and environmentally friendly process.

10 CFR 51.60 - Environmental report-materials licenses.

Code of Federal Regulations, 2011 CFR

2011-01-01

... oil and gas recovery. (vii) Construction and operation of a uranium enrichment facility. (2) Issuance... conversion of uranium hexafluoride pursuant to part 70 of this chapter. (ii) Possession and use of source material for uranium milling or production of uranium hexafluoride pursuant to part 40 of this chapter...

10 CFR 51.60 - Environmental report-materials licenses.

Code of Federal Regulations, 2010 CFR

2010-01-01

... oil and gas recovery. (vii) Construction and operation of a uranium enrichment facility. (2) Issuance... conversion of uranium hexafluoride pursuant to part 70 of this chapter. (ii) Possession and use of source material for uranium milling or production of uranium hexafluoride pursuant to part 40 of this chapter...

10 CFR 51.60 - Environmental report-materials licenses.

Code of Federal Regulations, 2013 CFR

2013-01-01

... oil and gas recovery. (vii) Construction and operation of a uranium enrichment facility. (2) Issuance... conversion of uranium hexafluoride pursuant to part 70 of this chapter. (ii) Possession and use of source material for uranium milling or production of uranium hexafluoride pursuant to part 40 of this chapter...

10 CFR 51.60 - Environmental report-materials licenses.

Code of Federal Regulations, 2014 CFR

2014-01-01

... oil and gas recovery. (vii) Construction and operation of a uranium enrichment facility. (2) Issuance... conversion of uranium hexafluoride pursuant to part 70 of this chapter. (ii) Possession and use of source material for uranium milling or production of uranium hexafluoride pursuant to part 40 of this chapter...

10 CFR 51.60 - Environmental report-materials licenses.

Code of Federal Regulations, 2012 CFR

2012-01-01

... oil and gas recovery. (vii) Construction and operation of a uranium enrichment facility. (2) Issuance... conversion of uranium hexafluoride pursuant to part 70 of this chapter. (ii) Possession and use of source material for uranium milling or production of uranium hexafluoride pursuant to part 40 of this chapter...

Design of a Fission 99 Mo Recovery Process and Implications toward Mo Adsorption Mechanism on Titania and Alumina Sorbents

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

Stepinski, Dominique C.; Youker, Amanda J.; Krahn, Elizabeth O.

2017-03-01

Molybdenum-99 is a parent of the most widely used medical isotope technetium-99m. Proliferation concerns have prompted development of alternative Mo production methods utilizing low enriched uranium. Alumina and titania sorbents were evaluated for separation of Mo from concentrated uranyl nitrate solutions. System, mass transfer, and isotherm parameters were determined to enable design of Mo separation processes under a wide range of conditions. A model-based approach was utilized to design representative commercial-scale column processes. The designs and parameters were verified with bench-scale experiments. The results are essential for design of Mo separation processes from irradiated uranium solutions, selection of support materialmore » and process optimization. Mo uptake studies show that adsorption decreases with increasing concentration of uranyl nitrate; howeveL, examination of Mo adsorption as a function of nitrate ion concentration shows no dependency, indicating that uranium competes with Mo for adsorption sites. These results are consistent with reports indicating that Mo forms inner-sphere complexes with titania and alumina surface groups.« less

76 FR 69295 - Strata Energy, Inc.; Establishment of Atomic Safety and Licensing Board

Federal Register 2010, 2011, 2012, 2013, 2014

2011-11-08

...: Strata Energy, Inc. (Ross In Situ Recovery Uranium Project) This proceeding involves a license... byproduct materials license at its Ross In Situ Recovery Uranium Project site located in Crook County... dated December 29, 1972, published in the Federal Register, 37 FR 28,710 (1972), and the Commission's...

75 FR 13141 - Powertech (USA), Inc.; Establishment of Atomic Safety and Licensing Board

Federal Register 2010, 2011, 2012, 2013, 2014

2010-03-18

... following proceeding: Powertech (USA) Inc. (Dewey-Burdock In Situ Uranium Recovery Facility). This Board is... Powertech (USA) Inc.'s application for a source materials license for an in situ uranium recovery facility... Commission dated December 29, 1972, published in the Federal Register, 37 FR 28,710 (1972), and the...

SOLVENT EXTRACTION PROCESS FOR THE RECOVERY OF METALS FROM PHOSPHORIC ACID

DOEpatents

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

1958-11-01

> A solvent extraction process is presented for recovering metal values including uranium, thorium, and other lanthanide and actinide elements from crude industrial phosphoric acid solutions. The process conslsts of contacting said solution with an immisclble organic solvent extractant containing a diluent and a material selected from the group consisting of mono and di alkyl phosphates, alkyl phosphonates and alkyl phosphites. The uranlum enters the extractant phase and is subsequently recovered by any of the methods known to the art. Recovery is improved if the phosphate solution is treated with a reducing agent such as iron or aluminum powder prior to the extraction step.

Basic Mechanisms of Radiation Effects on Electronic Materials, Devices, and Integrated Circuits

DTIC Science & Technology

1982-08-01

recovery time versus reciprocal tempera- ture derived from data of the type shown in Figure 18. . . .31 20 Several ways to alter the charje state of...and long-term recovery processes that occUr in neutron-irradiated silicon ........ 40 29 Annealing factor versus time for 11 ohm-cm p-type bulk silicon...radioactive ele- ments (such as uranium and thorium) which, when incorporated in packaged integrated circuits, can cause occasional transient upsets

Bidentate organophosphorus solvent extraction process for actinide recovery and partition

DOEpatents

Schulz, Wallace W.

1976-01-01

A liquid-liquid extraction process for the recovery and partitioning of actinide values from acidic nuclear waste aqueous solutions, the actinide values including trivalent, tetravalent and hexavalent oxidation states is provided and includes the steps of contacting the aqueous solution with a bidentate organophosphorous extractant to extract essentially all of the actinide values into the organic phase. Thereafter the respective actinide fractions are selectively partitioned into separate aqueous solutions by contact with dilute nitric or nitric-hydrofluoric acid solutions. The hexavalent uranium is finally removed from the organic phase by contact with a dilute sodium carbonate solution.

Use of steel and tantalum apparatus for molten Cd-Mg-Zn alloys

NASA Technical Reports Server (NTRS)

Bennett, G. A.; Burris, L., Jr.; Kyle, M. L.; Nelson, P. A.

1966-01-01

Steel and tantalum apparatus contains various ternary alloys of cadmium, zinc, and magnesium used in pyrochemical processes for the recovery of uranium-base reactor fuels. These materials exhibit good corrosion resistance at the high temperatures necessary for fuel separation in liquid metal-molten salt solvents.

78 FR 28897 - Lost Creek ISR, LLC, Lost Creek Uranium In-Situ Recovery Project; Sweetwater County, Wyoming

Federal Register 2010, 2011, 2012, 2013, 2014

2013-05-16

... ISR, LLC, Lost Creek Uranium In-Situ Recovery Project; Sweetwater County, Wyoming AGENCY: Nuclear... amendment, correction. SUMMARY: This document corrects a notice appearing in the Federal Register on April 3, 2013 [78 FR 20146], that listed, in tabular format, documents that related to the notice. This action...

Engineering study of 50 miscellaneous inactive underground radioactive waste tanks located at the Hanford Site, Washington

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

Freeman-Pollard, J.R.

1994-03-02

This engineering study addresses 50 inactive underground radioactive waste tanks. The tanks were formerly used for the following functions associated with plutonium and uranium separations and waste management activities in the 200 East and 200 West Areas of the Hanford Site: settling solids prior to disposal of supernatant in cribs and a reverse well; neutralizing acidic process wastes prior to crib disposal; receipt and processing of single-shell tank (SST) waste for uranium recovery operations; catch tanks to collect water that intruded into diversion boxes and transfer pipeline encasements and any leakage that occurred during waste transfer operations; and waste handlingmore » and process experimentation. Most of these tanks have not been in use for many years. Several projects have, been planned and implemented since the 1970`s and through 1985 to remove waste and interim isolate or interim stabilize many of the tanks. Some tanks have been filled with grout within the past several years. Responsibility for final closure and/or remediation of these tanks is currently assigned to several programs including Tank Waste Remediation Systems (TWRS), Environmental Restoration and Remedial Action (ERRA), and Decommissioning and Resource Conservation and Recovery Act (RCRA) Closure (D&RCP). Some are under facility landlord responsibility for maintenance and surveillance (i.e. Plutonium Uranium Extraction [PUREX]). However, most of the tanks are not currently included in any active monitoring or surveillance program.« less

RECOVERY OF THORIUM AND URANIUM VALUES FROM AQUEOUS SOLUTIONS

DOEpatents

Calkins, G.D.

1958-02-18

This patent deals with the separation and recovery of uranium from monazite sand. After initial treatment of the sand with sodium hydroxide, a precipitate is obtuined which contains the uranium, thorium, rare earths and some phosphorus. This precipitate is then dissolved in nitric acid. The bulk of the rare earths are removed from thls soiution by adding aa excess of alkali carbonate, causing precipitation of the rare earths together with part of the thorium present. The solution still contains a considerable amount of thorium, some rare earths, and practically all of the uranium originally present. Thorium and rare earth values are readily precipitated from such solution, and the uranium values thus isolated, by the addition of an excess hydrogen peroxide. The pH value of the solution is preferably adjusted to at least 9 prior to the addition of the peroxide.

Innovative Elution Processes for Recovering Uranium and Transition Metals from Amidoxime-based Adsorbents

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

Wai, Chien M.

Amidoxime-based polymer fibers are considered one of the most promising materials for sequestering uranium from seawater. The high-surface-area polymer fibers containing amidoxime and carboxylate groups synthesized by Oak Ridge National Lab (ORNL-AF1) show very high uranium adsorption capacities known in the literature. Effective elution of uranium and repeated use of the adsorbent are important factors affecting the cost of producing uranium from seawater using this material. Traditional acid leaching of uranium followed by KOH conditioning of the fiber causes chemical changes and physical damage to the ORNL-AF1 adsorbent. Two alkaline solution leaching methods were developed by this project, one usesmore » a highly concentrated (3 M) potassium bicarbonate solution at pH 8.3 and 40 °C; the other uses a mixture of sodium carbonate and hydrogen peroxide at pH 10.4. Both elution methods do not require KOH conditioning prior to reusing the fiber adsorbent. The conditions of eluting uranium from the amidoxime-based adsorbent using these alkaline solutions are confirmed by thermodynamic calculations. The bicarbonate elution method is selective for uranium recovery compared to other elution methods and causes no chemical change to the fiber material based on FTIR spectroscopy« less

Resource Conservation and Recovery Act (RCRA) Part B Permit Application for Production Associated Units at the Oak Ridge Y-12 Plant

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

Not Available

This is the RCRA required permit application for Radioactive and Hazardous Waste Management at the Oak Ridge Y-12 Plant for the following units: Building 9206 Container Storage Unit; Building 9212 Container Storage Unit; Building 9720-12 Container Storage Unit; Cyanide Treatment Unit. All four of these units are associated with the recovery of enriched uranium and other metals from wastes generated during the processing of nuclear materials.

Synthesis, Development, and Testing of High-Surface-Area Polymer-Based Adsorbents for the Selective Recovery of Uranium from Seawater

DOE PAGES

Oyola, Yatsandra; Janke, Christopher J.; Dai, Sheng

2016-02-29

The ocean contains uranium with an approximate concentration of 3.34 ppb, which can serve as an incredible supply source to sustain nuclear energy in the United States. Unfortunately, technology currently available to recover uranium from seawater is not efficient enough and mining uranium on land is still more economical. For this study, we have developed polymer-based adsorbents with high uranium adsorption capacities by grafting amidoxime onto high-surface-area polyethylene (PE) fibers. Various process conditions have been screened, in combination with developing a rapid testing protocol (<24 h), to optimize the process. These adsorbents are synthesized through radiation-induced grafting of acrylonitrile (AN)more » and methacrylic acid (MAA) onto PE fibers, followed by the conversion of nitriles to amidoximes and basic conditioning. In addition, the uranium adsorption capacity, measured in units of g U/kg ads, is greatly increased by reducing the diameter of the PE fiber or changing its morphology. An increase in the surface area of the PE polymer fiber allows for more grafting sites that are positioned in more-accessible locations, thereby increasing access to grafted molecules that would normally be located in the interior of a fiber with a larger diameter. Polymer fibers with hollow morphologies are able to adsorb beyond 1 order of magnitude more uranium from simulated seawater than current commercially available adsorbents. Finally, several high-surface-area fibers were tested in natural seawater and were able to extract 5–7 times more uranium than any adsorbent reported to date.« less

Cellular localization of uranium in the renal proximal tubules during acute renal uranium toxicity.

PubMed

Homma-Takeda, Shino; Kitahara, Keisuke; Suzuki, Kyoko; Blyth, Benjamin J; Suya, Noriyoshi; Konishi, Teruaki; Terada, Yasuko; Shimada, Yoshiya

2015-12-01

Renal toxicity is a hallmark of uranium exposure, with uranium accumulating specifically in the S3 segment of the proximal tubules causing tubular damage. As the distribution, concentration and dynamics of accumulated uranium at the cellular level is not well understood, here, we report on high-resolution quantitative in situ measurements by high-energy synchrotron radiation X-ray fluorescence analysis in renal sections from a rat model of uranium-induced acute renal toxicity. One day after subcutaneous administration of uranium acetate to male Wistar rats at a dose of 0.5 mg uranium kg(-1) body weight, uranium concentration in the S3 segment of the proximal tubules was 64.9 ± 18.2 µg g(-1) , sevenfold higher than the mean renal uranium concentration (9.7 ± 2.4 µg g(-1) ). Uranium distributed into the epithelium of the S3 segment of the proximal tubules and highly concentrated uranium (50-fold above mean renal concentration) in micro-regions was found near the nuclei. These uranium levels were maintained up to 8 days post-administration, despite more rapid reductions in mean renal concentration. Two weeks after uranium administration, damaged areas were filled with regenerating tubules and morphological signs of tissue recovery, but areas of high uranium concentration (100-fold above mean renal concentration) were still found in the epithelium of regenerating tubules. These data indicate that site-specific accumulation of uranium in micro-regions of the S3 segment of the proximal tubules and retention of uranium in concentrated areas during recovery are characteristics of uranium behavior in the kidney. Copyright © 2015 John Wiley & Sons, Ltd.

Geochemical data from groundwater at the proposed Dewey Burdock uranium in-situ recovery mine, Edgemont, South Dakota

USGS Publications Warehouse

Johnson, Raymond H.

2012-01-01

This report releases groundwater geochemistry data from samples that were collected in June 2011 at the Dewey Burdock proposed uranium in-situ recovery site near Edgemont, South Dakota. The sampling and analytical methods are summarized, and all of the data, including quality assurance/quality control information are provided in data tables.

78 FR 56944 - Strata Energy, Inc. (Ross In Situ Recovery Uranium Project); Notice of Atomic Safety and...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-09-16

..., Inc. (Ross In Situ Recovery Uranium Project); Notice of Atomic Safety and Licensing Board Reconstitution Pursuant to 10 CFR 2.313(c) and 2.321(b), the Atomic Safety and Licensing Board (Board) in the... Craig M. White to serve on the Board in place of Administrative Judge Kenneth L. Mossman. All...

Potential Aquifer Vulnerability in Regions Down-Gradient from Uranium In Situ Recovery (ISR) Sites

EPA Science Inventory

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 rock...

Recycle of enriched Mo targets for economic production of 99Mo/ 99mTc medical isotope without use of enriched uranium

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

Tkac, Peter; Vandegrift, George F.

2015-08-09

A new recycle process for recovery of enriched 98Mo or 100Mo used for production of 99Mo/ 99mTc medical isotope was developed. In this process, Mo is precipitated from spent NorthStar Mo/Tc generator solution containing ~200 g/L Mo as K 2MoO 4 in 5 M KOH using acetic acid and then washed with nitric acid. High purification factors from potassium were achieved, and typical Mo recovery yields were ~95 %. In conclusion, the recycle process was performed with up to 260 g of Mo per batch and can be easily implemented for processing of up to 400 g of Mo.

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

Reclamation with Recovery of Radionuclides and Toxic Metals from Contaminated Materials, Soils, and Wastes

NASA Technical Reports Server (NTRS)

Francis, A. J.; Dodge, C. J.

1993-01-01

A process has been developed at Brookhaven National Laboratory (BNL) for the removal of metals and radionuclides from contaminated materials, soils, and waste sites. In this process, citric acid, a naturally occurring organic complexing agent, is used to extract metals such as Ba, Cd, Cr, Ni, Zn, and radionuclides Co, Sr, Th, and U from solid wastes by formation of water soluble, metal-citrate complexes. Citric acid forms different types of complexes with the transition metals and actinides, and may involve formation of a bidentate, tridentate, binuclear, or polynuclear complex species. The extract containing radionuclide/metal complex is then subjected to microbiological degradation followed by photochemical degradation under aerobic conditions. Several metal citrate complexes are biodegraded, and the metals are recovered in a concentrated form with the bacterial biomass. Uranium forms binuclear complex with citric acid and is not biodegraded. The supernatant containing uranium citrate complex is separated and upon exposure to light, undergoes rapid degradation resulting in the formation of an insoluble, stable polymeric form of uranium. Uranium is recovered as a precipitate (polyuranate) in a concentrated form for recycling or for appropriate disposal. This treatment process, unlike others which use caustic reagents, does not create additional hazardous wastes for disposal and causes little damage to soil which can then be returned to normal use.

Aqueous biphasic extraction of uranium and thorium from contaminated soils. Final report

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

Chaiko, D.J.; Gartelmann, J.; Henriksen, J.L.

1995-07-01

The aqueous biphasic extraction (ABE) process for soil decontamination involves the selective partitioning of solutes and fine particulates between two immiscible aqueous phases. The biphase system is generated by the appropriate combination of a water-soluble polymer (e.g., polyethlene glycol) with an inorganic salt (e.g., sodium carbonate). Selective partitioning results in 99 to 99.5% of the soil being recovered in the cleaned-soil fraction, while only 0.5 to 1% is recovered in the contaminant concentrate. The ABE process is best suited to the recovery of ultrafine, refractory material from the silt and clay fractions of soils. During continuous countercurrent extraction tests withmore » soil samples from the Fernald Environmental Management Project site (Fernald, OH), particulate thorium was extracted and concentrated between 6- and 16-fold, while the uranium concentration was reduced from about 500 mg/kg to about 77 mg/kg. Carbonate leaching alone was able to reduce the uranium concentration only to 146 mg/kg. Preliminary estimates for treatment costs are approximately $160 per ton of dry soil. A detailed flowsheet of the ABE process is provided.« less

Recovery of fissile materials from nuclear wastes

DOEpatents

Forsberg, Charles W.

1999-01-01

A process for recovering fissile materials such as uranium, and plutonium, and rare earth elements, from complex waste feed material, and converting the remaining wastes into a waste glass suitable for storage or disposal. The waste feed is mixed with a dissolution glass formed of lead oxide and boron oxide resulting in oxidation, dehalogenation, and dissolution of metal oxides. Carbon is added to remove lead oxide, and a boron oxide fusion melt is produced. The fusion melt is essentially devoid of organic materials and halogens, and is easily and rapidly dissolved in nitric acid. After dissolution, uranium, plutonium and rare earth elements are separated from the acid and recovered by processes such as PUREX or ion exchange. The remaining acid waste stream is vitrified to produce a waste glass suitable for storage or disposal. Potential waste feed materials include plutonium scrap and residue, miscellaneous spent nuclear fuel, and uranium fissile wastes. The initial feed materials may contain mixtures of metals, ceramics, amorphous solids, halides, organic material and other carbon-containing material.

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...

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.

Recovery of Uranium from Sodium Carbonate Solutions with Dowex I Anion Exchangers; RECUPERACION DEL URANIO DE SOLUCIONES. DE CARBONATO SODICO MEDIANTE INTERCAMBIO ANIONICO CON RESINA DOWEX I

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

Urgell, M.; Bustamante, J.A.P.; Rodriguez, T.B.

1959-01-01

Synthetic carbonate solutions were used to study the recovery of uranium with Dowex-1. In the first part of the study the capacity of the resin for uranium fixation is investigated as a function of the solution concentration, and the flow velocity. The ideal thickness of the resin bed is investigated. The optimum conditions for elution with various salts and salt mixtures are established. In the last pant, the interference caused by an excess of carbonate and the presence of aluminate or phosphate is considered. (J.S.R.)

78 FR 31985 - Atomic Safety and Licensing Board Panel; Before the Licensing Board: G. Paul Bollwerk, III...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-05-28

... authorizes the operation of CBR's existing in situ uranium recovery (ISR) facility near Crawford, Nebraska... operate a satellite ISR facility, the Marsland Expansion Area (MEA) site, which is located in Dawes County, Nebraska, some eleven miles to the southeast of CBR's Crawford central processing facility. In response to...

Molybdenum isotope fractionation during acid leaching of a granitic uranium ore

NASA Astrophysics Data System (ADS)

Migeon, Valérie; Bourdon, Bernard; Pili, Eric; Fitoussi, Caroline

2018-06-01

As an attempt to prevent illicit trafficking of nuclear materials, it is critical to identify the origin and transformation of uranium materials from the nuclear fuel cycle based on chemical and isotope tracers. The potential of molybdenum (Mo) isotopes as tracers is considered in this study. We focused on leaching, the first industrial process used to release uranium from ores, which is also known to extract Mo depending on chemical conditions. Batch experiments were performed in the laboratory with pH ranging from 0.3 to 5.5 in sulfuric acid. In order to span a large range in uranium and molybdenum yields, oxidizers such as nitric acid, hydrogen peroxide and manganese dioxide were also added. An enrichment in heavy Mo isotopes is produced in the solution during leaching of a granitic uranium ore, when Mo recovery is not quantitative. At least two Mo reservoirs were identified in the ore: ∼40% as Mo oxides soluble in water or sulfuric acid, and ∼40% of Mo hosted in sulfides soluble in nitric acid or hydrogen peroxide. At pH > 1.8, adsorption and/or precipitation processes induce a decrease in Mo yields with time correlated with large Mo isotope fractionations. Quantitative models were used to evaluate the relative importance of the processes involved in Mo isotope fractionation: dissolution, adsorption, desorption, precipitation, polymerization and depolymerization. Model best fits are obtained when combining the effects of dissolution/precipitation, and adsorption/desorption onto secondary minerals. These processes are inferred to produce an equilibrium isotope fractionation, with an enrichment in heavy Mo isotopes in the liquid phase and in light isotopes in the solid phase. Quantification of Mo isotope fractionation resulting from uranium leaching is thus a promising tool to trace the origin and transformation of nuclear materials. Our observations of Mo leaching are also consistent with observations of natural Mo isotope fractionation taking place during chemical weathering in terrestrial environments where the role of secondary processes such as adsorption is significant.

PROCESS FOR THE RECOVERY OF METALS FROM HIGH-LIME CARNOTITE ORES

DOEpatents

Grinstead, R.R.

1959-01-20

A process is presented for recovering uranium values from a high-lime carnotite ore comprising contacting the ore dispersed in a finely divided state with a concentrated mineral acid, adding an industrial orgnnic solvent containing alkyl ontho or pyro phosphoric acids, alkyl phosphates or alkyl phosphonates so as to effect an organic phase into which the metal value is leached and then recovering the metal value from the organic phase.

77 FR 49825 - Notice of Availability of the Final Environmental Impact Statement for the Lost Creek Uranium In...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-08-17

...-166318] Notice of Availability of the Final Environmental Impact Statement for the Lost Creek Uranium In Situ Recovery Project in Sweetwater County, WY AGENCY: Bureau of Land Management, Interior. ACTION... (BLM) has prepared a Final Environmental Impact Statement (EIS) for the Lost Creek Uranium In Situ...

75 FR 76050 - Notice of the Nuclear Regulatory Commission Consent to Indirect Change of Control and Issuance of...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-12-07

... Ranch in situ recovery (ISR) project in Johnson and Campbell Counties, Wyoming. The project is currently in operating status, but is not producing uranium at this time. Materials License SUA-1569 authorizes Uranium One Americas, Inc., to possess uranium and byproduct material at its Moore Ranch ISR Project in...

RECOVERY OF URANIUM FROM AQUEOUS PHOSPHATE-CONTAINING SOLUTIONS

DOEpatents

Igelsrud, I.; Stephen, E.F.

1959-08-11

ABS>A method is presented for recovering hexavalent uranium from an acidic phosphaie solution. A high molecular weight amine, such as a mixture of cccoanut oil amines, is added to the solution in such amount as to give a ratio of about 2000 parts by weight of amine to 1 part by weight of uranium. The uranium is precipitated with the amines and the whole filtered from the solution. The uranium is leached from the amine mass by washing with aqueous sodium carbonate solution; and the amine mixture is available for reuse.

SUMMARY TECHNICAL REPORT FOR THE PERIOD JANUARY 1, 1961-MARCH 31, 1961

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

Burgett, R. ed

1961-05-01

Uranium and TBP Recovery from Waste Solvent. Laboratory and pilot-scale tests were carried out which demonstrated (1) that uranium in waste solvent can be removed by slurrying the solvent with activated charcoal, filtering the slurry, and washing the slurry with water and 3% Na/sub 2/CO/sub 3/ and (2) that TBP can be recovered from the waste solvent by splitting the solvent with HCl and distilling the TBP-rich phase. Improvement of Green Salt Quality. Denitration of ammonium uranyl trinitrate yielded a light, finely divided form of gamma -UO/ sub 3/ with a surface area higher than that of conventional batch potmore » powder; however, its reactivity in reduction and hydrofluorination tests was only moderately higher in comparison. Oxidation-reduction cycles were found to increase the reactivity of UO/sub 2/ toward hydrofluorination. The properties of various UO/sub 2/ samples were determined and correlated with the preparative methods used. Dehydration of Winlo Green Salt. About 27 tons of Winlo green salt was successfully dehydrated to a water content of -0.04% in a hydrofluorination reactor bank in the Green Salt Plant. Recovery of Uranium from MgF/sub 2/ Slag. A process for continuously digesting MgF/sub 2/ slag for uranium recovery was successfully tested on a plant scale. In this process, a water slurry of slag is transferred at a fixed rate and reacted with HCl, and the controlled feed rate reduces the hydrogen concentration. Graphite Liner for Bomb Reduction of Green Salt. An evaluation was made on machined graphite as a replacement for jolt-packed MgF/sub 2/ presently used to line reduction vessels for uranium metal production. Best results were obtained with a onepiece graphite liner fitted inside a steel vessel with an annulus of MgF/sub 2/ between liner and pot. Effects of Feed Material on Ingot Chemical Purity and Yields. The effects of various types of uranium feed stock on the chemical purity and yield of ingots were studied. The following results were obtained: (1) The H content was higher in ingots cast from melts contairing more derby material, (2) the O, N, and C contents of samples from ingot tops were signiicantly lower than those from ingot bottoms, (3) the crude ingot yields were lowest for pigots, briquettes, and heat-shocked grade III derbies, (4) pigots were deleterious to ingot chemical purity, (5) degreased drip crops and dingot extrnsion scrap were deleterious to core-to-good-core yield. Alpha Annealing of Uranium. The effect of a high alpha temperature anneal on the structure and growin index of beta heat treated uranium was evaluated. It was found that longer alpha annealing times gave greater recrystallization and that higher temperatures gave more rapid recrystallization. Delays of up to 6 months between beta heat treatment and alpha anneal did not affect either the recrystallization or the growth index. Billet Drilling. A LeBlond-Carlstedt Rapid Borer was tested as a urarium billet drilling machine and found to give satisfactory results, although some tool breakage occurred. (D.L.C.)« less

Solid-phase data from cores at the proposed Dewey Burdock uranium in-situ recovery mine, near Edgemont, South Dakota

USGS Publications Warehouse

Johnson, Raymond H.; Diehl, Sharon F.; Benzel, William M.

2013-01-01

This report releases solid-phase data from cores at the proposed Dewey Burdock uranium in-situ recovery site near Edgemont, South Dakota. These cores were collected by Powertech Uranium Corporation, and material not used for their analyses were given to the U.S. Geological Survey for additional sampling and analyses. These additional analyses included total carbon and sulfur, whole rock acid digestion for major and trace elements, 234U/238U activity ratios, X-ray diffraction, thin sections, scanning electron microscopy analyses, and cathodoluminescence. This report provides the methods and data results from these analyses along with a short summary of observations.

Method development and validation for simultaneous determination of IEA-R1 reactor’s pool water uranium and silicon content by ICP OES

NASA Astrophysics Data System (ADS)

Ulrich, J. C.; Guilhen, S. N.; Cotrim, M. E. B.; Pires, M. A. F.

2018-03-01

IPEN’s research reactor, IEA-R1, an open pool type research reactor moderated and cooled by light water. High quality water is a key factor in preventing the corrosion of the spent fuel stored in the pool. Leaching of radionuclides from the corroded fuel cladding may be prevented by an efficient water treatment and purification system. However, as a safety management policy, IPEN has adopted a water chemistry control which periodically monitors the levels of uranium (U) and silicon (Si) in the pool’s reactor, since IEA-R1 employs U3Si2-Al dispersion fuel. An analytical method was developed and validated for the determination of uranium and silicon by ICP OES. This work describes the validation process, in a context of quality assurance, including the parameters selectivity, linearity, quantification limit, precision and recovery.

The thermodynamics of pyrochemical processes for liquid metal reactor fuel cycles

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

Johnson, I.

1987-01-01

The thermodynamic basis for pyrochemical processes for the recovery and purification of fuel for the liquid metal reactor fuel cycle is described. These processes involve the transport of the uranium and plutonium from one liquid alloy to another through a molten salt. The processes discussed use liquid alloys of cadmium, zinc, and magnesium and molten chloride salts. The oxidation-reduction steps are done either chemically by the use of an auxiliary redox couple or electrochemically by the use of an external electrical supply. The same basic thermodynamics apply to both the salt transport and the electrotransport processes. Large deviations from idealmore » solution behavior of the actinides and lanthanides in the liquid alloys have a major influence on the solubilities and the performance of both the salt transport and electrotransport processes. Separation of plutonium and uranium from each other and decontamination from the more noble fission product elements can be achieved using both transport processes. The thermodynamic analysis is used to make process design computations for different process conditions.« 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

METHOD OF SEPARATING URANIUM FROM ALLOYS

DOEpatents

Chiotti, P.; Shoemaker, H.E.

1960-06-28

Uranium can be recovered from metallic uraniumthorium mixtures containing uranium in comparatively small amounts. The method of recovery comprises adding a quantity of magnesium to a mass to obtain a content of from 48 to 85% by weight; melting and forming a magnesium-thorium alloy at a temperature of between 585 and 800 deg C; agitating the mixture, allowing the mixture to settle whereby two phases, a thorium-containing magnesium-rich liquid phase and a solid uranium-rich phase, are formed; and separating the two phases.

Recovery of uranium values

DOEpatents

Brown, K. B.; Crouse, Jr., D. J.; Moore, J. G.

1959-03-10

A liquid-liquid extraction method is presented for recovering uranium values from an aqueous acidic solution by means of certain high molecular weight amine fn the amine classes of primary, secondary, heterocyclic secondary, tertiary, or heterocyclic tertiary. The uranium bearing aqueous acidic solution is contacted with the selected anine dissolved in a nonpolar waterimmiscible organfc solvent such as kerosene. The uranium which is substantially completely extracted by the organic phase may be stripped therefrom by water, and recovered from the aqueous phase by treatment into ammonia to precipitate ammonium diuranate.

RECOVERY OF URANIUM VALUES

DOEpatents

Brown, K.B.; Crouse, D.J. Jr.; Moore, J.G.

1959-03-10

A liquid-liquid extraction method is presented for recovering uranium values from an aqueous acidic solution by means of certain high molecular weight amine in the amine classes of primary, secondary, heterocyclic secondary, tertiary, or heterocyclic tertiary. The uranium bearing aqueous acidic solution is contacted with the selected amine dissolved in a nonpolar water-immiscible organic solvent such as kerosene. The uranium which is substantially completely exiracted by the organic phase may be stripped therefrom by waters and recovered from the aqueous phase by treatment into ammonia to precipitate ammonium diuranate.

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.

Removal of uranium from soil sample digests for ICP-OES analysis of trace metals

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

Foust, R.D. Jr.; Bidabad, M.

1996-10-01

An analytical procedure has been developed to quantitatively remove uranium from soil sample digests, permitting ICP-OES analysis of trace metals. The procedure involves digesting a soil sample with standard procedures (EPA SW-846, Method 3050), and passing the sample digestate through commercially available resin (U/TEVA{sm_bullet}Spec, Eichrom Industries, Inc.) containing diarryl amylphosphonate as the stationary phase. Quantitative removal of uranium was achieved with soil samples containing up to 60% uranium, and percent recoveries averaged better than 85% for 9 of the 10 metals evaluated (Ag, As, Cd. Cr, Cu, Ni, Pb, Se and Tl). The U/TEVA{sm_bullet}Spec column was regenerated by washing withmore » 200 mL of a 0.01 M oxalic acid/0.02 M nitric acid solution, permitting re-use of the column. GFAAS analysis of a sample spiked with 56.5% uranium, after treatment of the digestate with a U/TEVA{sm_bullet}Spec resin column, resulted in percent recoveries of 97% or better for all target metals.« less

RECOVERY OF URANIUM BY SECONDARY XANTHATE COMPLEXING

DOEpatents

Neville, O.K.

1959-09-01

A method is described for separating and recovering uranium values contained in an acidic aqueous solution together with thorium or protactinium values. In accordance with the invention, the acidic solution containing uranium in the uranyl form is contacted with an organic xanthate. The xanthate forms a urano-xanthate complex but is substantially non-reactive with thorium and protactinium. The urano-xanthate complex is recovered by organic solvent extraction.

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.

Laboratory column experiments and transport modeling to evaluate retardation of uranium in an aquifer downgradient of a uranium in-situ recovery site

DOE PAGES

Dangelmayr, Martin A.; Reimus, Paul W.; Wasserman, Naomi L.; ...

2017-05-01

The purpose of this study was to determine the attenuation potential and retardation of uranium in sediments taken from boreholes at the Smith-Ranch Highland in-situ recovery (ISR) site. Five column experiments with four different sediments were conducted to study the effects of variable mineralogy and alkalinity on uranium breakthrough. Uranium transport was modeled with PHREEQC using a generalized composite surface complexation model (GC SCM) with one, two, and, three generic surfaces, respectively. Reactive surface areas were approximated with PEST using BET derived surface areas to constrain fitting parameters. Uranium breakthrough was delayed by a factor of 1.68, 1.69 and 1.47more » relative to the non-reactive tracer for three of the 5 experiments at an alkalinity of 540 mg/l. A sediment containing smectite and kaolinite retained uranium by a factor of 2.80 despite a lower measured BET surface area. Decreasing alkalinity to 360 mg/l from 540 mg/l increased retardation by a factor of 4.26. Model fits correlated well to overall BET surface area in the three columns where clay content was less than 1%. For the sediment with clay, models consistently understated uranium retardation when reactive surface sites were restricted by BET results. Calcite saturation was shown to be a controlling factor for uranium desorption as the pH of the system changes. A pH of 6 during a secondary background water flush remobilized previously sorbed uranium resulting in a secondary uranium peak at twice the influent concentrations. Furthermore, this study demonstrates the potential of GC SCM models to predict uranium transport in sediments with homogenous mineral composition, but highlights the need for further research to understand the role of sediment clay composition and calcite saturation in uranium transport.« less

Laboratory column experiments and transport modeling to evaluate retardation of uranium in an aquifer downgradient of a uranium in-situ recovery site

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

Dangelmayr, Martin A.; Reimus, Paul W.; Wasserman, Naomi L.

The purpose of this study was to determine the attenuation potential and retardation of uranium in sediments taken from boreholes at the Smith-Ranch Highland in-situ recovery (ISR) site. Five column experiments with four different sediments were conducted to study the effects of variable mineralogy and alkalinity on uranium breakthrough. Uranium transport was modeled with PHREEQC using a generalized composite surface complexation model (GC SCM) with one, two, and, three generic surfaces, respectively. Reactive surface areas were approximated with PEST using BET derived surface areas to constrain fitting parameters. Uranium breakthrough was delayed by a factor of 1.68, 1.69 and 1.47more » relative to the non-reactive tracer for three of the 5 experiments at an alkalinity of 540 mg/l. A sediment containing smectite and kaolinite retained uranium by a factor of 2.80 despite a lower measured BET surface area. Decreasing alkalinity to 360 mg/l from 540 mg/l increased retardation by a factor of 4.26. Model fits correlated well to overall BET surface area in the three columns where clay content was less than 1%. For the sediment with clay, models consistently understated uranium retardation when reactive surface sites were restricted by BET results. Calcite saturation was shown to be a controlling factor for uranium desorption as the pH of the system changes. A pH of 6 during a secondary background water flush remobilized previously sorbed uranium resulting in a secondary uranium peak at twice the influent concentrations. Furthermore, this study demonstrates the potential of GC SCM models to predict uranium transport in sediments with homogenous mineral composition, but highlights the need for further research to understand the role of sediment clay composition and calcite saturation in uranium transport.« less

A Nuclear Reactor and Chemical Processing Design for Production of Molybdenum-99 with Crystalline Uranyl Nitrate Hexahydrate Fuel

NASA Astrophysics Data System (ADS)

Stange, Gary Michael

Medical radioisotopes are used in tens of millions of procedures every year to detect and image a wide variety of maladies and conditions in the human body. The most widely-used diagnostic radioisotope is technetium-99m, a metastable isomer of technetium-99 that is generated by the radioactive decay of molybdenum-99. For a number of reasons, the supply of molybdenum-99 has become unreliable and the techniques used to produce it have become unattractive. This has spurred the investigation of new technologies that avoid the use of highly enriched uranium to produce molybdenum-99 in the United States, where approximately half of the demand originates. The first goal of this research is to develop a critical nuclear reactor design powered by solid, discrete pins of low enriched uranium. Analyses of single-pin heat transfer and whole-core neutronics are performed to determine the required specifications. Molybdenum-99 is produced directly in the fuel of this reactor and then extracted through a series of chemical processing steps. After this extraction, the fuel is left in an aqueous state. The second goal of this research is to describe a process by which the uranium may be recovered from this spent fuel solution and reconstituted into the original fuel form. Fuel recovery is achieved through a crystallization step that generates solid uranyl nitrate hexahydrate while leaving the majority of fission products and transuranic isotopes in solution. This report provides background information on molybdenum-99 production and crystallization chemistry. The previously unknown thermal conductivity of the fuel material is measured. Following this is a description of the modeling and calculations used to develop a reactor concept. The operational characteristics of the reactor core model are analyzed and reported. Uranyl nitrate crystallization experiments have also been conducted, and the results of this work are presented here. Finally, a process flow scheme for uranium recovery is examined, in part qualitatively and in part quantitatively, based upon the preceding data garnered through literature review, modeling, and experimentation. The sum of this research is meant to allow for a complete understanding of the process flow, from the beginning of one production cycle to the beginning of another.

An evaluation of health risk to the public as a consequence of in situ uranium mining in Wyoming, USA

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

Ruedig, Elizabeth; Johnson, Thomas E.

In the United States there is considerable public concern regarding the health effects of in situ recovery uranium mining. These concerns focus principally on exposure to contaminants mobilized in groundwater by the mining process. However, the risk arising as a result of mining must be viewed in light of the presence of naturally occurring uranium ore and other constituents which comprise a latent hazard. The United States Environmental Protection Agency recently proposed new guidelines for successful restoration of an in situ uranium mine by limiting concentrations of thirteen groundwater constituents: arsenic, barium, cadmium, chromium, lead, mercury, selenium, silver, nitrate (asmore » nitrogen), molybdenum, radium, total uranium, and gross α activity. We investigated the changes occurring to these constituents at an ISR uranium mine in Wyoming, USA by comparing groundwater quality at baseline measurement to that at stability (post-restoration) testing. Of the groundwater constituents considered, only uranium and radium-226 showed significant (p < 0.05) deviation from site-wide baseline conditions in matched-wells. Uranium concentrations increased by a factor of 5.6 (95% CI 3.6–8.9 times greater) while radium-226 decreased by a factor of about one half (95% CI 0.42–0.75 times less). Change in risk was calculated using the RESRAD (onsite) code for an individual exposed as a resident-farmer; total radiation dose to a resident farmer decreased from pre-to post-mining by about 5.2 mSv y –1. As a result, higher concentrations of uranium correspond to increased biomarkers of nephrotoxicity, however the clinical significance of this increase is unclear.« less

An evaluation of health risk to the public as a consequence of in situ uranium mining in Wyoming, USA

DOE PAGES

Ruedig, Elizabeth; Johnson, Thomas E.

2015-08-30

In the United States there is considerable public concern regarding the health effects of in situ recovery uranium mining. These concerns focus principally on exposure to contaminants mobilized in groundwater by the mining process. However, the risk arising as a result of mining must be viewed in light of the presence of naturally occurring uranium ore and other constituents which comprise a latent hazard. The United States Environmental Protection Agency recently proposed new guidelines for successful restoration of an in situ uranium mine by limiting concentrations of thirteen groundwater constituents: arsenic, barium, cadmium, chromium, lead, mercury, selenium, silver, nitrate (asmore » nitrogen), molybdenum, radium, total uranium, and gross α activity. We investigated the changes occurring to these constituents at an ISR uranium mine in Wyoming, USA by comparing groundwater quality at baseline measurement to that at stability (post-restoration) testing. Of the groundwater constituents considered, only uranium and radium-226 showed significant (p < 0.05) deviation from site-wide baseline conditions in matched-wells. Uranium concentrations increased by a factor of 5.6 (95% CI 3.6–8.9 times greater) while radium-226 decreased by a factor of about one half (95% CI 0.42–0.75 times less). Change in risk was calculated using the RESRAD (onsite) code for an individual exposed as a resident-farmer; total radiation dose to a resident farmer decreased from pre-to post-mining by about 5.2 mSv y –1. As a result, higher concentrations of uranium correspond to increased biomarkers of nephrotoxicity, however the clinical significance of this increase is unclear.« less

An evaluation of health risk to the public as a consequence of in situ uranium mining in Wyoming, USA.

PubMed

Ruedig, Elizabeth; Johnson, Thomas E

2015-12-01

In the United States there is considerable public concern regarding the health effects of in situ recovery uranium mining. These concerns focus principally on exposure to contaminants mobilized in groundwater by the mining process. However, the risk arising as a result of mining must be viewed in light of the presence of naturally occurring uranium ore and other constituents which comprise a latent hazard. The United States Environmental Protection Agency recently proposed new guidelines for successful restoration of an in situ uranium mine by limiting concentrations of thirteen groundwater constituents: arsenic, barium, cadmium, chromium, lead, mercury, selenium, silver, nitrate (as nitrogen), molybdenum, radium, total uranium, and gross α activity. We investigated the changes occurring to these constituents at an ISR uranium mine in Wyoming, USA by comparing groundwater quality at baseline measurement to that at stability (post-restoration) testing. Of the groundwater constituents considered, only uranium and radium-226 showed significant (p < 0.05) deviation from site-wide baseline conditions in matched-wells. Uranium concentrations increased by a factor of 5.6 (95% CI 3.6-8.9 times greater) while radium-226 decreased by a factor of about one half (95% CI 0.42-0.75 times less). Change in risk was calculated using the RESRAD (onsite) code for an individual exposed as a resident-farmer; total radiation dose to a resident farmer decreased from pre-to post-mining by about 5.2 mSv y(-1). Higher concentrations of uranium correspond to increased biomarkers of nephrotoxicity, however the clinical significance of this increase is unclear. Published by Elsevier Ltd.

Behavior and Distribution of Heavy Metals Including Rare Earth Elements, Thorium, and Uranium in Sludge from Industry Water Treatment Plant and Recovery Method of Metals by Biosurfactants Application

PubMed Central

Gao, Lidi; Kano, Naoki; Sato, Yuichi; Li, Chong; Zhang, Shuang; Imaizumi, Hiroshi

2012-01-01

In order to investigate the behavior, distribution, and characteristics of heavy metals including rare earth elements (REEs), thorium (Th), and uranium (U) in sludge, the total and fractional concentrations of these elements in sludge collected from an industry water treatment plant were determined and compared with those in natural soil. In addition, the removal/recovery process of heavy metals (Pb, Cr, and Ni) from the polluted sludge was studied with biosurfactant (saponin and sophorolipid) elution by batch and column experiments to evaluate the efficiency of biosurfactant for the removal of heavy metals. Consequently, the following matters have been largely clarified. (1) Heavy metallic elements in sludge have generally larger concentrations and exist as more unstable fraction than those in natural soil. (2) Nonionic saponin including carboxyl group is more efficient than sophorolipid for the removal of heavy metals in polluted sludge. Saponin has selectivity for the mobilization of heavy metals and mainly reacts with heavy metals in F3 (the fraction bound to carbonates) and F5 (the fraction bound to Fe-Mn oxides). (3) The recovery efficiency of heavy metals (Pb, Ni, and Cr) reached about 90–100% using a precipitation method with alkaline solution. PMID:22693485

Fission-Produced 99Mo Without a Nuclear Reactor.

PubMed

Youker, Amanda J; Chemerisov, Sergey D; Tkac, Peter; Kalensky, Michael; Heltemes, Thad A; Rotsch, David A; Vandegrift, George F; Krebs, John F; Makarashvili, Vakho; Stepinski, Dominique C

2017-03-01

99 Mo, the parent of the widely used medical isotope 99m Tc, is currently produced by irradiation of enriched uranium in nuclear reactors. The supply of this isotope is encumbered by the aging of these reactors and concerns about international transportation and nuclear proliferation. Methods: We report results for the production of 99 Mo from the accelerator-driven subcritical fission of an aqueous solution containing low enriched uranium. The predominately fast neutrons generated by impinging high-energy electrons onto a tantalum convertor are moderated to thermal energies to increase fission processes. The separation, recovery, and purification of 99 Mo were demonstrated using a recycled uranyl sulfate solution. Conclusion: The 99 Mo yield and purity were found to be unaffected by reuse of the previously irradiated and processed uranyl sulfate solution. Results from a 51.8-GBq 99 Mo production run are presented. © 2017 by the Society of Nuclear Medicine and Molecular Imaging.

Looking Southwest at Reactor Box Furnaces With Reactor Boxes and ...

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

Looking Southwest at Reactor Box Furnaces With Reactor Boxes and Repossessed Uranium in Recycle Recovery Building - Hematite Fuel Fabrication Facility, Recycle Recovery Building, 3300 State Road P, Festus, Jefferson County, MO

Cost and Systems Analysis of Innovative Fuel Resources Concepts

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

Schneider, Erich; Byers, M.

Economically recovered uranium from seawater can have a transformative effect on the way policy makers view the long-term viability of uranium based fuel cycles. Seawater uranium, even when estimated to cost more than terrestrially mined uranium, is integral in establishing an economic backstop, thus reducing uncertainty in future nuclear power costs. While a passive recovery scheme relying on a field of polymer adsorbents prepared via radiation induced grafting has long been considered the leading technology for full scale deployment, non-trivial cost and logistical barriers persist. Consequently, university partners of the nation-wide consortium for seawater uranium recovery have developed variants ofmore » this technology, each aiming to address a substantial weakness. The focus of this NEUP project is the economic impacts of the proposed variant technologies. The team at University of Alabama has pursued an adsorbent synthesis method that replaces the synthetic fiber backbone with a natural waste product. Chitin fibers suitable for ligand grafting have been prepared from shrimp shell waste. These environmental benefits could be realized at a comparable cost to the reference fiber so long as the uptake can be increased or the chemical consumption cost decreased.« less

A graphene oxide/amidoxime hydrogel for enhanced uranium capture

PubMed Central

Wang, Feihong; Li, Hongpeng; Liu, Qi; Li, Zhanshuang; Li, Rumin; Zhang, Hongsen; Liu, Lianhe; Emelchenko, G. A.; Wang, Jun

2016-01-01

The efficient development of selective materials for the recovery of uranium from nuclear waste and seawater is necessary for their potential application in nuclear fuel and the mitigation of nuclear pollution. In this work, a graphene oxide/amidoxime hydrogel (AGH) exhibits a promising adsorption performance for uranium from various aqueous solutions, including simulated seawater. We show high adsorption capacities (Qm = 398.4 mg g−1) and high % removals at ppm or ppb levels in aqueous solutions for uranium species. In the presence of high concentrations of competitive ions such as Mg2+, Ca2+, Ba2+ and Sr2+, AGH displays an enhanced selectivity for uranium. For low uranium concentrations in simulated seawater, AGH binds uranium efficiently and selectively. The results presented here reveal that the AGH is a potential adsorbent for remediating nuclear industrial effluent and adsorbing uranium from seawater. PMID:26758649

SEPARATION OF THORIUM FROM URANIUM BY EXTRACTION

DOEpatents

Bohlmann, E.G.

1959-07-28

A method is presented for the recovery and separation of uranium and thorium values contained in an aqueous nitric acid solution which is more than 3 M in nitric acid. The uranium and thorium containing solution preferable about 7 M in nitric acid is contacted with tributyl phosphatekerosene mixture. Both U and Th are extracted by the immiscible organic. After phase separation the Th is selectively back extracted by contacting with an aqueous nitric acid solution preferably between 0.1 to 1.5 M in nitric acid. The uranium which is still in the organic extractant phase may be recovered by contacting with water.

Progress in Chile in the development of the fission {sup 99}Mo production using modified CINTICHEM

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

Schrader, R.; Klein, J.; Medel, J.

2008-07-15

Fission {sup 99}Mo will be produced in Chile irradiating low-enriched uranium (LEU) foil in a MTR research reactor. For the purpose of developing the capability to fabricate the target, which is done of uranium foil enclosed in swaged concentric aluminum tubes, dummy targets are being fabricated using 130 {mu}m copper foil instead of the uranium foil, wrapped in a 14{mu}m nickel fission-recoil barrier. Dummy targets using several dimensions of copper foil have been assembled; however, the emphasis is being set in targets fabricated using the dimensions of the LEU foil that KAERI will provide, i.e. 50 mm x 100mm xmore » 0.130 mm. The assembling of target using the last dimensions has not been free of difficulties. Neutronic calculations and preliminary thermal and fluid analyses were performed to estimate the fission products activity and the heat removal capability for a 13 grams LEU-foil annular target, which will be irradiated in the RECH-1 research reactor at the level power of 5 MW during 48 hours. In a fume hood, Cintichem processing of natural uranium shavings with the addition of different carriers were performed, obtaining recovery over 90% of the added Mo carrier. Expertise has been gained in (a) foil dissolution process in a dissolver locally designed, (b) in Mo precipitation process, and (c) preparation of the purification columns with AgC, C and HZrO. Additionally, the irradiated target cutting machine with an innovative design was finally assembled. (author)« less

The ultimate mineral processing challenge: Recovery of rare earths, phosphorus and uranium from Florida phosphatic clay

DOE PAGES

Zhang, Patrick; Liang, Haijun; Jin, Zhen; ...

2017-11-01

We report phosphate beneficiation in Florida generates more than one tonne of phosphatic clay, or slime, per tonne of phosphate rock produced. Since the start of the practice of large-scale washing and desliming for phosphate beneficiation, more than 2 Gt of slime has accumulated, containing approximately 600 Mt of phosphate rock, 600 kt of rare earth elements (REEs) and 80 million kilograms of uranium. The recovery of these valuable elements from the phosphatic clay is one of the most challenging endeavors in mineral processing, because the clay is extremely dilute, with an average solids concentration of 3 percent, and finemore » in size, with more than 50 percent having particle size smaller than 2 μm, and it contains nearly 50 percent clay minerals as well as large amounts of magnesium, iron and aluminum. With industry support and under funding from the Critical Materials Institute, the Florida Industrial and Phosphate Research Institute in conjunction with the Oak Ridge National Laboratory undertook the task to recover phosphorus, rare earths and uranium from Florida phosphatic clay. This paper presents the results from the preliminary testing of two approaches. The first approach involves three-stage cycloning using cyclones with diameters of 12.4 cm (5 in.), 5.08 cm (2 in.) and 2.54 cm (1 in.), respectively, to remove clay minerals followed by flotation and leaching. The second approach is a two-step leaching process. In the first step, selective leaching was conducted to remove magnesium, thus allowing the production of phosphoric acid suitable for the manufacture of diammonium phosphate (DAP) in the second leaching step. The results showed that multistage cycloning with small cyclones is necessary to remove clay minerals. Finally, selective leaching at about pH 3.2 using sulfuric acid was found to be effective for removing more than 80 percent of magnesium from the feed with minimal loss of phosphorus.« less

The ultimate mineral processing challenge: Recovery of rare earths, phosphorus and uranium from Florida phosphatic clay

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

Zhang, Patrick; Liang, Haijun; Jin, Zhen

We report phosphate beneficiation in Florida generates more than one tonne of phosphatic clay, or slime, per tonne of phosphate rock produced. Since the start of the practice of large-scale washing and desliming for phosphate beneficiation, more than 2 Gt of slime has accumulated, containing approximately 600 Mt of phosphate rock, 600 kt of rare earth elements (REEs) and 80 million kilograms of uranium. The recovery of these valuable elements from the phosphatic clay is one of the most challenging endeavors in mineral processing, because the clay is extremely dilute, with an average solids concentration of 3 percent, and finemore » in size, with more than 50 percent having particle size smaller than 2 μm, and it contains nearly 50 percent clay minerals as well as large amounts of magnesium, iron and aluminum. With industry support and under funding from the Critical Materials Institute, the Florida Industrial and Phosphate Research Institute in conjunction with the Oak Ridge National Laboratory undertook the task to recover phosphorus, rare earths and uranium from Florida phosphatic clay. This paper presents the results from the preliminary testing of two approaches. The first approach involves three-stage cycloning using cyclones with diameters of 12.4 cm (5 in.), 5.08 cm (2 in.) and 2.54 cm (1 in.), respectively, to remove clay minerals followed by flotation and leaching. The second approach is a two-step leaching process. In the first step, selective leaching was conducted to remove magnesium, thus allowing the production of phosphoric acid suitable for the manufacture of diammonium phosphate (DAP) in the second leaching step. The results showed that multistage cycloning with small cyclones is necessary to remove clay minerals. Finally, selective leaching at about pH 3.2 using sulfuric acid was found to be effective for removing more than 80 percent of magnesium from the feed with minimal loss of phosphorus.« less

76 FR 5216 - Notice of Availability of Final Supplemental Environmental Impact Statement for the Nichols Ranch...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-01-28

... Uranium Recovery Project, located in the Pumpkin Buttes Uranium Mining District within the Powder River.... Alternatives that were considered, but were eliminated from detailed analysis, include conventional mining and... an Agencywide Documents and Management System (ADAMS), which provides text and image files of the NRC...

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.

76 FR 37160 - Notice of Availability of Final Supplemental Environmental Impact Statement for the Lost Creek In...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-06-24

... Environmental Impact Statement for the Lost Creek In-Situ Recovery (ISR) Project in Sweetwater County, WY; Supplement to the Generic Environmental Impact Statement for In-Situ Leach Uranium Milling Facilities AGENCY... Statement (SEIS) to the Generic Environmental Impact Statement for In- Situ Leach Uranium Milling Facilities...

RECOVERY OF URANIUM FROM TUNGSTEN

DOEpatents

Newnam, K.

1959-02-01

A method is presented for the rccovery of uranium which has adhered to tungsten parts in electromagnetic isotope separation apparatus. Such a tungsten article is dissolved electrolytically in 20% NaOH by using the tungsten article as the anode. The resulting solution, containing soluble sodium lungstate and an insoluble slime, is then filtered. The slime residue is ignited successively with sodium nitrate and sodium pyrosulfate and leashed, and the resulting filtrates are combined with the original filtrate. Uranium is then recovered from the combined flltrates by diuranate precipitation.

Groundwater Restoration at Uranium In-Situ Recovery Mines, South Texas Coastal Plain

USGS Publications Warehouse

Hall, Susan

2009-01-01

This talk was presented by U.S. Geological Survey (USGS) geologist Susan Hall on May 11, 2009, at the Uranium 2009 conference in Keystone, Colorado, and on May 12, 2009, as part of an underground injection control track presentation at the Texas Commission on Environmental Quality (TCEQ) Environmental Trade Fair and Conference in Austin, Texas. Texas has been the location of the greatest number of uranium in-situ recovery (ISR) mines in the United States and was the incubator for the development of alkaline leach technology in this country. For that reason, the author chose to focus on the effectiveness of restoration at ISR mines by examining legacy mines developed in Texas. The best source for accurate information about restoration at Texas ISR mines is housed at the TCEQ offices in Austin. The bulk of this research is an analysis of those records.

Bioengineered Chimeric Spider Silk-Uranium Binding Proteins

PubMed Central

Krishnaji, Sreevidhya Tarakkad; Kaplan, David L.

2014-01-01

Heavy metals constitute a source of environmental pollution. Here, novel functional hybrid biomaterials for specific interactions with heavy metals are designed by bioengineering consensus sequence repeats from spider silk of Nephila clavipes with repeats of a uranium peptide recognition motif from a mutated 33-residue of calmodulin protein from Paramecium tetraurelia. The self-assembly features of the silk to control nanoscale organic/inorganic material interfaces provides new biomaterials for uranium recovery. With subsequent enzymatic digestion of the silk to concentrate the sequestered metals, options can be envisaged to use these new chimeric protein systems in environmental engineering, including to remediate environments contaminated by uranium. PMID:23212989

Recovery Act Weekly Video: 200 Area Asbestos Removal, U-Ancillary Demolition, 200 West Transfer Building Footings

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

None

A weekly update of the Recovery Act at work. Demolition of U-Ancillary that was contaminated with uranium and asbestos as well as removing asbestos from the Steam Generation Plant in the 200 East Area.

Recovery Act Weekly Video: 200 Area Asbestos Removal, U-Ancillary Demolition, 200 West Transfer Building Footings

ScienceCinema

None

2018-05-23

A weekly update of the Recovery Act at work. Demolition of U-Ancillary that was contaminated with uranium and asbestos as well as removing asbestos from the Steam Generation Plant in the 200 East Area.

Engineering assessment and feasibility study of Chattanooga Shale as a future source of uranium. [Preliminary mining; data on soils, meteorology, water resources, and biological resources

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

Not Available

This volume contains five appendixes: Chattanooga Shale preliminary mining study, soils data, meteorologic data, water resources data, and biological resource data. The area around DeKalb County in Tennessee is the most likely site for commercial development for recovery of uranium. (DLC)

77 FR 33782 - License Amendment To Construct and Operate New In Situ Leach Uranium Recovery Facility; Uranium...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-06-07

...-Filing system also distributes an email notice that provides access to the document to the NRC's Office... mail, or expedited delivery service to the Office of the Secretary, Sixteenth Floor, One White Flint... courier, express mail, or expedited delivery service upon depositing the document with the provider of the...

Spallation studies on shock loaded uranium

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

Tonks, D.L.; Hixson, R.; Gustavsen, R.L.

1997-12-31

Uranium samples at two different purity levels were used for spall strength measurements at three different stress levels. A 50 mm single-stage gas-gun was used to produce planar impact conditions using Z-cut quartz impactors. Samples of depleted uranium were taken from very high purity material and from material that had 300 ppm of carbon added. A pair of shots was done for each impact strength, one member of the pair with VISAR diagnostics and the second with soft recovery for metallographical examination. A series of increasing final stress states were chosen to effectively freeze the microstructural damage at three placesmore » in the development to full spall separation. This allowed determination of the dependence of spall mechanisms on stress level and sample purity. This report will discuss both the results of the metallurgical examination of soft recovered samples and the modeling of the free surface VISAR data. The micrographs taken from the recovered samples show brittle cracking as the spallation failure mechanism. Deformation induced twins are plentiful and obviously play a role in the spallation process. The twins are produced in the initial shock loading and, so, are present already before the fracture process begins. The 1 d characteristics code CHARADE has been used to model the free surface VISAR data.« less

Comprehensive Evaluation of Soil Near Uranium Tailings, Beishan City, China.

PubMed

Xun, Yan; Zhang, Xinjia; Chaoliang, Chen; Luo, Xuegang; Zhang, Yu

2018-06-01

To evaluate the impact of uranium tailings on soil composition and soil microbial, six soil samples at different distance from the uranium tailings (Beishan City, China) were collected for further analysis. Concentrations of radionuclides ( 238 U and 232 Th), heavy metals (Mn, Cd, Cr, Ni, Zn, and Pb) and organochlorine pesticide were determined by ICP-MS and GC, they were significantly higher than those of the control. And the Average Well Color Development as well as the Shannon, the Evenness, and the Simpson index were calculated to evaluate the soil microbial diversity. The carbon utilization model of soil microbial community was also analyzed by Biolog-eco. All results indicated that uranium tailings leaded to excessive radionuclides and heavy metals, and decreased the diversity of the soil microbial community. Our study will provide a valuable basis for soil quality evaluation around uranium tailing repositories and lay a foundation for the management and recovery of uranium tailings.

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.

CMB-8 material balance system

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

Langner, D.; Canada, T.; Ensslin, N.

1980-08-01

We describe the automated nondestructive assay (NDA) system installed at the Los Alamos Scientific Laboratory (LASL) Group CMB-8 uranium recovery facility. A random driver (RD) is used to measure the /sup 235/U content of various solids while a uranium solution assay system (USAS) measures the /sup 235/U or total uranium content of solutions over a concentration range of a few ppM to 400 g/l. Both instruments are interfaced to and controlled by a single minicomputer. The measurement principles, mechanical specifications, system software description, and operational instructions are described.

Isotopic analysis of uranium in natural waters by alpha spectrometry

USGS Publications Warehouse

Edwards, K.W.

1968-01-01

A method is described for the determination of U234/U238 activity ratios for uranium present in natural waters. The uranium is coprecipitated from solution with aluminum phosphate, extracted into ethyl acetate, further purified by ion exchange, and finally electroplated on a titanium disc for counting. The individual isotopes are determined by measurement of the alpha-particle energy spectrum using a high resolution low-background alpha spectrometer. Overall chemical recovery of about 90 percent and a counting efficiency of 25 percent allow analyses of water samples containing as little as 0.10 ?g/l of uranium. The accuracy of the method is limited, on most samples, primarily by counting statistics.

75 FR 3261 - Powertech (USA) Inc.; Dewey-Burdock Project; New Source Material License Application; Notice of...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-01-20

... in-situ uranium recovery (ISR, also known as in-situ leach) facilities, and would require restoration... Environmental Impact Statement for In-Situ Leach Uranium Milling Facilities (ISR GEIS) that was published in... published in the Federal Register on January 05, 2010 (75 FR 467-471). The purpose of this notice of intent...

13. VIEW OF A BBOX, WHICH WAS USED IN THE ...

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

13. VIEW OF A B-BOX, WHICH WAS USED IN THE FAST RECOVERY PROCESS. URANIUM OXIDE WAS TRANSFERRED FOR DISSOLUTION IN A ROOM WHICH HOUSED 3 ROWS OF B-BOXES. B-BOXES ARE CONTROLLED HOODS, SIMILAR TO LAB HOODS THAT OPERATED WITH HIGH AIR VELOCITIES AT THEIR OPENINGS TO ENSURE THAT THE VAPORS WERE CONTAINED WITHIN THE HOOD. (2/14/79) - Rocky Flats Plant, General Manufacturing, Support, Records-Central Computing, Southern portion of Plant, Golden, Jefferson County, CO

Removal of Trace Elements by Cupric Oxide Nanoparticles from Uranium In Situ Recovery Bleed Water and Its Effect on Cell Viability

PubMed Central

Schilz, Jodi R.; Reddy, K. J.; Nair, Sreejayan; Johnson, Thomas E.; Tjalkens, Ronald B.; Krueger, Kem P.; Clark, Suzanne

2015-01-01

In situ recovery (ISR) is the predominant method of uranium extraction in the United States. During ISR, uranium is leached from an ore body and extracted through ion exchange. The resultant production bleed water (PBW) contains contaminants such as arsenic and other heavy metals. Samples of PBW from an active ISR uranium facility were treated with cupric oxide nanoparticles (CuO-NPs). CuO-NP treatment of PBW reduced priority contaminants, including arsenic, selenium, uranium, and vanadium. Untreated and CuO-NP treated PBW was used as the liquid component of the cell growth media and changes in viability were determined by the MTT (3-(4,5-dimethylthiazol-2-yl)-2,5-diphenyltetrazolium bromide) assay in human embryonic kidney (HEK 293) and human hepatocellular carcinoma (Hep G2) cells. CuO-NP treatment was associated with improved HEK and HEP cell viability. Limitations of this method include dilution of the PBW by growth media components and during osmolality adjustment as well as necessary pH adjustment. This method is limited in its wider context due to dilution effects and changes in the pH of the PBW which is traditionally slightly acidic however; this method could have a broader use assessing CuO-NP treatment in more neutral waters. PMID:26132311

Selective separation of uranium using alizarin red S (ARS)-modified anion-exchange resin or by flotation of U-ARS chelate

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

Khalifa, M.E.

An alizarin red S (ARS)-modified anion exchange resin was prepared by a simple reaction of ARS with the anion exchange Doulite A101 and used for the efficient sorption of uranium from aqueous media. The effect of various parameters on the sorption of U(VI) (pH effect, sorption kinetics, resin capacity and breakthrough curves) was investigated. The modified resin sorbs U(VI) over a wide range of pH (2.8--5) with a maximum sorption capacity of 0.68 mmol/g at pH 3.2 to 4.0. Iron(III), Zr(IV), Ti(IV), Cu(II), and Th(IV) ions are also sorbed to different extents, but Be(II), Bi(III), Ca(II), Mg(II), Pb(II), Hg(II), Zn(II),more » Cd(II), Al(III), Mn(II), Co(II) and Ni(II) are not sorbed; thus, conditions for separating U(VI) from these metal ions have been identified. For eluting U(VI) from the resin, 0.2 mol/L HCl was used and the recovery recorded was as high as 99.9%. The use of ARS is extended to float uranium quantitatively and selectively from aqueous media at pH {approx} 4 by using oleic acid as a surfactant. The different parameters affecting the flotation process have also been investigated. Uranium(VI) has been effectively separated from natural water samples and certified uranium ores using both procedures.« less

Bacterial leaching of waste uranium materials.

PubMed

Barbic, F F; Bracilović, D M; Krajincanić, B V; Lucić, J L

1976-01-01

The effect of ferrobacteria and thiobacteria on the leaching of waste uranium materials from which 70-80% of uranium was previously leached by classical chemical hydrometallurgical procedure has been investigated. The bacteria used are found in the ore and the mine water of Zletovska River locality, Yugoslavia. Parameters of biological leaching were examined in the laboratory. Leaching conditions were changed with the aim of increasing the amount of uranium leached. The effect of pyrite added to the waste materials before the beginning of leaching has also been examined. Uranium leaching is directly proportional to the composition and number of ferrobacteria and thiobacteria, and increased by almost twice the value obtained from the same starting materials without using bacteria. Increased sulphuric acid concentrations stimulate considerably the rate of leaching. Uranium leaching is increased up to 20% while sulphuric acid consumption is simultaneously decreased by the addition of pyrite. Uranium concentrations in starting waste materials used for leaching were extremely low (0.0278 and 0.372% U) but about 60% recovery of uranium was obtained, with relatively low consumption of sulphuric acid.

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.

New Fiber Materials with Sorption Capacity at 5.0 g-U/kg Adsorbent under Marine Testing Conditions

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

Saito, Tomonori; Brown, S.; Das, Sadananda

The Fuel Resources program of the Fuel Cycle Research and Development program of the Office of Nuclear Energy (NE) has focused on assuring that nuclear fuel resources are available in the United States for a long term. An immense source of uranium is seawater, which contains an estimated amount of 4.5 billion tonnes of dissolved uranium. Extraction of the uranium resource in seawater 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 uraniummore » recovery from seawater. The goal is to develop advanced adsorbents to make the seawater uranium recovery technology a cost competitive, viable technology. Under this program, Oak Ridge National Laboratory (ORNL) has developed several novel adsorbents, which enhanced the uranium capacity 4-5 times from the state-of-the art Japanese adsorbents. Uranium exists uniformly at a concentration of ~3.3 ppb in seawater. Because of the vast volume of the oceans, the total estimated amount of uranium in seawater is approximately 1000 times larger than its amount in terrestrial resources. However, due to the low concentration, a significant challenge remains for making the extraction of uranium from seawater a commercially viable alternative technology. The biggest challenge for this technology to overcome to efficiently reduce the extraction cost is to develop adsorbents with increased uranium adsorption capacity. Two major approaches were investigated for synthesizing novel adsorbents with enhanced uranium adsorption capacity. One method utilized conventional radiation induced graft polymerization (RIGP) to synthesize adsorbents on high-surface area trunk fibers and the other method utilized a chemical grafting technique, atom-transfer radical polymerization (ATRP). Both approaches have shown promising uranium extraction capacities: RIGP adsorbent achieved 5.00 ± 0.15 g U/kg-ads., while ATRP adsorbent achieved 6.56 ± 0.33 g U/kg-ads., after 56 days of seawater exposure. These achieved values are the highest adsorption capacities ever reported for uranium extraction from seawater. The study successfully demonstrated new fiber materials with sorption capacity at 5.0 g-U/kg adsorbent under marine testing conditions. Further optimization, investigation of other new materials as well as deepening our understanding will develop adsorbents that have even higher uranium adsorption capacity, increased selectivity, and faster kinetics.« less

Uranium Adsorbent Fibers Prepared by Atom-Transfer Radical Polymerization (ATRP) from Poly(vinyl chloride)- co -chlorinated Poly(vinyl chloride) (PVC- co -CPVC) Fiber

DOE PAGES

Brown, Suree; Yue, Yanfeng; Kuo, Li-Jung; ...

2016-03-11

The need to secure future supplies of energy attracts researchers in several countries to a vast resource of nuclear energy fuel: uranium in seawater (estimated at 4.5 billion tons in seawater). In this study, we developed effective adsorbent fibers for the recovery of uranium from seawater via atom-transfer radical polymerization (ATRP) from a poly-(vinyl chloride)-co-chlorinated poly(vinyl chloride) (PVC-co-CPVC) fiber. ATRP was employed in the surface graft polymerization of acrylonitrile (AN) and tert-butyl acrylate (tBA), precursors for uranium-interacting functional groups, from PVC-co-CPVC fiber. The [tBA]/[AN] was systematically varied to identify the optimal ratio between hydrophilic groups (from tBA) and uranyl-binding ligandsmore » (from AN). The best performing adsorbent fiber, the one with the optimal [tBA]/[AN] ratio and a high degree of grafting (1390%), demonstrated uranium adsorption capacities that are significantly greater than those of the Japan Atomic Energy Agency (JAEA) reference fiber in natural seawater tests (2.42 3.24 g/kg in 42 days of seawater exposure and 5.22 g/kg in 49 days of seawater exposure, versus 1.66 g/kg in 42 days of seawater exposure and 1.71 g/kg in 49 days of seawater exposure for JAEA). Lastly, adsorption of other metal ions from seawater and their corresponding kinetics were also studied. The grafting of alternative monomers for the recovery of uranium from seawater is now under development by this versatile technique of ATRP.« less

Uranium Adsorbent Fibers Prepared by Atom-Transfer Radical Polymerization (ATRP) from Poly(vinyl chloride)- co -chlorinated Poly(vinyl chloride) (PVC- co -CPVC) Fiber

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

Brown, Suree; Yue, Yanfeng; Kuo, Li-Jung

The need to secure future supplies of energy attracts researchers in several countries to a vast resource of nuclear energy fuel: uranium in seawater (estimated at 4.5 billion tons in seawater). In this study, we developed effective adsorbent fibers for the recovery of uranium from seawater via atom-transfer radical polymerization (ATRP) from a poly-(vinyl chloride)-co-chlorinated poly(vinyl chloride) (PVC-co-CPVC) fiber. ATRP was employed in the surface graft polymerization of acrylonitrile (AN) and tert-butyl acrylate (tBA), precursors for uranium-interacting functional groups, from PVC-co-CPVC fiber. The [tBA]/[AN] was systematically varied to identify the optimal ratio between hydrophilic groups (from tBA) and uranyl-binding ligandsmore » (from AN). The best performing adsorbent fiber, the one with the optimal [tBA]/[AN] ratio and a high degree of grafting (1390%), demonstrated uranium adsorption capacities that are significantly greater than those of the Japan Atomic Energy Agency (JAEA) reference fiber in natural seawater tests (2.42 3.24 g/kg in 42 days of seawater exposure and 5.22 g/kg in 49 days of seawater exposure, versus 1.66 g/kg in 42 days of seawater exposure and 1.71 g/kg in 49 days of seawater exposure for JAEA). Lastly, adsorption of other metal ions from seawater and their corresponding kinetics were also studied. The grafting of alternative monomers for the recovery of uranium from seawater is now under development by this versatile technique of ATRP.« less

Uranium Adsorbent Fibers Prepared by Atom-Transfer Radical Polymerization (ATRP) from Poly(vinyl chloride)- co -chlorinated Poly(vinyl chloride) (PVC- co -CPVC) Fiber

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

Brown, Suree; Yue, Yanfeng; Kuo, Li-Jung

The need to secure future supplies of energy attracts researchers in several countries to a vast resource of nuclear energy fuel: uranium in seawater (estimated at 4.5 billion tons in seawater). In this study, we developed effective adsorbent fibers for the recovery of uranium from seawater via atom-transfer radical polymerization (ATRP) from a poly- (vinyl chloride)-co-chlorinated poly(vinyl chloride) (PVC-co-CPVC) fiber. ATRP was employed in the surface graft polymerization of acrylonitrile (AN) and tert-butyl acrylate (tBA), precursors for uranium-interacting functional groups, from PVC-co-CPVC fiber. The [tBA]/[AN] was systematically varied to identify the optimal ratio between hydrophilic groups (from tBA) and uranyl-bindingmore » ligands (from AN). The best performing adsorbent fiber, the one with the optimal [tBA]/[AN] ratio and a high degree of grafting (1390%), demonstrated uranium adsorption capacities that are significantly greater than those of the Japan Atomic Energy Agency (JAEA) reference fiber in natural seawater tests (2.42-3.24 g/kg in 42 days of seawater exposure and 5.22 g/kg in 49 days of seawater exposure, versus 1.66 g/kg in 42 days of seawater exposure and 1.71 g/kg in 49 days of seawater exposure for JAEA). Adsorption of other metal ions from seawater and their corresponding kinetics were also studied. The grafting of alternative monomers for the recovery of uranium from seawater is now under development by this versatile technique of ATRP.« less

Recovery of Uranium from Seawater: Preparation and Development of Polymer-Supported Extractants

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

Spiro, Alexandratos

2013-12-01

A new series of polymer-supported extractants is proposed for the removal and recovery of uranium from seawater. The objective is to produce polymers with improved stability, loading capacity, and sorption kinetics compared to what is found with amidoximes. The target ligands are diphosphonates and aminomethyldiphosphonates. Small molecule analogues, especially of aminomethyldiphos-phonates, have exceptionally high stability constants for the uranyl ion. The polymeric diphosphonates will have high affinities due to their ability to form six-membered rings with the uranyl ion while the aminomethyldiphosphonates may have yet higher affinities due to possible tridentate coordination and their greater acidity. A representative set ofmore » the polymers to be prepared are indicated.« 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.

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.

Recovery of tritium from tritiated molecules

DOEpatents

Swansiger, William A.

1987-01-01

A method of recovering tritium from tritiated compounds comprises the steps of heating tritiated water and other co-injected tritiated compounds in a preheater to temperatures of about 600.degree. C. The mixture is injected into a reactor charged with a mixture of uranium and uranium dioxide. The injected mixture undergoes highly exothermic reactions with the uranium causing reaction temperatures to occur in excess of the melting point of uranium, and complete decomposition of the tritiated compounds to remove tritium therefrom. The uranium dioxide functions as an insulating material and heat sink preventing the reactor side walls from attaining reaction temperatures to thereby minimize tritium permeation rates. The uranium dioxide also functions as a diluent to allow for volumetric expansion of the uranium as it is converted to uranium dioxide. The reactor vessel is preferably stainless steel of sufficient mass so as to function as a heat sink preventing the reactor side walls from approaching high temperatures. A disposable copper liner extends between the reaction chamber and stainless steel outer vessel to prevent alloying of the uranium with the outer vessel. Apparatus used to carry out the method of the invention is also disclosed.

SUMMARY TECHNICAL REPORT ON FEED MATERIALS FOR THE PERIOD APRIL 1, 1959 TO JUNE 30, 1959

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

Simmons, J.W. ed.

1959-07-20

Anaconda Acld, Kermac, Moab, Rifle, and Texas Zinc uranium concentrates were evaluated (the laboratory portlon of feed material evaluation). Laboratory equilibrium tests and Pilot Plant 2-inch-column extraction tests demonstrated effective distribution of uranium into a TBPkerosene solvent from aqueous phases containing as little as 0.5N HNO/sub 3/ and varying amounts of added metal nitrates (NaNO/sub 3/). The concentration of assoclated nitric acid in dilute aqueous nitric acld solutions was determined after values were obtained for the equillbrium constant for the reaction of tri-n-butyl phosphate with associated nitric acid and for the equilibrium distribution constant for the partition of associated nitricmore » acld into tri-n-butyl phosphate. Optimum partition of uranium into tri-n-butyl phosphate was realized in the laboratory by using an aqueous uranyl nitrate solution containing sufficient hydrogen ions to promote extraction and a low concentration of associated nitric acid. An Ohmart system for controlling the uranium profile in the A'' extractlon column was installed on Refinery pulse columns. Use of this system improved control but did not stop all column upsets. The effect of 13 to l89 ppm sodium contaminatlon upon hydrofluorination conversion of teraperature at the site of the reaction. Uranyl sulfate was shown to undergo an enantiotroplc transitlon at 755 deg C and to decompose to U/sub 3/O/sub 8/ in an atmosphere of oxygen sulfur dioxide, which gases are evolved during decoraposition. Decontamination of sodium, calcium, nickel, magnesium, gadolinium, and dysprosium was achieved in a laboratory investigatlon of the ADU process. UO/sub 2/ produced by reductions programmed from 700 to ll00 deg F was hydrofluorinated at programmed temperatures of 550 to 1100 deg F and isothermally at ll00 deg F. Good conversion was obtained for material whose source was ADU calcined at 1200 deg F. Uranium derbles were classified by the present method of derby grading and were then examined for slag coverage, slag volume, and slag weight. There was a high degree of overlap of these parameters for adjacent grades. A hydraulic separator for separatlng uranlum from magnesium and magnesium fluorlde was fabrlcated. Excellent separatlon was obtained for +l6 mesh material. A hydrochloric acid dissolution- UF/sub 4/ precipitation process for routing scrap materials to the reductlon-to- metal step was examined. The purification obtained was noted, and process conditions were varied to determine their effect upon UF/sub 4/ density, UF/sub 4/ purity and precipitation time. Three types of uranium scrap were subjected to the HCl dissolution-aqueous precipitation Winlo process to determine the purification achieved. Green salt made from dolomitlc bomb liner residues was found to be grossly contaminated. Acceptable green salt was raade from pickle liquor treated with formaldehyde and from pickle liquor plus black oxide. Nominal 80% yields were obtained in the recovery of magnesium metal by reaction of calcium carblde with magnesium fluoride slag and in the recovery of HF by the reactlon of sulfuric acid wlth magnesium fluoride slag. A sample holder for use in quantitative preferred orientation studies was fabricated. The holder, designed to fit a North American Philips Gonionweter, will accommodate specimens up to l 13/16 inches in diameter and incorporates a precision ball bearing. A satisfactory technique was developed for the analysis of uranium metal for traces of fluoride. A direct flame photometric method is glven for the determination of magnesium in uranium ore concentrates. No chemical separation step is required, except for high-iron-content ores. (auth)« less

ADSORPTION METHOD FOR SEPARATING THORIUM VALUES FROM URANIUM VALUES

DOEpatents

Boyd, G.E.; Russell, E.R.; Schubert, J.

1959-08-01

An improved ion exchange method is described for recovery of uranium and thorium values as separate functions from an aqueous acidic solution containing less than 10/sup -3/ M thorium ions and between 0.1 and 1 M uranyl ions. The solution is passed through a bed of cation exchange resin in the acid form to adsorb all the thorium ions and a portion of the uranyl ions. The uranium is eluted by means of aqueous 0.1 to 0.4 M sulfuric acid. The thorium may then be stripped from the resin by elution with aqueous 0.5 M oxalic acid.

ADSORPTION METHOD FOR SEPARATING THORIUM VALUES FROM URANIUM VALUES

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

Boyd, G.E.; Russell, E.R.; Schubert, J.

An improved ion exchange method is described for recovery of uranium and thorium values as separate functions from an aqueous acidic solution containing less than 10/sup -3/ M thorium ions and between 0.1 and 1 M uranyl ions. The solution is passed through a bed of cation exchange resin in the acid form to adsorb all the thorium ions and a portion of the uranyl ions. The uranium is eluted by means of aqueous 0.1 to 0.4 M sulfuric acid. The thorium may then be stripped from the resin by elution with aqueous 0.5 M oxalic acid.

Experiments and Modeling of Uranium Adsorption in the Presence of Other Ions in Simulated Seawater

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

Ladshaw, Austin; Das, Sadananda; Liao, Wei-Po

2015-11-19

Seawater contains uranium at an average concentration of 3.3 ppb, as well as a variety of other ions at either overwhelmingly higher or similar concentrations, which complicate the recovery of uranium. This report describes an investigation of the effects of various factors such as uranium speciation and presence of salts including sodium, calcium, magnesium, and bicarbonate, as well as trace elements such as vanadium on uranium adsorption kinetics in laboratory experiments. Adsorption models are also developed to describe the experimental data of uranium extraction from seawater. Results show that the presence of calcium and magnesium significantly slows down the uraniummore » adsorption kinetics. Vanadium can replace uranium from amidoxime-based adsorbent in the presence of sodium in the solution. Results also show that bicarbonate in the solution strongly competes with amidoxime for binding uranium, and thus slows down the uranium adsorption kinetics. Developed on the basis of the experimental findings, the model is capable of describing the effects of pH, ionic strength, temperature, and concentration of various species. The results of this work are useful in the understanding of the important factors that control the adsorbent capacity and kinetics of uranium uptake by amidoxime-based adsorbents.« less

Uranium quantification in semen by inductively coupled plasma mass spectrometry

USGS Publications Warehouse

Todorov, Todor I.; Ejnik, John W.; Guandalini, Gustavo S.; Xu, Hanna; Hoover, Dennis; Anderson, Larry W.; Squibb, Katherine; McDiarmid, Melissa A.; Centeno, Jose A.

2013-01-01

In this study we report uranium analysis for human semen samples. Uranium quantification was performed by inductively coupled plasma mass spectrometry. No additives, such as chymotrypsin or bovine serum albumin, were used for semen liquefaction, as they showed significant uranium content. For method validation we spiked 2 g aliquots of pooled control semen at three different levels of uranium: low at 5 pg/g, medium at 50 pg/g, and high at 1000 pg/g. The detection limit was determined to be 0.8 pg/g uranium in human semen. The data reproduced within 1.4–7% RSD and spike recoveries were 97–100%. The uranium level of the unspiked, pooled control semen was 2.9 pg/g of semen (n = 10). In addition six semen samples from a cohort of Veterans exposed to depleted uranium (DU) in the 1991 Gulf War were analyzed with no knowledge of their exposure history. Uranium levels in the Veterans’ semen samples ranged from undetectable (<0.8 pg/g) to 3350 pg/g. This wide concentration range for uranium in semen is consistent with known differences in current DU body burdens in these individuals, some of whom have retained embedded DU fragments.

Method of uranium reclamation from aqueous systems by reactive ion exchange. [US DOE patent application; anion exchange resin of copolymerized divinyl-benzene and styrene having quarternary ammonium groups and bicarbonate ligands

DOEpatents

Maya, L.

1981-11-05

A reactive ion exchange method for separation and recovery of values of uranium, neptunium, plutonium, or americium from substantially neutral aqueous systems of said metals comprises contacting said system with an effective amount of a basic anion exchange resin of copolymerized divinyl-benzene and styrene having quarternary ammonium groups and bicarbonate ligands to achieve nearly 100% sorption of said actinyl ion onto said resin and an aqueous system practically free of said actinyl ions. The method is operational over an extensive range of concentrations from about 10/sup -6/ M to 1.0 M actinyl ion and a pH range of about 4 to 7. The method has particulr application to treatment of waste streams from Purex-type nuclear fuel reprocessing facilities and hydrometallurgical processes involving U, Np, P, or Am.

Process for electrolytically preparing uranium metal

DOEpatents

Haas, Paul A.

1989-01-01

A process for making uranium metal from uranium oxide by first fluorinating uranium oxide to form uranium tetrafluoride and next electrolytically reducing the uranium tetrafluoride with a carbon anode to form uranium metal and CF.sub.4. The CF.sub.4 is reused in the fluorination reaction rather than being disposed of as a hazardous waste.

Process for electrolytically preparing uranium metal

DOEpatents

Haas, Paul A.

1989-08-01

A process for making uranium metal from uranium oxide by first fluorinating uranium oxide to form uranium tetrafluoride and next electrolytically reducing the uranium tetrafluoride with a carbon anode to form uranium metal and CF.sub.4. The CF.sub.4 is reused in the fluorination reaction rather than being disposed of as a hazardous waste.

Macroporous monoliths for trace metal extraction from seawater

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

Yue, Yanfeng; Mayes, Richard; Gill, Gary A.

2015-05-29

The viability of seawater-based uranium recovery depends on the uranium adsorption rate and capacity, since the concentration of uranium in the oceans is relatively low (3.3 μgL⁻¹). An important consideration for a fast adsorption is to maximize the adsorption properties of adsorbents such as surface areas and pore structures, which can greatly improve the kinetics of uranium extraction and the adsorption capacity simultaneously. Following this consideration, macroporous monolith adsorbents were prepared from the copolymerization of acrylonitrile (AN) and N, N’-methylenebis(acrylamide) (MBAAm) based on a cryogel method using both hydrophobic and hydrophilic monomers. The monolithic sorbents were tested with simulated seawatermore » containing a high uranyl concentration (–6 ppm) and the uranium adsorption results showed that the adsorption capacities are strongly influenced by the ratio of monomer to the crosslinker, i.e., the density of the amidoxime groups. The preliminary seawater testing indicates the high salinity content of seawater does not hinder the adsorption of uranium.« less

Alternative alkaline conditioning of amidoxime based adsorbent for uranium extraction from seawater

DOE PAGES

Das, Sadananda; Liao, Wei -Po; Byers, Maggie Flicker; ...

2015-10-18

Alkaline conditioning of the amidoxime based adsorbents is a significant step in the preparation of the adsorbent for uranium uptake from seawater. The effects of various alkaline conditioning parameters such as the type of alkaline reagent, reaction temperature, and reaction time were investigated with respect to uranium adsorption capacity from simulated seawater (spiked with 8 ppm uranium) and natural seawater (from Sequim Bay, WA). An adsorbent (AF1) was prepared at the Oak Ridge National Laboratory by radiation-induced graft polymerization (RIGP) with acrylonitrile and itaconic acid onto high-surface-area polyethylene fibers. For the AF1 adsorbent, sodium hydroxide emerged as a better reagentmore » for alkaline conditioning over potassium hydroxide, which has typically been used in previous studies, because of higher uranium uptake capacity and lower cost over the other candidate alkaline reagents investigated in this study. Furthermore, the use of sodium hydroxide in place of potassium hydroxide is shown to result in a 21–30% decrease in the cost of uranium recovery.« less

Macroporous monoliths for trace metal extraction from seawater

DOE PAGES

Yue, Yanfeng; Mayes, Richard T.; Gill, Gary; ...

2015-05-29

The viability of seawater-based uranium recovery depends on the uranium adsorption rate and capacity, since the concentration of uranium in the oceans is relatively low (3.3 gL -1). An important consideration for a fast adsorption is to maximize the adsorption properties of adsorbents such as surface areas and pore structures, which can greatly improve the kinetics of uranium extraction and the adsorption capacity simultaneously. Following this consideration, macroporous monolith adsorbents were prepared from the copolymerization of acrylonitrile (AN) and N,N -methylenebis(acrylamide) (MBAAm) based on a cryogel method using both hydrophobic and hydrophilic monomers. The monolithic sorbents were tested with simulatedmore » seawater containing a high uranyl concentration (–6 ppm) and the uranium adsorption results showed that the adsorption capacities are strongly influenced by the ratio of monomer to the crosslinker, i.e., the density of the amidoxime groups. Furthermore, the preliminary seawater testing indicates the high salinity content of seawater does not hinder the adsorption of uranium.« less

Alternative Alkaline Conditioning of Amidoxime Based Adsorbent for Uranium Extraction from Seawater

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

Das, S.; Liao, W. -P.; Flicker Byers, M.

2016-04-20

Alkaline conditioning of the amidoxime based adsorbents is a significant step in the preparation of the adsorbent for uranium uptake from seawater. The effects of various alkaline conditioning parameters such as the type of alkaline reagent, reaction temperature, and reaction time were investigated with respect to uranium adsorption capacity from simulated seawater (spiked with 8 ppm uranium) and natural seawater (from Sequim Bay, WA). An adsorbent (AF1) was prepared at the Oak Ridge National Laboratory by radiation-induced graft polymerization (RIGP) with acrylonitrile and itaconic acid onto high-surface-area polyethylene fibers. For the AF1 adsorbent, sodium hydroxide emerged as a better reagentmore » for alkaline conditioning over potassium hydroxide, which has typically been used in previous studies, because of higher uranium uptake capacity and lower cost over the other candidate alkaline reagents investigated in this study. Use of sodium hydroxide in place of potassium hydroxide is shown to result in a 21-30% decrease in the cost of uranium recovery.« less

Alternative alkaline conditioning of amidoxime based adsorbent for uranium extraction from seawater

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

Das, Sadananda; Liao, Wei -Po; Byers, Maggie Flicker

Alkaline conditioning of the amidoxime based adsorbents is a significant step in the preparation of the adsorbent for uranium uptake from seawater. The effects of various alkaline conditioning parameters such as the type of alkaline reagent, reaction temperature, and reaction time were investigated with respect to uranium adsorption capacity from simulated seawater (spiked with 8 ppm uranium) and natural seawater (from Sequim Bay, WA). An adsorbent (AF1) was prepared at the Oak Ridge National Laboratory by radiation-induced graft polymerization (RIGP) with acrylonitrile and itaconic acid onto high-surface-area polyethylene fibers. For the AF1 adsorbent, sodium hydroxide emerged as a better reagentmore » for alkaline conditioning over potassium hydroxide, which has typically been used in previous studies, because of higher uranium uptake capacity and lower cost over the other candidate alkaline reagents investigated in this study. Furthermore, the use of sodium hydroxide in place of potassium hydroxide is shown to result in a 21–30% decrease in the cost of uranium recovery.« less

75 FR 6066 - Extension of Public Comment Period on the Draft Environmental Impact Statement for the Nichols...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-02-05

... Environmental Impact Statement for the Nichols Ranch In-Situ Recovery Project in Campbell and Johnson Counties, WY; Supplement to the Generic Environmental Impact Statement for In-Situ Leach Uranium Milling... Environmental Impact Statement (Draft SEIS) for the Nichols Ranch In- Situ Recovery (ISR) Project closes on...

76 FR 71082 - Strata Energy, Inc., Ross Uranium Recovery Project; New Source Material License Application...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-11-16

...-situ recovery (ISR), also known as in-situ leach, facilities and restoration of the aquifer from which... Supplemental Environmental Impact Statement (SEIS) to the Generic Environmental Impact Statement for In-Situ... located in Crook County, Wyoming, 32 miles northeast of Gillette, Wyoming and 30 miles northwest of...

76 FR 37111 - Environmental Impacts Statements; Notice of Availability

Federal Register 2010, 2011, 2012, 2013, 2014

2011-06-24

...-5410 EIS No. 20110197, Final EIS, NRC, AK, Lost Creek In-Situ Uranium Recovery (ISR) Project, Proposal to Construct, Operate, Conduit Aquifer Restoration, and Decommission an In-Situ Recovery (ISR.../13/2011 Through 06/17/2011 Pursuant to 40 CFR 1506.9. Notice In accordance with Section 309(a) of the...

75 FR 52736 - Environmental Impacts Statements; Notice of Availability

Federal Register 2010, 2011, 2012, 2013, 2014

2010-08-27

... . EIS No. 20100337, Final EIS, NRC, WY, Moore Ranch In-Situ Uranium Recovery (ISR) Project, Proposal to Construct, Operate, Conduct Aquifer Restoration, and Decommission an In-Situ Recovery (ISR) Facility, NUREG... Through 08/20/2010. Pursuant to 40 CFR 1506.9. Notice In accordance with Section 309(a) of the Clean Air...

Organic Separation Test Results

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

Russell, Renee L.; Rinehart, Donald E.; Peterson, Reid A.

2014-09-22

Separable organics have been defined as “those organic compounds of very limited solubility in the bulk waste and that can form a separate liquid phase or layer” (Smalley and Nguyen 2013), and result from three main solvent extraction processes: U Plant Uranium Recovery Process, B Plant Waste Fractionation Process, and Plutonium Uranium Extraction (PUREX) Process. The primary organic solvents associated with tank solids are TBP, D2EHPA, and NPH. There is concern that, while this organic material is bound to the sludge particles as it is stored in the tanks, waste feed delivery activities, specifically transfer pump and mixer pump operations,more » could cause the organics to form a separated layer in the tank farms feed tank. Therefore, Washington River Protection Solutions (WRPS) is experimentally evaluating the potential of organic solvents separating from the tank solids (sludge) during waste feed delivery activities, specifically the waste mixing and transfer processes. Given the Hanford Tank Waste Treatment and Immobilization Plant (WTP) waste acceptance criteria per the Waste Feed Acceptance Criteria document (24590-WTP-RPT-MGT-11-014) that there is to be “no visible layer” of separable organics in the waste feed, this would result in the batch being unacceptable to transfer to WTP. This study is of particular importance to WRPS because of these WTP requirements.« less

IMPROVED PROCESS OF PLUTONIUM CARRIER PRECIPITATION

DOEpatents

Faris, B.F.

1959-06-30

This patent relates to an improvement in the bismuth phosphate process for separating and recovering plutonium from neutron irradiated uranium, resulting in improved decontamination even without the use of scavenging precipitates in the by-product precipitation step and subsequently more complete recovery of the plutonium in the product precipitation step. This improvement is achieved by addition of fluomolybdic acid, or a water soluble fluomolybdate, such as the ammonium, sodium, or potassium salt thereof, to the aqueous nitric acid solution containing tetravalent plutonium ions and contaminating fission products, so as to establish a fluomolybdate ion concentration of about 0.05 M. The solution is then treated to form the bismuth phosphate plutonium carrying precipitate.

Process for electroslag refining of uranium and uranium alloys

DOEpatents

Lewis, P.S. Jr.; Agee, W.A.; Bullock, J.S. IV; Condon, J.B.

1975-07-22

A process is described for electroslag refining of uranium and uranium alloys wherein molten uranium and uranium alloys are melted in a molten layer of a fluoride slag containing up to about 8 weight percent calcium metal. The calcium metal reduces oxides in the uranium and uranium alloys to provide them with an oxygen content of less than 100 parts per million. (auth)

Influence of Current Velocity on Uranium Adsorption from Seawater Using an Amidoxime-based Polymer Fiber Adsorbent

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

Ladshaw, Austin; Kuo, Li-Jung; Strivens, Jonathan

2017-02-08

Passive adsorption using amidoxime-based polymeric adsorbents is being developed for uranium recovery from seawater. The local oceanic current velocity where the adsorbent is deployed is a key variable in determining locations that will maximize uranium adsorption rates. Two independent experimental approaches using flow-through columns and recirculating flumes were used to assess the influence of linear velocity on uranium uptake kinetics by the adsorbent. Little to no difference was observed in the uranium adsorption rate vs. linear velocity for seawater exposure in flow-through columns. In contrast, adsorption results from seawater exposure in a recirculating flume showed a nearly linear trend withmore » current velocity. The difference in adsorbent performance between columns and flume can be attributed to (i) flow resistance provided by the adsorbent braid in the flume and (ii) enhancement in braid movement (fluttering) with increasing linear velocity.« less

Influence of Current Velocity on Uranium Adsorption from Seawater Using an Amidoxime-Based Polymer Fiber Adsorbent

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

Ladshaw, Austin; Kuo, Li-Jung; Strivens, Jonathan

2017-02-17

Passive adsorption using amidoxime-based polymeric adsorbents is being developed for uranium recovery from seawater. The local oceanic current velocity where the adsorbent is deployed is a key variable in determining locations that will maximize uranium adsorption rates. Two independent experimental approaches using flow-through columns and recirculating flumes were used to assess the influence of linear velocity on uranium uptake kinetics by the adsorbent. Little to no difference was observed in the uranium adsorption rate vs. linear velocity for seawater exposure in flow-through columns. In contrast, adsorption results from seawater exposure in a recirculating flume showed a nearly linear trend withmore » current velocity. The difference in adsorbent performance between columns and flume can be attributed to (i) flow resistance provided by the adsorbent braid in the flume and (ii) enhancement in braid movement (fluttering) with increasing linear velocity.« less

Biogeochemical aspects of uranium mineralization, mining, milling, and remediation

USGS Publications Warehouse

Campbell, Kate M.; Gallegos, Tanya J.; Landa, Edward R.

2015-01-01

Natural uranium (U) occurs as a mixture of three radioactive isotopes: 238U, 235U, and 234U. Only 235U is fissionable and makes up about 0.7% of natural U, while 238U is overwhelmingly the most abundant at greater than 99% of the total mass of U. Prior to the 1940s, U was predominantly used as a coloring agent, and U-bearing ores were mined mainly for their radium (Ra) and/or vanadium (V) content; the bulk of the U was discarded with the tailings (Finch et al., 1972). Once nuclear fission was discovered, the economic importance of U increased greatly. The mining and milling of U-bearing ores is the first step in the nuclear fuel cycle, and the contact of residual waste with natural water is a potential source of contamination of U and associated elements to the environment. Uranium is mined by three basic methods: surface (open pit), underground, and solution mining (in situ leaching or in situ recovery), depending on the deposit grade, size, location, geology and economic considerations (Abdelouas, 2006). Solid wastes at U mill tailings (UMT) sites can include both standard tailings (i.e., leached ore rock residues) and solids generated on site by waste treatment processes. The latter can include sludge or “mud” from neutralization of acidic mine/mill effluents, containing Fe and a range of coprecipitated constituents, or barium sulfate precipitates that selectively remove Ra (e.g., Carvalho et al., 2007). In this chapter, we review the hydrometallurgical processes by which U is extracted from ore, the biogeochemical processes that can affect the fate and transport of U and associated elements in the environment, and possible remediation strategies for site closure and aquifer restoration.This paper represents the fourth in a series of review papers from the U.S. Geological Survey (USGS) on geochemical aspects of UMT management that span more than three decades. The first paper (Landa, 1980) in this series is a primer on the nature of tailings and radionuclide mobilization from them. The second paper (Landa, 1999) includes coverage of research carried out under the U.S. Department of Energy’s Uranium Mill Tailings Remedial Action Program (UMTRA). The third paper (Landa, 2004) reflects the increased focus of researchers on biotic effects in UMT environs. This paper expands the focus to U mining, milling, and remedial actions, and includes extensive coverage of the increasingly important alkaline in situ recovery and groundwater restoration.

Experimental design and optimization of leaching process for recovery of valuable chemical elements (U, La, V, Mo, Yb and Th) from low-grade uranium ore.

PubMed

Zakrzewska-Koltuniewicz, Grażyna; Herdzik-Koniecko, Irena; Cojocaru, Corneliu; Chajduk, Ewelina

2014-06-30

The paper deals with experimental design and optimization of leaching process of uranium and associated metals from low-grade, Polish ores. The chemical elements of interest for extraction from the ore were U, La, V, Mo, Yb and Th. Sulphuric acid has been used as leaching reagent. Based on the design of experiments the second-order regression models have been constructed to approximate the leaching efficiency of elements. The graphical illustrations using 3-D surface plots have been employed in order to identify the main, quadratic and interaction effects of the factors. The multi-objective optimization method based on desirability approach has been applied in this study. The optimum condition have been determined as P=5 bar, T=120 °C and t=90 min. Under these optimal conditions, the overall extraction performance is 81.43% (for U), 64.24% (for La), 98.38% (for V), 43.69% (for Yb) and 76.89% (for Mo) and 97.00% (for Th). Copyright © 2014 Elsevier B.V. All rights reserved.

Dry halide method for separating the components of spent nuclear fuels

DOEpatents

Christian, Jerry Dale; Thomas, Thomas Russell; Kessinger, Glen F.

1998-01-01

The invention is a nonaqueous, single method for processing multiple spent nuclear fuel types by separating the fission- and transuranic products from the nonradioactive and fissile uranium product. The invention has four major operations: exposing the spent fuels to chlorine gas at temperatures preferably greater than 1200.degree. C. to form volatile metal chlorides; removal of the fission product chlorides, transuranic product chlorides, and any nickel chloride and chromium chloride in a molten salt scrubber at approximately 400.degree. C.; fractional condensation of the remaining volatile chlorides at temperatures ranging from 164.degree. C. to 2.degree. C.; and regeneration and recovery of the transferred spent molten salt by vacuum distillation. The residual fission products, transuranic products, and nickel- and chromium chlorides are converted to fluorides or oxides for vitrification. The method offers the significant advantages of a single, compact process that is applicable to most of the diverse nuclear fuels, minimizes secondary wastes, segregates fissile uranium from the high level wastes to resolve potential criticality concerns, segregates nonradioactive wastes from the high level wastes for volume reduction, and produces a common waste form glass or glass-ceramic.

Dry halide method for separating the components of spent nuclear fuels

DOEpatents

Christian, J.D.; Thomas, T.R.; Kessinger, G.F.

1998-06-30

The invention is a nonaqueous, single method for processing multiple spent nuclear fuel types by separating the fission and transuranic products from the nonradioactive and fissile uranium product. The invention has four major operations: exposing the spent fuels to chlorine gas at temperatures preferably greater than 1200 C to form volatile metal chlorides; removal of the fission product chlorides, transuranic product chlorides, and any nickel chloride and chromium chloride in a molten salt scrubber at approximately 400 C; fractional condensation of the remaining volatile chlorides at temperatures ranging from 164 to 2 C; and regeneration and recovery of the transferred spent molten salt by vacuum distillation. The residual fission products, transuranic products, and nickel- and chromium chlorides are converted to fluorides or oxides for vitrification. The method offers the significant advantages of a single, compact process that is applicable to most of the diverse nuclear fuels, minimizes secondary wastes, segregates fissile uranium from the high level wastes to resolve potential criticality concerns, segregates nonradioactive wastes from the high level wastes for volume reduction, and produces a common waste form glass or glass-ceramic. 3 figs.

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...

Process for the removal of radium from acidic solutions containing same

DOEpatents

Scheitlin, F.M.

The invention is a process for the removal of radium from acidic aqueous solutions. In one aspect, the invention is a process for removing radium from an inorganic-acid solution. The process comprises contacting the solution with coal fly ash to effect adsorption of the radium on the ash. The radium-containing ash then is separated from the solution. The process is simple, comparatively inexpensive, and efficient. High radium-distribution coefficients are obtained even at room temperature. Coal fly ash is an inexpensive, acid-resistant, high-surface-area material which is available in large quantities throughout the United States. The invention is applicable, for example, to the recovery of /sup 226/Ra from nitric acid solutions which have been used to leach radium from uranium-mill tailings.

A Uranium Bioremediation Reactive Transport Benchmark

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

Yabusaki, Steven B.; Sengor, Sevinc; Fang, Yilin

A reactive transport benchmark problem set has been developed based on in situ uranium bio-immobilization experiments that have been performed at a former uranium mill tailings site in Rifle, Colorado, USA. Acetate-amended groundwater stimulates indigenous microorganisms to catalyze the reduction of U(VI) to a sparingly soluble U(IV) mineral. The interplay between the flow, acetate loading periods and rates, microbially-mediated and geochemical reactions leads to dynamic behavior in metal- and sulfate-reducing bacteria, pH, alkalinity, and reactive mineral surfaces. The benchmark is based on an 8.5 m long one-dimensional model domain with constant saturated flow and uniform porosity. The 159-day simulation introducesmore » acetate and bromide through the upgradient boundary in 14-day and 85-day pulses separated by a 10 day interruption. Acetate loading is tripled during the second pulse, which is followed by a 50 day recovery period. Terminal electron accepting processes for goethite, phyllosilicate Fe(III), U(VI), and sulfate are modeled using Monod-type rate laws. Major ion geochemistry modeled includes mineral reactions, as well as aqueous and surface complexation reactions for UO2++, Fe++, and H+. In addition to the dynamics imparted by the transport of the acetate pulses, U(VI) behavior involves the interplay between bioreduction, which is dependent on acetate availability, and speciation-controlled surface complexation, which is dependent on pH, alkalinity and available surface complexation sites. The general difficulty of this benchmark is the large number of reactions (74), multiple rate law formulations, a multisite uranium surface complexation model, and the strong interdependency and sensitivity of the reaction processes. Results are presented for three simulators: HYDROGEOCHEM, PHT3D, and PHREEQC.« less

75 FR 6065 - Extension of Public Comment Period on the Draft Environmental Impact Statement for the Moore...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-02-05

... Environmental Impact Statement for the Moore Ranch In-Situ Recovery Project in Campbell County, WY; Supplement to the Generic Environmental Impact Statement for In-Situ Leach Uranium Milling Facilities AGENCY... Statement (Draft SEIS) for the Moore Ranch In-Situ Recovery (ISR) Project closes on February 1, 2010. The...

Characterization of cores from an in-situ recovery mined uranium deposit in Wyoming: Implications for post-mining restoration

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

WoldeGabriel, G.; Boukhalfa, H.; Ware, S. D.

In-situ recovery (ISR) of uranium (U) from sandstone-type roll-front deposits is a technology that involves the injection of solutions that consist of ground water fortified with oxygen and carbonate to promote the oxidative dissolution of U, which is pumped to recovery facilities located at the surface that capture the dissolved U and recycle the treated water. The ISR process alters the geochemical conditions in the subsurface creating conditions that are more favorable to the migration of uranium and other metals associated with the uranium deposit. There is a lack of clear understanding of the impact of ISR mining on themore » aquifer and host rocks of the post-mined site and the fate of residual U and other metals within the mined ore zone. We performed detailed petrographic, mineralogical, and geochemical analyses of several samples taken from about 7 m of core of the formerly the ISR-mined Smith Ranch–Highland uranium deposit in Wyoming. We show that previously mined cores contain significant residual uranium (U) present as coatings on pyrite and carbonaceous fragments. Coffinite was identified in three samples. Core samples with higher organic (> 1 wt.%) and clay (> 6–17 wt.%) contents yielded higher 234U/ 238U activity ratios (1.0–1.48) than those with lower organic and clay fractions. The ISR mining was inefficient in mobilizing U from the carbonaceous materials, which retained considerable U concentrations (374–11,534 ppm). This is in contrast with the deeper part of the ore zone, which was highly depleted in U and had very low 234U/ 238U activity ratios. This probably is due to greater contact with the lixiviant (leaching solution) during ISR mining. EXAFS analyses performed on grains with the highest U and Fe concentrations reveal that Fe is present in a reduced form as pyrite and U occurs mostly as U(IV) complexed by organic matter or as U(IV) phases of carbonate complexes. Moreover, U–O distances of ~ 2.05 Å were noted, indicating the potential formation of other poorly defined U(IV/VI) species. We also noted a small contribution from Udouble bond; length as m-dashO at 1.79 Å, which indicates that U is partially oxidized. There is no apparent U–S or U–Fe interaction in any of the U spectra analyzed. However, SEM analysis of thin sections prepared from the same core material reveals surficial U associated with pyrite which is probably a minor fraction of the total U present as thin coatings on the surface of pyrite. Our data show the presence of different structurally variable uranium forms associated with the mined cores. U associated with carbonaceous materials is probably from the original U mobilization that accumulated in the organic matter-rich areas under reducing conditions during shallow burial diagenesis. U associated with pyrite represents a small fraction of the total U and was likely deposited as a result of chemical reduction by pyrite. Our data suggest that areas rich in carbonaceous materials had limited exposure to the lixiviant solution, continue to be reducing, and still hold significant U resources. Because of their limited access to fluid flow, these areas might not contribute significantly to post-mining U release or attenuation. Areas with pyrite that are accessible to fluids seem to be more reactive and could act as reductants and facilitate U reduction and accumulation, limiting its migration.« less

Characterization of cores from an in-situ recovery mined uranium deposit in Wyoming: Implications for post-mining restoration

DOE PAGES

WoldeGabriel, G.; Boukhalfa, H.; Ware, S. D.; ...

2014-10-08

In-situ recovery (ISR) of uranium (U) from sandstone-type roll-front deposits is a technology that involves the injection of solutions that consist of ground water fortified with oxygen and carbonate to promote the oxidative dissolution of U, which is pumped to recovery facilities located at the surface that capture the dissolved U and recycle the treated water. The ISR process alters the geochemical conditions in the subsurface creating conditions that are more favorable to the migration of uranium and other metals associated with the uranium deposit. There is a lack of clear understanding of the impact of ISR mining on themore » aquifer and host rocks of the post-mined site and the fate of residual U and other metals within the mined ore zone. We performed detailed petrographic, mineralogical, and geochemical analyses of several samples taken from about 7 m of core of the formerly the ISR-mined Smith Ranch–Highland uranium deposit in Wyoming. We show that previously mined cores contain significant residual uranium (U) present as coatings on pyrite and carbonaceous fragments. Coffinite was identified in three samples. Core samples with higher organic (> 1 wt.%) and clay (> 6–17 wt.%) contents yielded higher 234U/ 238U activity ratios (1.0–1.48) than those with lower organic and clay fractions. The ISR mining was inefficient in mobilizing U from the carbonaceous materials, which retained considerable U concentrations (374–11,534 ppm). This is in contrast with the deeper part of the ore zone, which was highly depleted in U and had very low 234U/ 238U activity ratios. This probably is due to greater contact with the lixiviant (leaching solution) during ISR mining. EXAFS analyses performed on grains with the highest U and Fe concentrations reveal that Fe is present in a reduced form as pyrite and U occurs mostly as U(IV) complexed by organic matter or as U(IV) phases of carbonate complexes. Moreover, U–O distances of ~ 2.05 Å were noted, indicating the potential formation of other poorly defined U(IV/VI) species. We also noted a small contribution from Udouble bond; length as m-dashO at 1.79 Å, which indicates that U is partially oxidized. There is no apparent U–S or U–Fe interaction in any of the U spectra analyzed. However, SEM analysis of thin sections prepared from the same core material reveals surficial U associated with pyrite which is probably a minor fraction of the total U present as thin coatings on the surface of pyrite. Our data show the presence of different structurally variable uranium forms associated with the mined cores. U associated with carbonaceous materials is probably from the original U mobilization that accumulated in the organic matter-rich areas under reducing conditions during shallow burial diagenesis. U associated with pyrite represents a small fraction of the total U and was likely deposited as a result of chemical reduction by pyrite. Our data suggest that areas rich in carbonaceous materials had limited exposure to the lixiviant solution, continue to be reducing, and still hold significant U resources. Because of their limited access to fluid flow, these areas might not contribute significantly to post-mining U release or attenuation. Areas with pyrite that are accessible to fluids seem to be more reactive and could act as reductants and facilitate U reduction and accumulation, limiting its migration.« less

Investigations into the Reusability of Amidoxime-Based Polymeric Adsorbents for Seawater Uranium Extraction

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

Kuo, Li-Jung; Pan, Horng-Bin; Wai, Chien M.

The ability to re-use amidoxime-based polymeric adsorbents is a critical component in reducing the overall cost of the technology to extract uranium from seawater. This report describes an evaluation of adsorbent reusability in multiple re-use (adsorption/stripping) cycles in real seawater exposures with potassium bicarbonate (KHCO3) elution using several amidoxime-based polymeric adsorbents. The KHCO3 elution technique achieved ~100% recovery of uranium adsorption capacity in the first re-use. Subsequent re-uses showed significant drops in adsorption capacity. After the 4th re-use with the ORNL AI8 adsorbent, the 56-day adsorption capacity dropped to 28% of its original capacity. FTIR spectra revealed that there wasmore » a conversion of the amidoxime ligands to carboxylate groups during extended seawater exposure, becoming more significant with longer the exposure time. Ca and Mg adsorption capacities also increased with each re-use cycle supporting the hypothesis that long term exposure resulted in converting amidoxime to carboxylate, enhancing the adsorption of Ca and Mg. Shorter seawater exposure (adsorption/stripping) cycles (28 vs. 42 days) had higher adsorption capacities after re-use, but the shorter exposure cycle time did not produce an overall better performance in terms of cumulative exposure time. Recovery of uranium capacity in re-uses may also vary across different adsorbent formulations. Through multiple re-use the adsorbent AI8 can harvest 10 g uranium/kg adsorbent in ~140 days, using a 28-day adsorption/stripping cycle, a performance much better than would be achieved with a single use of the adsorbent through very long-term exposure (saturation capacity = 7.4 g U/kg adsorbent). A time dependent seawater exposure model to evaluate the cost associated with reusing amidoxime-based adsorbents in real seawater exposures was developed. The cost to extract uranium from seawater ranged from $610-830/kg U was predicted. Model simulation suggests that a short seawater exposure cycle (< 15 days) is the optimal deployment period for lower uranium production cost in seawater uranium mining.« less

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

Washiya, Tadahiro; Komaki, Jun; Funasaka, Hideyuki

Japan Atomic Energy Agency (JAEA) has been developing the new aqueous reprocessing system named 'NEXT' (New Extraction system for TRU recovery)1-2, which provides many advantages as waste volume reduction, cost savings by advanced components and simplification of process operation. Advanced head-end systems in the 'NEXT' process consist of fuel disassembly system, fuel shearing system and continuous dissolver system. We developed reliable fuel disassembly system with innovative procedure, and short-length shearing system and continuous dissolver system can be provided highly concentrated dissolution to adapt to the uranium crystallization process. We have carried out experimental studies, and fabrication of engineering-scale test devicesmore » to confirm the systems performance. In this paper, research and development of advanced head-end systems are described. (authors)« less

Removal of radium from acidic solutions containing same by adsorption on coal fly ash

DOEpatents

Scheitlin, Frank M.

1984-01-01

The invention is a process for the removal of radium from acidic aqueous solutions. In one aspect, the invention is a process for removing radium from an inorganic-acid solution. The process comprises contacting the solution with coal fly ash to effect adsorption of the radium on the ash. The radium-containing ash then is separated from the solution. The process is simple, comparatively inexpensive, and efficient. High radium-distribution coefficients are obtained even at room temperature. Coal fly ash is an inexpensive, acid-resistant, high-surface-area material which is available in large quantities throughout the United States. The invention is applicable, for example, to the recovery of .sup.226 Ra from nitric acid solutions which have been used to leach radium from uranium-mill tailings.

Plutonium Decontamination of Uranium using CO2 Cleaning

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

Blau, M

A concern of the Department of Energy (DOE) Environmental Management (EM) and Defense Programs (DP), and of the Los Alamos National Laboratory (LANL) and the Lawrence Livermore National Laboratory (LLNL), is the disposition of thousands of legacy and recently generated plutonium (Pu)-contaminated, highly enriched uranium (HEU) parts. These parts take up needed vault space. This presents a serious problem for LLNL, as site limit could result in the stoppage of future weapons work. The Office of Fissile Materials Disposition (NN-60) will also face a similar problem as thousands of HEU parts will be created with the disassembly of site-return pitsmore » for plutonium recovery when the Pit Disassembly and Conversion Facility (PDCF) at the Savannah River Site (SRS) becomes operational. To send HEU to the Oak Ridge National Laboratory and the Y-12 Plant for disposition, the contamination for metal must be less than 20 disintegrations per minute (dpm) of swipable transuranic per 100 cm{sup 2} of surface area or the Pu bulk contamination for oxide must be less than 210 parts per billion (ppb). LANL has used the electrolytic process on Pu-contaminated HEU weapon parts with some success. However, this process requires that a different fixture be used for every configuration; each fixture cost approximately $10K. Moreover, electrolytic decontamination leaches the uranium metal substrate (no uranium or plutonium oxide) from the HEU part. The leaching rate at the uranium metal grain boundaries is higher than that of the grains and depends on the thickness of the uranium oxide layer. As the leaching liquid flows past the HEU part, it carries away plutonium oxide contamination and uranium oxide. The uneven uranium metal surface created by the leaching becomes a trap for plutonium oxide contamination. In addition, other DOE sites have used CO{sub 2} cleaning for Pu decontamination successfully. In the 1990's, the Idaho National Engineering Laboratory investigated this technology and showed that CO{sub 2} pellet blasting (or CO{sub 2} cleaning) reduced both fixed and smearable contamination on tools. In 1997, LLNL proved that even tritium contamination could be removed from a variety of different matrices using CO{sub 2}cleaning. CO{sub 2} cleaning is a non-toxic, nonconductive, nonabrasive decontamination process whose primary cleaning mechanisms are: (1) Impact of the CO{sub 2} pellets loosens the bond between the contaminant and the substrate. (2) CO{sub 2} pellets shatter and sublimate into a gaseous state with large expansion ({approx}800 times). The expanding CO{sub 2} gas forms a layer between the contaminant and the substrate that acts as a spatula and peels off the contaminant. (3) Cooling of the contaminant assists in breaking its bond with the substrate. Thus, LLNL conducted feasibility testing to determine if CO{sub 2} pellet blasting could remove Pu contamination (e.g., uranium oxide) from uranium metal without abrading the metal matrix. This report contains a summary of events and the results of this test.« less

ELUTION OF URANIUM FROM RESIN

DOEpatents

McLEan, D.C.

1959-03-10

A method is described for eluting uranium from anion exchange resins so as to decrease vanadium and iron contamination and permit recycle of the major portion of the eluats after recovery of the uranium. Diminution of vanadium and iron contamination of the major portion of the uranium is accomplished by treating the anion exchange resin, which is saturated with uranium complex by adsorption from a sulfuric acid leach liquor from an ore bearing uranium, vanadium and iron, with one column volume of eluant prepared by passing chlorine into ammonium hydroxide until the chloride content is about 1 N and the pH is about 1. The resin is then eluted with 8 to 9 column volumes of 0.9 N ammonium chloride--0.1 N hydrochloric acid solution. The eluants are collected separately and treated with ammonia to precipitate ammonium diuranate which is filtered therefrom. The uranium salt from the first eluant is contaminated with the major portion of ths vanadium and iron and is reworked, while the uranium recovered from the second eluant is relatively free of the undesirable vanadium and irons. The filtrate from the first eluant portion is discarded. The filtrate from the second eluant portion may be recycled after adding hydrochloric acid to increase the chloride ion concentration and adjust the pH to about 1.

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

Characterization of the biochemical-pathway of uranium (VI) reduction in facultative anaerobic bacteria.

PubMed

Mtimunye, Phalazane J; Chirwa, Evans M N

2014-10-01

Cultures of U(VI) reducing bacteria sourced from abandoned uranium mine tailing dam were evaluated for their ability to reduce U(VI) to U(IV). The species in the cultures reduced U(VI) in solutions with initial U(VI) concentration up to 400mgL(-)(1) under a near neutral pH of 6.5. The electron flow pathway and fate of reduced species was also analysed in the individual species in order to evaluate the potential for control and optimisation of the reduction potential at the biochemical level. The results showed that U(VI) reduction in live cells was completely blocked by the NADH-dehydrogenase inhibitor, rotenone (C23H22O6), and thioredoxin inhibitor, cadmium chloride (CdCl2), showing that U(VI) reduction involves the electron flow through NADH-dehydrogenase, a primary electron donor to the electron transport respiratory (ETR) system. Mass balance analysis of uranium species aided by visual and electron microscopy suggest that most U(VI) reduction occurred on the cell surface of the isolated species. This finding indicates the possibility of easy uranium recovery for beneficial use through biological remediation. Should the U(VI) be reduced inside the cell, recovery would require complete disruption of the cells and therefore would be difficult. The study contributes new knowledge on the underlying mechanisms in the U(VI) reduction in facultative anaerobes. Copyright © 2014 Elsevier Ltd. All rights reserved.

Geochemical and radiological characterization of soils from former radium processing sites

USGS Publications Warehouse

Landa, E.R.

1984-01-01

Soil samples were collected from former radium processing sites in Denver, CO, and East Orange, NJ. Particle-size separations and radiochemical analyses of selected samples showed that while the greatest contents of both 226Ra and U were generally found in the finest (< 45 ??m) fraction, the pattern was not always of progressive increase in radionuclide content with decreasing particle size. Leaching tests on these samples showed a large portion of the 225Ra and U to be soluble in dilute hydrochloric acid. Radon-emanation coefficients measured for bulk samples of contaminated soil were about 20%. Recovery of residual uranium and vanadium, as an adjunct to any remedial action program, appears unlikely due to economic considerations.

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.

Determination of uranium in tap water by ICP-MS.

PubMed

El Himri, M; Pastor, A; de la Guardia, M

2000-05-01

A fast and accurate procedure has been developed for the determination of uranium at microg L(-1) level in tap and mineral water. The method is based on the direct introduction of samples, without any chemical pre-treatment, into an inductively coupled plasma mass spectrometer (ICP-MS). Uranium was determined at the mass number 238 using Rh as internal standard. The method provides a limit of detection of 2 ng L(-1) and a good repeatability with relative standard deviation values (RSD) about 3% for five independent analyses of samples containing 73 microg L(-1) of uranium. Recovery percentage values found for the determination of uranium in spiked natural samples varied between 91% and 106%. Results obtained are comparable with those found by radiochemical methods for natural samples and of the same order for the certified content of a reference material, thus indicating the accuracy of the ICP-MS procedure without the need of using isotope dilution. A series of mineral and tap waters from different parts of Spain and Morocco were analysed.

Engineering assessment of inactive uranium mill tailings: Maybell Site, Maybell, Colorado

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

None

1981-09-01

Ford, Bacon and Davis Utah Inc. has reevaluated the Maybell site in order to revise the October 1977 engineering assessment of the problems resulting from the existence of radioactive uranium mill tailings at Maybell, Colorado. This engineering assessment has included the preparation of topographic maps, the performance of core drillings and radiometric measurements sufficient to determine areas and volumes of tailings and radiation exposures of individuals and nearby populations, the investigations of site hydrology and meteorology, and the evaluation and costing of alternative corrective actions. Radon gas released from the 2.6 million dry tons of tailings at the Maybell sitemore » constitutes the most significant environmental impact, although windblown tailings and external gamma radiation also are factors. The two alternative actions presented in this engineering assessment range from millsite decontamination with the addition of 3 m of stabilization cover material (Option I), to disposal of the tailings in a nearby open pit mine and decontamination of the tailings site (Option II). Cost estimates for the two options are about $11,700,000 for stabilization in-place and about $22,700,000 for disposal within a distance of 2 mi. Three principal alternatives for the reprocessing of the Maybell tailings were examined: (a) heap leaching; (b) treatment at an existing mill; and (c) reprocessing at a new conventional mill constructed for tailings reprocessing. The cost of the uranium recovered would be about $125 and $165/lb of U/sub 3/O/sub 8/ by heap leach and conventional plant processes, respectively. The spot market price for uranium was $25/lb early in 1981. Therefore, reprocessing the tailings for uranium recovery is not economically attractive at present.« less

Engineering assessment of inactive uranium mill tailings: Maybell Site, Maybell, Colorado. Summary

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

None

1981-09-01

Ford, Bacon and Davis Utah Inc. has reevaluated the Maybell site in order to revise the October 1977 engineering assessment of the problems resulting from the existence of radioactive uranium mill tailings at Maybell, Colorado. This engineering assessment has included the preparation of topographic maps, the performance of core drillings and radiometric measurements sufficient to determine areas and volumes of tailings and radiation exposures of individuals and nearby populations, the investigations of site hydrology and meteorology, and the evaluation and costing of alternative corrective actions. Radon gas released from the 2.6 million dry tons of tailings at the Maybell sitemore » constitutes the most significant environmental impact, although windblown tailings and external gamma radiation also are factors. The two alternative actions presented in this engineering assessment range from millsite decontamination with the addition of 3 m of stabilization cover material (Option I), to disposal of the tailings in a nearby open pit mine and decontamination of the tailings site (Option II). Cost estimates for the two options are about $11,700,000 for stabilization in-place and about $22,700,000 for disposal within a distance of 2 mi. Three principal alternatives for the reprocessing of the Maybell tailings were examined: (a) heap leaching; (b) treatment at an existing mill; and (c) reprocessing at a new conventional mill constructed for tailings reprocessing. The cost of the uranium recovered would be about $125 and $165/lb of U/sub 3/O/sub 8/ by heap leach and conventional plant processes, respectively. The spot market price for uranium was $25/lb early in 1981. Therefore, reprocessing the tailings for uranium recovery is not economically attractive at present.« less

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.

76 FR 30696 - Reimbursement for Costs of Remedial Action at Active Uranium and Thorium Processing Sites

Federal Register 2010, 2011, 2012, 2013, 2014

2011-05-26

... DEPARTMENT OF ENERGY Reimbursement for Costs of Remedial Action at Active Uranium and Thorium...) acceptance of claims in FY 2011 from eligible active uranium and thorium processing site licensees for... incurred by licensees at active uranium and thorium processing sites to remediate byproduct material...

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.

Release behavior of uranium in uranium mill tailings under environmental conditions.

PubMed

Liu, Bo; Peng, Tongjiang; Sun, Hongjuan; Yue, Huanjuan

2017-05-01

Uranium contamination is observed in sedimentary geochemical environments, but the geochemical and mineralogical processes that control uranium release from sediment are not fully appreciated. Identification of how sediments and water influence the release and migration of uranium is critical to improve the prevention of uranium contamination in soil and groundwater. To understand the process of uranium release and migration from uranium mill tailings under water chemistry conditions, uranium mill tailing samples from northwest China were investigated with batch leaching experiments. Results showed that water played an important role in uranium release from the tailing minerals. The uranium release was clearly influenced by contact time, liquid-solid ratio, particle size, and pH under water chemistry conditions. Longer contact time, higher liquid content, and extreme pH were all not conducive to the stabilization of uranium and accelerated the uranium release from the tailing mineral to the solution. The values of pH were found to significantly influence the extent and mechanisms of uranium release from minerals to water. Uranium release was monitored by a number of interactive processes, including dissolution of uranium-bearing minerals, uranium desorption from mineral surfaces, and formation of aqueous uranium complexes. Considering the impact of contact time, liquid-solid ratio, particle size, and pH on uranium release from uranium mill tailings, reducing the water content, decreasing the porosity of tailing dumps and controlling the pH of tailings were the key factors for prevention and management of environmental pollution in areas near uranium mines. Copyright © 2017 Elsevier Ltd. All rights reserved.

The importance of colloids and mires for the transport of uranium isotopes through the Kalix River watershed and Baltic Sea

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

Porcelli, D.; Wasserburg, G.J.; Andersson, P.S.

The importance of colloids and organic deposits for the transport of uranium isotopes from continental source regions and through the estuarine environment was investigated in the mire-rich Kalix River drainage basin in northern Sweden and the Baltic Sea. Ultrafiltration techniques were used to separate uranium and other elements associated with colloids > 10 kD and >3 kD from {open_quotes}solute{close_quotes} uranium and provided consistent results and high recovery rates for uranium as well as for other elements from large volume samples. Uranium concentrations in 0.45 {mu}m-filtered Kalix River water samples increased by a factor of 3 from near the headwaters inmore » the Caledonides to the river mouth while major cation concentrations were relatively constant. {sup 234}U {sup 238}U ratios were high ({delta}{sup 234}U = 770-1500) throughout the basin, without showing any simple pattern, and required a supply of {sup 234}U-rich water. Throughout the Kalix River, a large fraction (30-90%) of the uranium is carried by >10 kD colloids, which is compatible with uranium complexation with humic acids. No isotopic differences were found between colloid-associated and solute uranium. Within the Baltic Sea, about half of the uranium is removed at low salinities. The proportion that is lost is equivalent to that of river-derived colloid-bound uranium, suggesting that while solute uranium behaves conservatively during estuarine mixing, colloid-bound uranium is lost due to rapid flocculation of colloidal material. The association of uranium with colloids therefore may be an important parameter in determining uranium estuarine behavior. Mire peats in the Kalix River highly concentrate uranium and are potentially a significant source of recoil {sup 234}U to the mirewaters and river waters. However, mirewater data clearly demonstrate that only small {sup 234}U/{sup 238}U shifts are generated relative to inflowing groundwater. 63 refs., 8 figs., 3 tabs.« less

ANIONIC EXCHANGE PROCESS FOR THE RECOVERY OF URANIUM AND VANADIUM FROM CARBONATE SOLUTIONS

DOEpatents

Bailes, R.H.; Ellis, D.A.; Long, R.S.

1958-12-16

Uranium and vanadium can be economically purified and recovered from non- salt roast carbonate leach liquors by adsorption on a strongly basic anionic exchange resin and subsequent selective elution by one of three alternative methods. Method 1 comprises selectively eluting uranium from the resin with an ammonium sulfate solution followed by eluting vanadium from the resin with either 5 M NaCl, saturated (NH/sub 4/)/sub 2/CO/sub 3/, saturated NaHCO/sub 3/, 1 M NaOH, or saturated S0/sub 2/ solutions. Method II comprises selectively eluting vanadium from the resin with either concentrated NaCl or S0/sub 2/ solutions subsequent to pretreatment of the column with either S0/sub 2/ gas, 1 N HCl, or 0.1 N H/sub 2/8O/sub 4/ followed by eluting uranium from the resin with solutions containing 0.9 M NH/sub 4/Cl or NaCl and 0.1 Cl. Method III comprises flowing the carbonate leac solutlon through a first column of a strongly basic anlonlc exchange resin untll vanadium breakthrough occurs, so that the effluent solution is enriched ln uranium content and the vanadium is chiefly retalned by the resln, absorbing the uranlum from the enriched effluent solution on a second column of a strongly basic anionic exchange resin, pretreating the first column with either 0.1 N HCl, 0.1 H/sub 2/SO/sub 4/, C0/sub 2/ gas, or ammonium sulfate, selectively eluting the vanadlum from the column with saturated S0/sub 2/ solution, pretreatlng the second column with either 0.1 N HCl or S0/sub 2/ gas, selectively eluting residual vanadium from the column with saturated S0/sub 2/ solution, and then eluting the uranium from the column with either 0.1 N HCl and 1 N NaCl orO.l N HCl and 1 N NH/sub 4/Cl.

PROCESS OF RECOVERING URANIUM

DOEpatents

Carter, J.M.; Larson, C.E.

1958-10-01

A process is presented for recovering uranium values from calutron deposits. The process consists in treating such deposits to produce an oxidlzed acidic solution containing uranium together with the following imparities: Cu, Fe, Cr, Ni, Mn, Zn. The uranium is recovered from such an impurity-bearing solution by adjusting the pH of the solution to the range 1.5 to 3.0 and then treating the solution with hydrogen peroxide. This results in the precipitation of uranium peroxide which is substantially free of the metal impurities in the solution. The peroxide precipitate is then separated from the solution, washed, and calcined to produce uranium trioxide.

Biomining: metal recovery from ores with microorganisms.

PubMed

Schippers, Axel; Hedrich, Sabrina; Vasters, Jürgen; Drobe, Malte; Sand, Wolfgang; Willscher, Sabine

2014-01-01

Biomining is an increasingly applied biotechnological procedure for processing of ores in the mining industry (biohydrometallurgy). Nowadays the production of copper from low-grade ores is the most important industrial application and a significant part of world copper production already originates from heap or dump/stockpile bioleaching. Conceptual differences exist between the industrial processes of bioleaching and biooxidation. Bioleaching is a conversion of an insoluble valuable metal into a soluble form by means of microorganisms. In biooxidation, on the other hand, gold is predominantly unlocked from refractory ores in large-scale stirred-tank biooxidation arrangements for further processing steps. In addition to copper and gold production, biomining is also used to produce cobalt, nickel, zinc, and uranium. Up to now, biomining has merely been used as a procedure in the processing of sulfide ores and uranium ore, but laboratory and pilot procedures already exist for the processing of silicate and oxide ores (e.g., laterites), for leaching of processing residues or mine waste dumps (mine tailings), as well as for the extraction of metals from industrial residues and waste (recycling). This chapter estimates the world production of copper, gold, and other metals by means of biomining and chemical leaching (bio-/hydrometallurgy) compared with metal production by pyrometallurgical procedures, and describes new developments in biomining. In addition, an overview is given about metal sulfide oxidizing microorganisms, fundamentals of biomining including bioleaching mechanisms and interface processes, as well as anaerobic bioleaching and bioleaching with heterotrophic microorganisms.

Strategic Aspects of Pacific Cooperation Proposals,

DTIC Science & Technology

1983-08-01

partner, as a consequence of rapid economic growth that began after a postwar recovery . Roughly 13% of the USA’s imports come from Japan, and approximately...38- uranium , minerals, and agriculture products. Production interdependencies between the two countries are small, as the 80... Recovery ", Pacific Affairs, 53, 2, Summer, 1980, 245-268. 26. See Lowell Dittmer, "China in 1980: Modernization and its .- Discontents" Asian Survey

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

Doyle, Jamie L.; Kuhn, Kevin John; Byerly, Benjamin

Nuclear forensic publications, performance tests, and research and development efforts typically target the bulk global inventory of intentionally safeguarded materials, such as plutonium (Pu) and uranium (U). Other materials, such as neptunium (Np), pose a nuclear security risk as well. Trafficking leading to recovery of an interdicted Np sample is a realistic concern especially for materials originating in countries that reprocesses fuel. Using complementary forensic methods, potential signatures for an unknown Np oxide sample were investigated. Measurement results were assessed against published Np processes to present hypotheses as to the original intended use, method of production, and origin for thismore » Np oxide.« less

Bioprocessing of ores: Application to space resources

NASA Technical Reports Server (NTRS)

Johansson, Karl R.

1992-01-01

The role of microorganisms in the oxidation and leaching of various ores (especially those of copper, iron, and uranium) is well known. This role is increasingly being applied by the mining, metallurgy, and sewage industries in the bioconcentration of metal ions from natural receiving waters and from waste waters. It is concluded that bioprocessing using bacteria in closed reactors may be a variable option for the recovery of metals from the lunar regolith. Obviously, considerable research must be done to define the process, specify the appropriate bacteria, determine the necessary conditions and limitations, and evaluate the overall feasibility.

Dissolution of synthetic uranium dibutyl phosphate deposits in oxidizing and reducing chemical formulations.

PubMed

Rufus, A L; Sathyaseelan, V S; Narasimhan, S V; Velmurugan, S

2013-06-15

Permanganate and nitrilotriacetic acid (NTA) based dilute chemical formulations were evaluated for the dissolution of uranium dibutyl phosphate (U-DBP), a compound that deposits over the surfaces of nuclear reprocessing plants and waste storage tanks. A combination of an acidic, oxidizing treatment (nitric acid with permanganate) followed by reducing treatment (NTA based formulation) efficiently dissolved the U-DBP deposits. The dissolution isotherm of U-DBP in its as precipitated form followed a logarithmic fit. The same chemical treatment was also effective in dissolving U-DBP coated on the surface of 304-stainless steel, while resulting in minimal corrosion of the stainless steel substrate material. Investigation of uranium recovery from the resulting decontamination solutions by ion exchange with a bed of mixed anion and cation resins showed quantitative removal of uranium. Copyright © 2013 Elsevier B.V. All rights reserved.

Inhalation and Ingestion Intakes with Associated Dose Estimates for Level II and Level III Personnel Using Capstone Study Data

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

Szrom, Fran; Falo, Gerald A.; Lodde, Gordon M.

2009-03-01

Depleted uranium (DU) intake rates and subsequent dose rates were estimated for personnel entering armored combat vehicles perforated with DU penetrators (level II and level III personnel) using data generated during the Capstone Depleted Uranium (DU) Aerosol Study. Inhalation intake rates and associated dose rates were estimated from cascade impactors worn by sample recovery personnel and from cascade impactors that served as area monitors. Ingestion intake rates and associated dose rates were estimated from cotton gloves worn by sample recovery personnel and from wipe test samples from the interior of vehicles perforated with large caliber DU munitions. The mean DUmore » inhalation intake rate for level II personnel ranged from 0.447 mg h-1 based on breathing zone monitor data (in and around a perforated vehicle) to 14.5 mg h-1 based on area monitor data (in a perforated vehicle). The mean DU ingestion intake rate for level II ranged from 4.8 mg h-1 to 38.9 mg h-1 based on the wipe test data including surface to glove transfer factors derived from the Capstone data. Based on glove contamination data, the mean DU ingestion intake rates for level II and level III personnel were 10.6 mg h-1 was and 1.78 mg h-1, respectively. Effective dose rates and peak kidney uranium concentration rates were calculated based on the intake rates. The peak kidney uranium concentration rate cannot be multiplied by the total exposure duration when multiple intakes occur because uranium will clear from the kidney between the exposures.« less

SHINE and Mini-SHINE Column Designs for Recovery of Mo from 140 g-U/L Uranyl Sulfate

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

Stepinski, Dominique C.; Vandegrift, George F.

Argonne is assisting SHINE Medical Technologies (SHINE) in their efforts to develop an accelerator-driven process that utilizes a uranyl-sulfate solution for the production of fission Mo-99. In an effort to design a Mo-recovery system for the SHINE project using low-enriched uranium (LEU), we conducted batch, breakthrough, and pulse tests to determine the Mo isotherm, mass-transfer zone (MTZ), and system parameters for a 130 g-U/L uranyl sulfate solution at pH 1 and 80°C, as described previously. The VERSE program was utilized to calculate the MTZ under various loading times and velocities. The results were then used to design Mo separation andmore » recovery columns employing a pure titania sorbent (110-μm particles, S110, and 60 Å pore size). The plant-scale column designs assume Mo will be separated from 271 L of a 141 g-U/L uranyl sulfate solution, pH 1, containing 0.0023 mM Mo. The VERSE-designed recovery systems have been tested and verified in laboratory-scale experiments, and this approach was found to be very successful.« less

PROCESS OF PREPARING URANIUM CARBIDE

DOEpatents

Miller, W.E.; Stethers, H.L.; Johnson, T.R.

1964-03-24

A process of preparing uranium monocarbide is de scribed. Uranium metal is dissolved in cadmium, zinc, cadmium-- zinc, or magnesium-- zinc alloy and a small quantity of alkali metal is added. Addition of stoichiometric amounts of carbon at 500 to 820 deg C then precipitates uranium monocarbide. (AEC)

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

Hydrometallurgical methods of recovery of scandium from the wastes of various technologies

NASA Astrophysics Data System (ADS)

Molchanova, T. V.; Akimova, I. D.; Smirnov, K. M.; Krylova, O. K.; Zharova, E. V.

2017-03-01

The recovery of scandium from the wastes of the production of uranium, titanium, iron-vanadium, and alumina is studied. The applied acid schemes of scandium transfer to a solution followed by ion-exchange recovery and extraction concentration of scandium ensure the precipitation of crude scandium oxides containing up to 5% Sc2O3. Scandium oxides of 99.96-99.99% purity are formed after additional refining of these crude oxides according to an extraction technology using a mixture 15% multiradical phosphine oxide or Cyanex-925 + 15% tributyl phosphate in kerosene.

Geochemical and radiological characterization of soils from former radium processing sites.

PubMed

Landa, E R

1984-02-01

Soil samples were collected from former radium processing sites in Denver, CO, and East Orange, NJ. Particle-size separations and radiochemical analyses of selected samples showed that while the greatest contents of both 226Ra and U were generally found in the finest (less than 45 micron) fraction, the pattern was not always of progressive increase in radionuclide content with decreasing particle size. Leaching tests on these samples showed a large portion of the 226Ra and U to be soluble in dilute hydrochloric acid. Radon-emanation coefficients measured for bulk samples of contaminated soil were about 20%. Recovery of residual uranium and vanadium, as an adjunct to any remedial action program, appears unlikely due to economic considerations.

SOLVENT EXTRACTION PROCESS FOR URANIUM FROM CHLORIDE SOLUTIONS

DOEpatents

Blake, C.A. Jr.; Brown, K.B.; Horner, D.E.

1960-05-24

An improvement was made in a uranium extraction process wherein the organic extractant is a phosphine oxide. An aqueous solution containing phosphate ions or sulfate ions together with uranium is provided with a source of chloride ions during the extraction step. The presence of the chloride ions enables a phosphine oxide to extract uranium in the presence of strong uranium- complexing ions such as phosphate or sulfate ions.

Decontamination of uranium-contaminated waste oil using supercritical fluid and nitric acid.

PubMed

Sung, Jinhyun; Kim, Jungsoo; Lee, Youngbae; Seol, Jeunggun; Ryu, Jaebong; Park, Kwangheon

2011-07-01

The waste oil used in nuclear fuel processing is contaminated with uranium because of its contact with materials or environments containing uranium. Under current law, waste oil that has been contaminated with uranium is very difficult to dispose of at a radioactive waste disposal site. To dispose of the uranium-contaminated waste oil, the uranium was separated from the contaminated waste oil. Supercritical R-22 is an excellent solvent for extracting clean oil from uranium-contaminated waste oil. The critical temperature of R-22 is 96.15 °C and the critical pressure is 49.9 bar. In this study, a process to remove uranium from the uranium-contaminated waste oil using supercritical R-22 was developed. The waste oil has a small amount of additives containing N, S or P, such as amines, dithiocarbamates and dialkyldithiophosphates. It seems that these organic additives form uranium-combined compounds. For this reason, dissolution of uranium from the uranium-combined compounds using nitric acid was needed. The efficiency of the removal of uranium from the uranium-contaminated waste oil using supercritical R-22 extraction and nitric acid treatment was determined.

Aftermath of Uranium Ore Processing on Floodplains: Lasting Effects of Uranium on Soil and Microbes

NASA Astrophysics Data System (ADS)

Tang, H.; Boye, K.; Bargar, J.; Fendorf, S. E.

2016-12-01

A former uranium ore processing site located between the Wind River and the Little Wind River near the city of Riverton, Wyoming, has generated a uranium plume in the groundwater within the floodplain. Uranium is toxic and poses a threat to human health. Thus, controlling and containing the spread of uranium will benefit the human population. The primary source of uranium was removed from the processing site, but a uranium plume still exists in the groundwater. Uranium in its reduced form is relatively insoluble in water and therefore is retained in organic rich, anoxic layers in the subsurface. However, with the aid of microbes uranium becomes soluble in water which could expose people and the environment to this toxin, if it enters the groundwater and ultimately the river. In order to better understand the mechanisms controlling uranium behavior in the floodplains, we examined sediments from three sediment cores (soil surface to aquifer). We determined the soil elemental concentrations and measured microbial activity through the use of several instruments (e.g. Elemental Analyzer, X-ray Fluorescence, MicroResp System). Through the data collected, we aim to obtain a better understanding of how the interaction of geochemical factors and microbial metabolism affect uranium mobility. This knowledge will inform models used to predict uranium behavior in response to land use or climate change in floodplain environments.

10 CFR 765.11 - Reimbursable costs.

Code of Federal Regulations, 2011 CFR

2011-01-01

... DEPARTMENT OF ENERGY REIMBURSEMENT FOR COSTS OF REMEDIAL ACTION AT ACTIVE URANIUM AND THORIUM PROCESSING... uranium processing site licensees shall not exceed $6.25, as adjusted for inflation, multiplied by the... remedial action incurred at all active uranium processing sites shall not exceed $350 million. This...

10 CFR 765.11 - Reimbursable costs.

Code of Federal Regulations, 2012 CFR

2012-01-01

... DEPARTMENT OF ENERGY REIMBURSEMENT FOR COSTS OF REMEDIAL ACTION AT ACTIVE URANIUM AND THORIUM PROCESSING... uranium processing site licensees shall not exceed $6.25, as adjusted for inflation, multiplied by the... remedial action incurred at all active uranium processing sites shall not exceed $350 million. This...

10 CFR 765.11 - Reimbursable costs.

Code of Federal Regulations, 2013 CFR

2013-01-01

... DEPARTMENT OF ENERGY REIMBURSEMENT FOR COSTS OF REMEDIAL ACTION AT ACTIVE URANIUM AND THORIUM PROCESSING... uranium processing site licensees shall not exceed $6.25, as adjusted for inflation, multiplied by the... remedial action incurred at all active uranium processing sites shall not exceed $350 million. This...

10 CFR 765.11 - Reimbursable costs.

Code of Federal Regulations, 2014 CFR

2014-01-01

... DEPARTMENT OF ENERGY REIMBURSEMENT FOR COSTS OF REMEDIAL ACTION AT ACTIVE URANIUM AND THORIUM PROCESSING... uranium processing site licensees shall not exceed $6.25, as adjusted for inflation, multiplied by the... remedial action incurred at all active uranium processing sites shall not exceed $350 million. This...

10 CFR 765.11 - Reimbursable costs.

Code of Federal Regulations, 2010 CFR

2010-01-01

... DEPARTMENT OF ENERGY REIMBURSEMENT FOR COSTS OF REMEDIAL ACTION AT ACTIVE URANIUM AND THORIUM PROCESSING... uranium processing site licensees shall not exceed $6.25, as adjusted for inflation, multiplied by the... remedial action incurred at all active uranium processing sites shall not exceed $350 million. This...

Method of preparing uranium nitride or uranium carbonitride bodies

DOEpatents

Wilhelm, Harley A.; McClusky, James K.

1976-04-27

Sintered uranium nitride or uranium carbonitride bodies having a controlled final carbon-to-uranium ratio are prepared, in an essentially continuous process, from U.sub.3 O.sub.8 and carbon by varying the weight ratio of carbon to U.sub.3 O.sub.8 in the feed mixture, which is compressed into a green body and sintered in a continuous heating process under various controlled atmospheric conditions to prepare the sintered bodies.

Recovery of protactinium from molten fluoride nuclear fuel compositions

DOEpatents

Baes, C.F. Jr.; Bamberger, C.; Ross, R.G.

1973-12-25

A method is provided for separating protactinium from a molten fluonlde salt composition consisting essentially of at least one alkali and alkaline earth metal fluoride and at least one soluble fluoride of uranium or thorium which comprises oxidizing the protactinium in said composition to the + 5 oxidation state and contacting said composition with an oxide selected from the group consisting of an alkali metal oxide, an alkaline earth oxide, thorium oxide, and uranium oxide, and thereafter isolating the resultant insoluble protactinium oxide product from said composition. (Official Gazette)

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)

Survey Study of Trunk Materials for Direct ATRP Grafting

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

Saito, Tomonori; Chatterjee, Sabornie; Johnson, Joseph C.

2015-02-01

In previous study, we demonstrated a new method to prepare polymeric fiber adsorbents via a chemical-grafting method, namely atom-transfer radical polymerization (ATRP), and identified parameters affecting their uranium adsorption capacity. However, ATRP chemical grafting in the previous study still utilized conventional radiation-induced graft polymerization (RIGP) to introduce initiation sites on fibers. Therefore, the objective of the present study is to perform survey study of trunk fiber materials for direct ATRP chemical grafting method without RIGP for the preparation of fiber adsorbents for uranium recovery from seawater.

Radon gas: contractor liability for an indoor health hazard.

PubMed

Shuko, C M

1986-01-01

Many families throughout the United States have recently detected dangerously high concentrations of radon gas inside their homes. Radon, a carcinogenic gas produced from uranium, has been discovered in structures overlying uranium-bearing rock. This discovery may result in litigation to determine contractor liability for building upon radon-releasing rock sites. This Note examines the strengths and weaknesses of the various theories of contractor liability and considers potential statutory claims under the Clean Air Act. The Note suggests, as an alternative approach to recovery, a proposed regulatory scheme and implementation plan.

Alpha spectrometric characterization of process-related particle size distributions from active particle sampling at the Los Alamos National Laboratory uranium foundry

NASA Astrophysics Data System (ADS)

Plionis, A. A.; Peterson, D. S.; Tandon, L.; LaMont, S. P.

2010-03-01

Uranium particles within the respirable size range pose a significant hazard to the health and safety of workers. Significant differences in the deposition and incorporation patterns of aerosols within the respirable range can be identified and integrated into sophisticated health physics models. Data characterizing the uranium particle size distribution resulting from specific foundry-related processes are needed. Using personal air sampling cascade impactors, particles collected from several foundry processes were sorted by activity median aerodynamic diameter onto various Marple substrates. After an initial gravimetric assessment of each impactor stage, the substrates were analyzed by alpha spectrometry to determine the uranium content of each stage. Alpha spectrometry provides rapid non-distructive isotopic data that can distinguish process uranium from natural sources and the degree of uranium contribution to the total accumulated particle load. In addition, the particle size bins utilized by the impactors provide adequate resolution to determine if a process particle size distribution is: lognormal, bimodal, or trimodal. Data on process uranium particle size values and distributions facilitate the development of more sophisticated and accurate models for internal dosimetry, resulting in an improved understanding of foundry worker health and safety.

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.

PROCESS FOR UTILIZING ORGANIC ORTHOPHOSPHATE EXTRACTANTS

DOEpatents

Grinstead, R.R.

1958-11-11

A process is presented for recovering uranium from its ores, the steps comprising producing the uranium in solution in the trivalent state, extracting the uranium from solution in an lmmiscible organic solvent extract phase which lncludes mono and dialkyl orthophosphorlc acid esters having a varying number of carbon atoms on the alkyl substituent, amd recovering the uranium from tbe extract phase.

Uranium Adsorbent Fibers Prepared by Atom-Transfer Radical Polymerization from Chlorinated Polypropylene and Polyethylene Trunk Fibers

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

Brown, Suree; Chatterjee, Sabornie; Li, Meijun

Seawater contains a large amount of uranium (~4.5 billion tons) which can serve as a limitless supply of an energy source. However, in order to make the recovery of uranium from seawater economically feasible, lower manufacturing and deployment costs are required, and thus, solid adsorbents must have high uranium uptake, reusability, and high selectivity toward uranium. In this study, atom-transfer radical polymerization (ATRP), without the radiation-induced graft polymerization (RIGP), was used for grafting acrylonitrile (AN) and tert-butyl acrylate (tBA) from a new class of trunk fibers, forming adsorbents in a readily deployable form. The new class of trunk fibers wasmore » prepared by the chlorination of PP round fiber, hollow-gear-shaped PP fiber, and hollow-gear-shaped PE fiber. During ATRP, degrees of grafting (d.g.) varied according to the structure of active chlorine sites on trunk fibers and ATRP conditions, and the d.g. as high as 2570% was obtained. Resulting adsorbent fibers were evaluated in U-spiked simulated seawater and the maximum adsorption capacity of 146.6 g U/kg, much higher than that of a standard adsorbent JAEA fiber (75.1 g/kg), was obtained. This new type of trunk fibers can be used for grafting a variety of uranium-interacting ligands, including designed ligands that are highly selective toward uranium.« less

Uranium Adsorbent Fibers Prepared by Atom-Transfer Radical Polymerization from Chlorinated Polypropylene and Polyethylene Trunk Fibers

DOE PAGES

Brown, Suree; Chatterjee, Sabornie; Li, Meijun; ...

2015-12-10

Seawater contains a large amount of uranium (~4.5 billion tons) which can serve as a limitless supply of an energy source. However, in order to make the recovery of uranium from seawater economically feasible, lower manufacturing and deployment costs are required, and thus, solid adsorbents must have high uranium uptake, reusability, and high selectivity toward uranium. In this study, atom-transfer radical polymerization (ATRP), without the radiation-induced graft polymerization (RIGP), was used for grafting acrylonitrile (AN) and tert-butyl acrylate (tBA) from a new class of trunk fibers, forming adsorbents in a readily deployable form. The new class of trunk fibers wasmore » prepared by the chlorination of PP round fiber, hollow-gear-shaped PP fiber, and hollow-gear-shaped PE fiber. During ATRP, degrees of grafting (d.g.) varied according to the structure of active chlorine sites on trunk fibers and ATRP conditions, and the d.g. as high as 2570% was obtained. Resulting adsorbent fibers were evaluated in U-spiked simulated seawater and the maximum adsorption capacity of 146.6 g U/kg, much higher than that of a standard adsorbent JAEA fiber (75.1 g/kg), was obtained. This new type of trunk fibers can be used for grafting a variety of uranium-interacting ligands, including designed ligands that are highly selective toward uranium.« 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.

10 CFR 765.2 - Scope and applicability.

Code of Federal Regulations, 2014 CFR

2014-01-01

... DEPARTMENT OF ENERGY REIMBURSEMENT FOR COSTS OF REMEDIAL ACTION AT ACTIVE URANIUM AND THORIUM PROCESSING... uranium or thorium processing sites as a result of byproduct material generated as an incident of sales to the United States. (b) Costs of remedial action at active uranium or thorium processing sites are...

10 CFR 765.2 - Scope and applicability.

Code of Federal Regulations, 2011 CFR

2011-01-01

... DEPARTMENT OF ENERGY REIMBURSEMENT FOR COSTS OF REMEDIAL ACTION AT ACTIVE URANIUM AND THORIUM PROCESSING... uranium or thorium processing sites as a result of byproduct material generated as an incident of sales to the United States. (b) Costs of remedial action at active uranium or thorium processing sites are...

10 CFR 765.2 - Scope and applicability.

Code of Federal Regulations, 2013 CFR

2013-01-01

... DEPARTMENT OF ENERGY REIMBURSEMENT FOR COSTS OF REMEDIAL ACTION AT ACTIVE URANIUM AND THORIUM PROCESSING... uranium or thorium processing sites as a result of byproduct material generated as an incident of sales to the United States. (b) Costs of remedial action at active uranium or thorium processing sites are...

10 CFR 765.2 - Scope and applicability.

Code of Federal Regulations, 2012 CFR

2012-01-01

... DEPARTMENT OF ENERGY REIMBURSEMENT FOR COSTS OF REMEDIAL ACTION AT ACTIVE URANIUM AND THORIUM PROCESSING... uranium or thorium processing sites as a result of byproduct material generated as an incident of sales to the United States. (b) Costs of remedial action at active uranium or thorium processing sites are...

10 CFR 765.2 - Scope and applicability.

Code of Federal Regulations, 2010 CFR

2010-01-01

... DEPARTMENT OF ENERGY REIMBURSEMENT FOR COSTS OF REMEDIAL ACTION AT ACTIVE URANIUM AND THORIUM PROCESSING... uranium or thorium processing sites as a result of byproduct material generated as an incident of sales to the United States. (b) Costs of remedial action at active uranium or thorium processing sites are...

PROCESS FOR PRODUCTION OF URANIUM HEXAFLUORIDE

DOEpatents

Fowler, R.D.

1958-11-01

A process is described for the manufacture of uranium bexafluoride which consists in contacting an oxide of uranium simultaneously with elemental carbon and elemental fluorine at an elevated temperature, using a proportion of the carbon to the oxide about 50% in excess of that theoretically required to combine with f the oxygen as C0/.sub 2/. The process has the advantage that the uranium oxide is reduced by tbe carbon aad converted to the hexafluoride in a single operation.

The separation of lanthanides and actinides in supercritical fluid carbon dioxide

DOE PAGES

Mincher, Bruce J.; Wai, Chien M.; Fox, Robert V.; ...

2015-10-28

Supercritical fluid carbon dioxide presents an attractive alternative to conventional solvents for recovery of the actinides and lanthanides. Carbon dioxide is a good solvent for fluorine and phosphate-containing ligands, including the traditional tributylphosphate ligand used in process-scale uranium separations. Actinide and lanthanide oxides may even be directly dissolved in carbon dioxide containing the complexes formed between these ligands and mineral acids, obviating the need for large volumes of acids for leaching and dissolution, and the corresponding organic liquid–liquid solvent extraction solutions. As a result, examples of the application of this novel technology for actinide and lanthanide separations are presented.

Nuclear forensic analysis of a non-traditional actinide sample

DOE PAGES

Doyle, Jamie L.; Kuhn, Kevin John; Byerly, Benjamin; ...

2016-06-15

Nuclear forensic publications, performance tests, and research and development efforts typically target the bulk global inventory of intentionally safeguarded materials, such as plutonium (Pu) and uranium (U). Other materials, such as neptunium (Np), pose a nuclear security risk as well. Trafficking leading to recovery of an interdicted Np sample is a realistic concern especially for materials originating in countries that reprocesses fuel. Using complementary forensic methods, potential signatures for an unknown Np oxide sample were investigated. Measurement results were assessed against published Np processes to present hypotheses as to the original intended use, method of production, and origin for thismore » Np oxide.« less

Nuclear forensic analysis of a non-traditional actinide sample.

PubMed

Doyle, Jamie L; Kuhn, Kevin; Byerly, Benjamin; Colletti, Lisa; Fulwyler, James; Garduno, Katherine; Keller, Russell; Lujan, Elmer; Martinez, Alexander; Myers, Steve; Porterfield, Donivan; Spencer, Khalil; Stanley, Floyd; Townsend, Lisa; Thomas, Mariam; Walker, Laurie; Xu, Ning; Tandon, Lav

2016-10-01

Nuclear forensic publications, performance tests, and research and development efforts typically target the bulk global inventory of intentionally safeguarded materials, such as plutonium (Pu) and uranium (U). Other materials, such as neptunium (Np), pose a nuclear security risk as well. Trafficking leading to recovery of an interdicted Np sample is a realistic concern especially for materials originating in countries that reprocesses fuel. Using complementary forensic methods, potential signatures for an unknown Np oxide sample were investigated. Measurement results were assessed against published Np processes to present hypotheses as to the original intended use, method of production, and origin for this Np oxide. Published by Elsevier B.V.

PROCESS FOR PRODUCING ALKYL ORTHOPHOSPHORIC ACID EXTRACTANTS

DOEpatents

Grinstead, R.R.

1962-01-23

A process is given for producing superior alkyl orthophosphoric acid extractants for use in solvent extraction methods to recover and purify various metals such as uranium and vanadium. The process comprises slurrying P/sub 2/O/ sub 5/ in a solvent diluent such as kerosene, benzene, isopropyl ether, and the like. An alipbatic alcohol having from nine to seventeen carbon atoms, and w- hcrein ihc OH group is situated inward of the terminal carbon atoms, is added to the slurry while the reaction temperature is mainiained below 60 deg C. The alcohol is added in the mole ratio of about 2 to l, alcohol to P/sub 2/O/sub 5/. A pyrophosphate reaotion product is formed in the slurry-alcohol mixture. Subsequently, the pyrophosphate reaction product is hydrolyzed with dilute mineral acid to produce the desired alkyl orthophosphoric aeid extractant. The extraetant may then be separated and utilized in metal-recovery, solvent- extraction processes. (AEC)

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

Evapotranspiration And Geochemical Controls On Groundwater Plumes At Arid Sites: Toward Innovative Alternate End-States For Uranium Processing And Tailings Facilities

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

Looney, Brian B.; Denham, Miles E.; Eddy-Dilek, Carol A.

2014-01-08

Management of legacy tailings/waste and groundwater contamination are ongoing at the former uranium milling site in Tuba City AZ. The tailings have been consolidated and effectively isolated using an engineered cover system. For the existing groundwater plume, a system of recovery wells extracts contaminated groundwater for treatment using an advanced distillation process. The ten years of pump and treat (P&T) operations have had minimal impact on the contaminant plume – primarily due to geochemical and hydrological limits. A flow net analysis demonstrates that groundwater contamination beneath the former processing site flows in the uppermost portion of the aquifer and exitsmore » the groundwater as the plume transits into and beneath a lower terrace in the landscape. The evaluation indicates that contaminated water will not reach Moenkopi Wash, a locally important stream. Instead, shallow groundwater in arid settings such as Tuba City is transferred into the vadose zone and atmosphere via evaporation, transpiration and diffuse seepage. The dissolved constituents are projected to precipitate and accumulate as minerals such as calcite and gypsum in the deep vadose zone (near the capillary fringe), around the roots of phreatophyte plants, and near seeps. The natural hydrologic and geochemical controls common in arid environments such as Tuba City work together to limit the size of the groundwater plume, to naturally attenuate and detoxify groundwater contaminants, and to reduce risks to humans, livestock and the environment. The technical evaluation supports an alternative beneficial reuse (“brownfield”) scenario for Tuba City. This alternative approach would have low risks, similar to the current P&T scenario, but would eliminate the energy and expense associated with the active treatment and convert the former uranium processing site into a resource for future employment of local citizens and ongoing benefit to the Native American Nations.« less

FLAME DENITRATION AND REDUCTION OF URANIUM NITRATE TO URANIUM DIOXIDE

DOEpatents

Hedley, W.H.; Roehrs, R.J.; Henderson, C.M.

1962-06-26

A process is given for converting uranyl nitrate solution to uranium dioxide. The process comprises spraying fine droplets of aqueous uranyl nitrate solution into a hightemperature hydrocarbon flame, said flame being deficient in oxygen approximately 30%, retaining the feed in the flame for a sufficient length of time to reduce the nitrate to the dioxide, and recovering uranium dioxide. (AEC)

Modernization at the Y-12 National Security Complex: A Case for Additional Experimental Benchmarks

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

Thornbury, M. L.; Juarez, C.; Krass, A. W.

Efforts are underway at the Y-12 National Security Complex (Y-12) to modernize the recovery, purification, and consolidation of un-irradiated, highly enriched uranium metal. Successful integration of advanced technology such as Electrorefining (ER) eliminates many of the intermediate chemistry systems and processes that are the current and historical basis of the nuclear fuel cycle at Y-12. The cost of operations, the inventory of hazardous chemicals, and the volume of waste are significantly reduced by ER. It also introduces unique material forms and compositions related to the chemistry of chloride salts for further consideration in safety analysis and engineering. The work hereinmore » briefly describes recent investigations of nuclear criticality for 235UO2Cl2 (uranyl chloride) and 6LiCl (lithium chloride) in aqueous solution. Of particular interest is the minimum critical mass of highly enriched uranium as a function of the molar ratio of 6Li to 235U. The work herein also briefly describes recent investigations of nuclear criticality for 235U metal reflected by salt mixtures of 6LiCl or 7LiCl (lithium chloride), KCl (potassium chloride), and 235UCl3 or 238UCl3 (uranium tri-chloride). Computational methods for analysis of nuclear criticality safety and published nuclear data are employed in the absence of directly relevant experimental criticality benchmarks.« less

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

Zhang, Zhicheng; Rao, Linfeng; Abney, Carter W.

Adsorbents developed for the recovery of uranium from seawater display poor selectivity over other transition metals present in the ocean, with vanadium particularly problematic. To improve selectivity, an indispensable step is the positive identification of metal binding environments following actual seawater deployment. In this work we apply x-ray absorption fine structure (XAFS) spectroscopy to directly investigate the vanadium binding environment on seawater-deployed polyamidoxime adsorbents. Comparison of the x-ray absorption near edge spectra (XANES) reveal marked similarities to recently a reported non-oxido vanadium (V) structure formed upon binding with cyclic imidedioxime, a byproduct of generating amidoxime functionalities. Density functional theory (DFT)more » calculations provided a series of putative vanadium binding environments for both vanadium (IV) and vanadium (V) oxidation states, and with both amidoxime and cyclic imidedioxime. Fits of the extended XAFS (EXAFS) data confirmed vanadium (V) is bound exclusively by the cyclic imidedioxime moiety in a 1:2 metal:ligand fashion, though a modest structural distortion is also observed compared to crystal structure data and computationally optimized geometries which is attributed to morphology effects from the polymer graft chain and the absence of crystal packing interactions. These results demonstrate that improved selectivity for uranium over vanadium can be achieved by suppressing the formation of cyclic imidedioxime during preparation of polyamidoxime adsorbents for seawater uranium recovery.« less

Case Study in Corporate Memory Recovery: Hanford Tank Farms Miscellaneous Underground Waste Storage Tanks - 15344

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

Washenfelder, D. J.; Johnson, J. M.; Turknett, J. C.

In addition to managing the 177 underground waste storage tanks containing 212,000 m3 (56 million gal) of radioactive waste at the U. S. Department of Energy’s Hanford Site 200 Area Tank Farms, Washington River Protection Solutions LLC is responsible for managing numerous small catch tanks and special surveillance facilities. These are collectively known as “MUSTs” - Miscellaneous Underground Storage Tanks. The MUSTs typically collected drainage and flushes during waste transfer system piping changes; special surveillance facilities supported Tank Farm processes including post-World War II uranium recovery and later fission product recovery from tank wastes. Most were removed from service followingmore » deactivation of the single-shell tank system in 1980 and stabilized by pumping the remaining liquids from them. The MUSTs were isolated by blanking connecting transfer lines and adding weatherproofing to prevent rainwater entry. Over the next 30 years MUST operating records were dispersed into large electronic databases or transferred to the National Archives Regional Center in Seattle, Washington. During 2014 an effort to reacquire the historical bases for the MUSTs’ published waste volumes was undertaken. Corporate Memory Recovery from a variety of record sources allowed waste volumes to be initially determined for 21 MUSTs, and waste volumes to be adjusted for 37 others. Precursors and symptoms of Corporate Memory Loss were identified in the context of MUST records recovery.« less

78 FR 21352 - Update on Reimbursement for Costs of Remedial Action at Active Uranium and Thorium Processing Sites

Federal Register 2010, 2011, 2012, 2013, 2014

2013-04-10

... DEPARTMENT OF ENERGY Update on Reimbursement for Costs of Remedial Action at Active Uranium and Thorium Processing Sites AGENCY: Department of Energy. ACTION: Notice of the Title X claims during fiscal... at active uranium and thorium processing sites to remediate byproduct material generated as an...

40 CFR 471.70 - Applicability; description of the uranium forming subcategory.

Code of Federal Regulations, 2011 CFR

2011-07-01

... uranium forming subcategory. 471.70 Section 471.70 Protection of Environment ENVIRONMENTAL PROTECTION... SOURCE CATEGORY Uranium Forming Subcategory § 471.70 Applicability; description of the uranium forming... introductions of pollutants into publicly owned treatment works from the process operations of the uranium...

40 CFR 471.70 - Applicability; description of the uranium forming subcategory.

Code of Federal Regulations, 2010 CFR

2010-07-01

... uranium forming subcategory. 471.70 Section 471.70 Protection of Environment ENVIRONMENTAL PROTECTION... SOURCE CATEGORY Uranium Forming Subcategory § 471.70 Applicability; description of the uranium forming... introductions of pollutants into publicly owned treatment works from the process operations of the uranium...

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)

31 CFR 540.316 - Uranium enrichment.

Code of Federal Regulations, 2013 CFR

2013-07-01

... 31 Money and Finance:Treasury 3 2013-07-01 2013-07-01 false Uranium enrichment. 540.316 Section... FOREIGN ASSETS CONTROL, DEPARTMENT OF THE TREASURY HIGHLY ENRICHED URANIUM (HEU) AGREEMENT ASSETS CONTROL REGULATIONS General Definitions § 540.316 Uranium enrichment. The term uranium enrichment means the process of...

31 CFR 540.316 - Uranium enrichment.

Code of Federal Regulations, 2014 CFR

2014-07-01

... 31 Money and Finance:Treasury 3 2014-07-01 2014-07-01 false Uranium enrichment. 540.316 Section... FOREIGN ASSETS CONTROL, DEPARTMENT OF THE TREASURY HIGHLY ENRICHED URANIUM (HEU) AGREEMENT ASSETS CONTROL REGULATIONS General Definitions § 540.316 Uranium enrichment. The term uranium enrichment means the process of...

31 CFR 540.316 - Uranium enrichment.

Code of Federal Regulations, 2011 CFR

2011-07-01

... 31 Money and Finance:Treasury 3 2011-07-01 2011-07-01 false Uranium enrichment. 540.316 Section... FOREIGN ASSETS CONTROL, DEPARTMENT OF THE TREASURY HIGHLY ENRICHED URANIUM (HEU) AGREEMENT ASSETS CONTROL REGULATIONS General Definitions § 540.316 Uranium enrichment. The term uranium enrichment means the process of...

31 CFR 540.316 - Uranium enrichment.

Code of Federal Regulations, 2012 CFR

2012-07-01

... 31 Money and Finance:Treasury 3 2012-07-01 2012-07-01 false Uranium enrichment. 540.316 Section... FOREIGN ASSETS CONTROL, DEPARTMENT OF THE TREASURY HIGHLY ENRICHED URANIUM (HEU) AGREEMENT ASSETS CONTROL REGULATIONS General Definitions § 540.316 Uranium enrichment. The term uranium enrichment means the process of...

Simultaneous extraction and preconcentration of uranium and thorium in aqueous samples by new modified mesoporous silica prior to inductively coupled plasma optical emission spectrometry determination.

PubMed

Yousefi, Seyed Reza; Ahmadi, Seyed Javad; Shemirani, Farzaneh; Jamali, Mohammad Reza; Salavati-Niasari, Masoud

2009-11-15

A new synthesized modified mesoporous silica (MCM-41) using 5-nitro-2-furaldehyde (fural) was applied as an effective sorbent for the solid phase extraction of uranium(VI) and thorium(IV) ions from aqueous solution for the measurement by inductively coupled plasma optical emission spectrometry (ICP OES). The influences of some analytical parameters on the quantitative recoveries of the analyte ions were investigated in batch method. Under optimal conditions, the analyte ions were sorbed by the sorbent at pH 5.5 and then eluted with 1.0 mL of 1.0 mol L(-1) HNO(3). The preconcentration factor was 100 for a 100mL sample volume. The limits of detection (LOD) obtained for uranium(VI) and thorium(IV) were 0.3 microg L(-1). The maximum sorption capacity of the modified MCM-41 was found to be 47 and 49 mg g(-1) for uranium(VI) and thorium(IV), respectively. The sorbent exhibited good stability, reusability, high adsorption capacity and fast rate of equilibrium for sorption/desorption of uranium and thorium ions. The applicability of the synthesized sorbent was examined using CRM and real water samples.

XAS and TRLIF spectroscopy of uranium and neptunium in seawater.

PubMed

Maloubier, Melody; Solari, Pier Lorenzo; Moisy, Philippe; Monfort, Marguerite; Den Auwer, Christophe; Moulin, Christophe

2015-03-28

Seawater contains radionuclides at environmental levels; some are naturally present and others come from anthropogenic nuclear activity. In this report, the molecular speciation in seawater of uranium(VI) and neptunium(V) at a concentration of 5 × 10(-5) M has been investigated for the first time using a combination of two spectroscopic techniques: Time-resolved laser-induced fluorescence (TRLIF) for U and extended X-ray absorption fine structure (EXAFS) for U and Np at the LIII edge. In parallel, the theoretical speciation of uranium and neptunium in seawater at the same concentration is also discussed and compared to spectroscopic data. The uranium complex was identified as the neutral carbonato calcic complex UO2(CO3)3Ca2, which has been previously described in other natural systems. In the case of neptunium, the complex identified is mainly a carbonato complex whose exact stoichiometry is more difficult to assess. The knowledge of the actinide molecular speciation and reactivity in seawater is of fundamental interest in the particular case of uranium recovery and more generally regarding the actinide life cycle within the biosphere in the case of accidental release. This is the first report of actinide direct speciation in seawater medium that can complement inventory data.

Development of practical decontamination process for the removal of uranium from gravel.

PubMed

Kim, Ilgook; Kim, Gye-Nam; Kim, Seung-Soo; Choi, Jong-Won

2018-01-01

In this study, a practical decontamination process was developed to remove uranium from gravel using a soil washing method. The effects of critical parameters including particle size, H 2 SO 4 concentration, temperature, and reaction time on uranium removal were evaluated. The optimal condition for two-stage washing of gravel was found to be particle size of 1-2 mm, 1.0 M H 2 SO 4 , temperature of 60°C, and reaction time of 3 h, which satisfied the required uranium concentration for self-disposal. Furthermore, most of the extracted uranium was removed from the waste solution by precipitation, implying that the treated solution can be reused as washing solution. These results clearly demonstrated that our proposed process can be indeed a practical technique to decontaminate uranium-polluted gravel.

PROCESSES OF RECOVERING URANIUM FROM A CALUTRON

DOEpatents

Baird, D.O.; Zumwalt, L.R.

1958-07-15

An improved process is described for recovering the residue of a uranium compound which has been subjected to treatment in a calutron, from the parts of the calutron disposed in the source region upon which the residue is deposited. The process may be utilized when the uranium compound adheres to a surface containing metals of the group consisting of copper, iron, chromium, and nickel. The steps comprise washing the surface with an aqueous acidic oxidizing solvent for the uranium whereby there is obtained an acidic aqueous Solution containing uranium as uranyl ions and metals of said group as impurities, treating the acidic solution with sodium acetate in the presenee of added sodium nitrate to precipitate the uranium as sodium uranyl acetate away from the impurities in the solution, and separating the sodium uranyl acetate from the solution.

Development of experimental approach to examine U occurrence continuity over the extended area reconnoitory boreholes: Lostoin Block, West Khasi Hills district, Meghalaya (India).

PubMed

Kukreti, B M; Kumar, Pramod; Sharma, G K

2015-10-01

Exploratory drilling was undertaken in the Lostoin block, West Khasi Hills district of Meghalaya based on the geological extension to the major uranium deposit in the basin. Gamma ray logging of drilled boreholes shows considerable subsurface mineralization in the block. However, environmental and exploration related challenges such as climatic, logistic, limited core drilling and poor core recovery etc. in the block severely restricted the study of uranium exploration related index parameters for the block with a high degree confidence. The present study examines these exploration related challenges and develops an integrated approach using representative sampling of reconnoitory boreholes in the block. Experimental findings validate a similar geochemically coherent nature of radio elements (K, Ra and Th) in the Lostoin block uranium hosting environment with respect to the known block of Mahadek basin and uranium enrichment is confirmed by the lower U to Th correlation index (0.268) of hosting environment. A mineralized zone investigation in the block shows parent (refers to the actual parent uranium concentration at a location and not a secondary concentration such as the daughter elements which produce the signal from a total gamma ray measurement) favoring uranium mineralization. The confidence parameters generated under the present study have implications for the assessment of the inferred category of uranium ore in the block and setting up a road map for the systematic exploration of large uranium potential occurring over extended areas in the basin amid prevailing environmental and exploratory impediments. Copyright © 2015 Elsevier Ltd. All rights reserved.

Quantitative determination of environmental levels of uranium, thorium and plutonium in bone by solvent extraction and alpha spectrometry

NASA Astrophysics Data System (ADS)

Singh, Narayani P.; Zimmerman, Carol J.; Lewis, Laura L.; Wrenn, McDonald E.

1984-06-01

Solvent extraction and alpha-spectrometry have been emplyed in the quantitative simultaneous determination of uranium. thorium and plutonium. The bone specimens, spiked with 232U, 229Th and 242Pu tracers, are wet ashed with HNO 3 followed by alternate additions of a new drops of HNO 3 and H 2O 2. Uranium is reduced to the tetravalent state with 200 mg SnCl 2 and 25 ml HI. Uranium, thorium and plutonium are then coprecipitated with calcium as oxalate, heated to 550°C, dissolved in 50 ml HCl, and the acidity adjusted to 10 M. Uranium and plutonium are extracted into a 20% tri-lauryl amine (TLA) solution in xylene, leaving thorium in the aqueous phase. Plutonium is first back-extracted from the TLA phase by shaking with a 1:1.5 volume of 0.05 M NH 4I in 8 M HCl, which reduces Pu(IV) to Pu(III). Uranium is then back-extracted with an equal volume of 0.1 M HCl. Thorium, which was left in the aqueous phase, is evaporated to dryness, dissolved in 4 M HNO 3, and the acidity adjusted to 4 M. Thorium is then extracted into 20% TLA solution in xylene pre-equilibrated with 4 M HNO 3, and back-extracted with 10 M HCl. Uranium, thorium, and plutonium are then electrodeposited separately onto platinum discs and counted by an alpha-spectrometer with a multi-channel analyzer and surface barrier silicon diodes. The mean recoveries of uranium, thorium, and plutonium in bovine, dog, and human bones were over 70%.

PROCESS FOR REMOVING NOBLE METALS FROM URANIUM

DOEpatents

Knighton, J.B.

1961-01-31

A pyrometallurgical method is given for purifying uranium containing ruthenium and palladium. The uranium is disintegrated and oxidized by exposure to air and then the ruthenium and palladium are extracted from the uranium with molten zinc.

PREPARATION OF URANIUM-ALUMINUM ALLOYS

DOEpatents

Moore, R.H.

1962-09-01

A process is given for preparing uranium--aluminum alloys from a solution of uranium halide in an about equimolar molten alkali metal halide-- aluminum halide mixture and excess aluminum. The uranium halide is reduced and the uranium is alloyed with the excess aluminum. The alloy and salt are separated from each other. (AEC)

Recovery of Uranium from Seawater: Modified Polyacrylonitrile Fibers as Selective Extractants

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

Alexandratos, Spiro D.

2017-03-15

A new bifunctional fiber has been prepared and found to have a significant loading capacity of uranium from real seawater. The fiber support is polyacrylonitrile and bifunctionality is provided by amidoxime and either diethylenetriamine (DETA) or ethylenediamine (EDA) ligands. The key feature is adjusting the hydrophilic /lipophilic balance within the fiber and this was best accomplished by partially acetylating or carboxylating EDA ligands. The bifunctional carboxylated EDA /AO fiber had a loading capacity of 3.83 mg U/g fiber at the Pacific Northwest National Laboratory with a 21 day contact time in real seawater.

PROCESS FOR PREPARING URANIUM METAL

DOEpatents

Prescott, C.H. Jr.; Reynolds, F.L.

1959-01-13

A process is presented for producing oxygen-free uranium metal comprising contacting iodine vapor with crude uranium in a reaction zone maintained at 400 to 800 C to produce a vaporous mixture of UI/sub 4/ and iodine. Also disposed within the maction zone is a tungsten filament which is heated to about 1600 C. The UI/sub 4/, upon contacting the hot filament, is decomposed to molten uranium substantially free of oxygen.

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.

Column Testing and 1D Reactive Transport Modeling to Evaluate Uranium Plume Persistence Processes

NASA Astrophysics Data System (ADS)

Johnson, R. H.; Morrison, S.; Morris, S.; Tigar, A.; Dam, W. L.; Dayvault, J.

2015-12-01

At many U.S. Department of Energy Office of Legacy Management sites, 100 year natural flushing was selected as a remedial option for groundwater uranium plumes. However, current data indicate that natural flushing is not occurring as quickly as expected and solid-phase and aqueous uranium concentrations are persistent. At the Grand Junction, Colorado office site, column testing was completed on core collected below an area where uranium mill tailings have been removed. The total uranium concentration in this core was 13.2 mg/kg and the column was flushed with laboratory-created water with no uranium and chemistry similar to the nearby Gunnison River. The core was flushed for a total of 91 pore volumes producing a maximum effluent uranium concentration of 6,110 μg/L at 2.1 pore volumes and a minimum uranium concentration of 36.2 μg/L at the final pore volume. These results indicate complex geochemical reactions at small pore volumes and a long tailing affect at greater pore volumes. Stop flow data indicate the occurrence of non-equilibrium processes that create uranium concentration rebound. These data confirm the potential for plume persistence, which is occurring at the field scale. 1D reactive transport modeling was completed using PHREEQC (geochemical model) and calibrated to the column test data manually and using PEST (inverse modeling calibration routine). Processes of sorption, dual porosity with diffusion, mineral dissolution, dispersion, and cation exchange were evaluated separately and in combination. The calibration results indicate that sorption and dual porosity are major processes in explaining the column test data. These processes are also supported by fission track photographs that show solid-phase uranium residing in less mobile pore spaces. These procedures provide valuable information on plume persistence and secondary source processes that may be used to better inform and evaluate remedial strategies, including natural flushing.

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

40 CFR 421.326 - Pretreatment standards for new sources.

Code of Federal Regulations, 2011 CFR

2011-07-01

... uranium processed in the refinery Chromium (total) 27.14 11.00 Copper 93.88 44.74 Nickel 40.34 27.14... uranium processed in the refinery Chromium (total) 1.689 0.685 Copper 5.844 2.785 Nickel 2.511 1.689... per million pounds) of uranium processed in the refinery Chromium (total) 2.357 0.955 Copper 8.152 3...

40 CFR 421.326 - Pretreatment standards for new sources.

Code of Federal Regulations, 2010 CFR

2010-07-01

... uranium processed in the refinery Chromium (total) 27.14 11.00 Copper 93.88 44.74 Nickel 40.34 27.14... uranium processed in the refinery Chromium (total) 1.689 0.685 Copper 5.844 2.785 Nickel 2.511 1.689... per million pounds) of uranium processed in the refinery Chromium (total) 2.357 0.955 Copper 8.152 3...

Pyroprocessing of Light Water Reactor Spent Fuels Based on an Electrochemical Reduction Technology

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

Ohta, Hirokazu; Inoue, Tadashi; Sakamura, Yoshiharu

A concept of pyroprocessing light water reactor (LWR) spent fuels based on an electrochemical reduction technology is proposed, and the material balance of the processing of mixed oxide (MOX) or high-burnup uranium oxide (UO{sub 2}) spent fuel is evaluated. Furthermore, a burnup analysis for metal fuel fast breeder reactors (FBRs) is conducted on low-decontamination materials recovered by pyroprocessing. In the case of processing MOX spent fuel (40 GWd/t), UO{sub 2} is separately collected for {approx}60 wt% of the spent fuel in advance of the electrochemical reduction step, and the product recovered through the rare earth (RE) removal step, which hasmore » the composition uranium:plutonium:minor actinides:fission products (FPs) = 76.4:18.4:1.7:3.5, can be applied as an ingredient of FBR metal fuel without a further decontamination process. On the other hand, the electroreduced alloy of high-burnup UO{sub 2} spent fuel (48 GWd/t) requires further decontamination of residual FPs by an additional process such as electrorefining even if RE FPs are removed from the alloy because the recovered plutonium (Pu) is accompanied by almost the same amount of FPs in addition to RE. However, the amount of treated materials in the electrorefining step is reduced to {approx}10 wt% of the total spent fuel owing to the prior UO{sub 2} recovery step. These results reveal that the application of electrochemical reduction technology to LWR spent oxide fuel is a promising concept for providing FBR metal fuel by a rationalized process.« less

Manganese biomining: A review.

PubMed

Das, A P; Sukla, L B; Pradhan, N; Nayak, S

2011-08-01

Biomining comprises of processing and extraction of metal from their ores and concentrates using microbial techniques. Currently this is used by the mining industry to extract copper, uranium and gold from low grade ores but not for low grade manganese ore in industrial scale. The study of microbial genomes, metabolites and regulatory pathways provide novel insights to the metabolism of bioleaching microorganisms and their synergistic action during bioleaching operations. This will promote understanding of the universal regulatory responses that the biomining microbial community uses to adapt to their changing environment leading to high metal recovery. Possibility exists of findings ways to imitate the entire process during industrial manganese biomining endeavor. This paper reviews the current status of manganese biomining research operations around the world, identifies factors that drive the selection of biomining as a processing technology, describes challenges in exploiting these innovations, and concludes with a discussion of Mn biomining's future. Copyright © 2011 Elsevier Ltd. All rights reserved.

Development of spent fuel reprocessing process based on selective sulfurization: Study on the Pu, Np and Am sulfurization

NASA Astrophysics Data System (ADS)

Kirishima, Akira; Amano, Yuuki; Nihei, Toshifumi; Mitsugashira, Toshiaki; Sato, Nobuaki

2010-03-01

For the recovery of fissile materials from spent nuclear fuel, we have proposed a novel reprocessing process based on selective sulfurization of fission products (FPs). The key concept of this process is utilization of unique chemical property of carbon disulfide (CS2), i.e., it works as a reductant for U3O8 but works as a sulfurizing agent for minor actinides and lanthanides. Sulfurized FPs and minor actinides (MA) are highly soluble to dilute nitric acid while UO2 and PuO2 are hardly soluble, therefore, FPs and MA can be removed from Uranium and Plutonium matrix by selective dissolution. As a feasibility study of this new concept, the sulfurization behaviours of U, Pu, Np, Am and Eu are investigated in this paper by the thermodynamical calculation, phase analysis of chemical analogue elements and tracer experiments.

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)

15. VIEW OF LABORATORY EQUIPMENT IN THE BUILDING 771 ANALYTICAL ...

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

15. VIEW OF LABORATORY EQUIPMENT IN THE BUILDING 771 ANALYTICAL LABORATORY. THE LAB ANALYZED SAMPLES FOR PLUTONIUM, AMERICIUM, URANIUM, NEPTUNIUM, AND OTHER RADIOACTIVE ISOTOPES. (9/25/62) - Rocky Flats Plant, Plutonium Recovery & Fabrication Facility, North-central section of plant, Golden, Jefferson County, CO

Solid-phase extraction microfluidic devices for matrix removal in trace element assay of actinide materials

DOE PAGES

Gao, Jun; Manard, Benjamin Thomas; Castro, Alonso; ...

2017-02-02

Advances in sample nebulization and injection technology have significantly reduced the volume of solution required for trace impurity analysis in plutonium and uranium materials. Correspondingly, we have designed and tested a novel chip-based microfluidic platform, containing a 100-µL or 20-µL solid-phase microextraction column, packed by centrifugation, which supports nuclear material mass and solution volume reductions of 90% or more compared to standard methods. Quantitative recovery of 28 trace elements in uranium was demonstrated using a UTEVA chromatographic resin column, and trace element recovery from thorium (a surrogate for plutonium) was similarly demonstrated using anion exchange resin AG MP-1. Of ninemore » materials tested, compatibility of polyvinyl chloride (PVC), polypropylene (PP), and polytetrafluoroethylene (PTFE) chips with the strong nitric acid media was highest. Finally, the microcolumns can be incorporated into a variety of devices and systems, and can be loaded with other solid-phase resins for trace element assay in high-purity metals.« less

Solid-phase extraction microfluidic devices for matrix removal in trace element assay of actinide materials

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

Gao, Jun; Manard, Benjamin Thomas; Castro, Alonso

Advances in sample nebulization and injection technology have significantly reduced the volume of solution required for trace impurity analysis in plutonium and uranium materials. Correspondingly, we have designed and tested a novel chip-based microfluidic platform, containing a 100-µL or 20-µL solid-phase microextraction column, packed by centrifugation, which supports nuclear material mass and solution volume reductions of 90% or more compared to standard methods. Quantitative recovery of 28 trace elements in uranium was demonstrated using a UTEVA chromatographic resin column, and trace element recovery from thorium (a surrogate for plutonium) was similarly demonstrated using anion exchange resin AG MP-1. Of ninemore » materials tested, compatibility of polyvinyl chloride (PVC), polypropylene (PP), and polytetrafluoroethylene (PTFE) chips with the strong nitric acid media was highest. Finally, the microcolumns can be incorporated into a variety of devices and systems, and can be loaded with other solid-phase resins for trace element assay in high-purity metals.« less

PROCESS OF ELECTROPLATING METALS WITH ALUMINUM

DOEpatents

Schickner, W.C.

1960-04-26

A process of electroplating aluminum on metals from a nonaqueous bath and a novel method of pretreating or conditioning the metal prior to electrodeposition of the aluminum are given. The process of this invention, as applied by way of example to the plating of uranium, comprises the steps of plating the uranium with the barrier inetal, immersing the barrier-coated uranium in fatty acid, and electrolyzing a water-free diethyl ether solution of aluminum chloride and lithium hydride while making the uranium the cathode until an aluminum deposit of the desired thickness has been formed. According to another preferred embodiment the barrier-coated uranium is immersed in an isopropyl alcohol solution of sterato chromic chloride prior to the fatty acid treatment of this invention.

Assessment of a Hydroxyapatite Permeable Reactive Barrier to Remediate Uranium at the Old Rifle Site Colorado.

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

Moore, Robert C.; Szecsody, James; Rigali, Mark J.

We have performed an initial evaluation and testing program to assess the effectiveness of a hydroxyapatite (Ca10(PO4)6(OH)2) permeable reactive barrier and source area treatment to decrease uranium mobility at the Department of Energy (DOE) former Old Rifle uranium mill processing site in Rifle, western Colorado. Uranium ore was processed at the site from the 1940s to the 1970s. The mill facilities at the site as well as the uranium mill tailings previously stored there have all been removed. Groundwater in the alluvial aquifer beneath the site still contains elevated concentrations of uranium, and is currently used for field tests tomore » study uranium behavior in groundwater and investigate potential uranium remediation technologies. The technology investigated in this work is based on in situ formation of apatite in sediment to create a subsurface apatite PRB and also for source area treatment. The process is based on injecting a solution containing calcium citrate and sodium into the subsurface for constructing the PRB within the uranium plume. As the indigenous sediment micro-organisms biodegrade the injected citrate, the calcium is released and reacts with the phosphate to form hydroxyapatite (precipitate). This paper reports on proof-of-principle column tests with Old Rifle sediment and synthetic groundwater.« less

10 CFR 765.20 - Procedures for submitting reimbursement claims.

Code of Federal Regulations, 2011 CFR

2011-01-01

... Section 765.20 Energy DEPARTMENT OF ENERGY REIMBURSEMENT FOR COSTS OF REMEDIAL ACTION AT ACTIVE URANIUM... reimbursement ceiling for any active uranium or thorium processing site; (5) Any revision in the per dry short ton limit on reimbursement for all active uranium processing sites; and (6) Any other relevant...

10 CFR 765.20 - Procedures for submitting reimbursement claims.

Code of Federal Regulations, 2013 CFR

2013-01-01

... Section 765.20 Energy DEPARTMENT OF ENERGY REIMBURSEMENT FOR COSTS OF REMEDIAL ACTION AT ACTIVE URANIUM... reimbursement ceiling for any active uranium or thorium processing site; (5) Any revision in the per dry short ton limit on reimbursement for all active uranium processing sites; and (6) Any other relevant...

10 CFR 765.20 - Procedures for submitting reimbursement claims.

Code of Federal Regulations, 2012 CFR

2012-01-01

... Section 765.20 Energy DEPARTMENT OF ENERGY REIMBURSEMENT FOR COSTS OF REMEDIAL ACTION AT ACTIVE URANIUM... reimbursement ceiling for any active uranium or thorium processing site; (5) Any revision in the per dry short ton limit on reimbursement for all active uranium processing sites; and (6) Any other relevant...

10 CFR 765.20 - Procedures for submitting reimbursement claims.

Code of Federal Regulations, 2014 CFR

2014-01-01

... Section 765.20 Energy DEPARTMENT OF ENERGY REIMBURSEMENT FOR COSTS OF REMEDIAL ACTION AT ACTIVE URANIUM... reimbursement ceiling for any active uranium or thorium processing site; (5) Any revision in the per dry short ton limit on reimbursement for all active uranium processing sites; and (6) Any other relevant...

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.)

Uranium from Seawater Program Review; Fuel Resources Uranium from Seawater Program DOE Office of Nuclear Energy

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

None

2013-07-01

For nuclear energy to remain sustainable in the United States, economically viable sources of uranium beyond terrestrial ores must be developed. The goal of this program is to develop advanced adsorbents that can extract uranium from seawater at twice the capacity of the best adsorbent developed by researchers at the Japan Atomic Energy Agency (JAEA), 1.5 mg U/g adsorbent. A multidisciplinary team from Oak Ridge National Laboratory, Lawrence Berkeley National Laboratory, Pacific Northwest National Laboratory, and the University of Texas at Austin was assembled to address this challenging problem. Polymeric adsorbents, based on the radiation grafting of acrylonitrile and methacrylicmore » acid onto high surface-area polyethylene fibers followed by conversion of the nitriles to amidoximes, have been developed. These poly(acrylamidoxime-co-methacrylic acid) fibers showed uranium adsorption capacities for the extraction of uranium from seawater that exceed 3 mg U/g adsorbent in testing at the Pacific Northwest National Laboratory Marine Sciences Laboratory. The essence of this novel technology lies in the unique high surface-area trunk material that considerably increases the grafting yield of functional groups without compromising its mechanical properties. This technology received an R&D100 Award in 2012. In addition, high surface area nanomaterial adsorbents are under development with the goal of increasing uranium adsorption capacity by taking advantage of the high surface areas and tunable porosity of carbon-based nanomaterials. Simultaneously, de novo structure-based computational design methods are being used to design more selective and stable ligands and the most promising candidates are being synthesized, tested and evaluated for incorporation onto a support matrix. Fundamental thermodynamic and kinetic studies are being carried out to improve the adsorption efficiency, the selectivity of uranium over other metals, and the stability of the adsorbents. Understanding the rate-limiting step of uranium uptake from seawater is also essential in designing an effective uranium recovery system. Finally, economic analyses have been used to guide these studies and highlight what parameters, such as capacity, recyclability, and stability, have the largest impact on the cost of extraction of uranium from seawater. Initially, the cost estimates by the JAEA for extraction of uranium from seawater with braided polymeric fibers functionalized with amidoxime ligands were evaluated and updated. The economic analyses were subsequently updated to reflect the results of this project while providing insight for cost reductions in the adsorbent development through “cradle-to-grave” case studies for the extraction process. This report highlights the progress made over the last three years on the design, synthesis, and testing of new materials to extract uranium for seawater. This report is organized into sections that highlight the major research activities in this project: (1) Chelate Design and Modeling, (2) Thermodynamics, Kinetics and Structure, (3) Advanced Polymeric Adsorbents by Radiation Induced Grafting, (4) Advanced Nanomaterial Adsorbents, (5) Adsorbent Screening and Modeling, (6) Marine Testing, and (7) Cost and Energy Assessment. At the end of each section, future research directions are briefly discussed to highlight the challenges that still remain to reduce the cost of extractions of uranium for seawater. Finally, contributions from the Nuclear Energy University Programs (NEUP), which complement this research program, are included at the end of this report.« less

Highly Preorganized Ligand 1,10-Phenanthroline-2,9-dicarboxylic Acid for the Selective Recovery of Uranium from Seawater in the Presence of Competing Vanadium Species

DOE PAGES

Lashley, Mark A.; Ivanov, Alexander S.; Bryantsev, Vyacheslav S.; ...

2016-09-30

Studies of the complexation of new promising ligands with uranyl (UO 2 2+) and other seawater cations can aid the development of more efficient, selective, and robust sorbents for the recovery of uranium from seawater. Here, we propose that the ligand design principles based on structural preorganization can be successfully applied to obtain a dramatic enhancement in UO 2 2+ ion binding affinity and selectivity. This concept is exemplified through the investigation of the com-plexes of UO 2 2+, VO 2+, and VO 2+ with the highly preorganized ligand PDA (1,10-phenanthroline-2,9-dicarboxylic acid) using a combination of fluores-cence and absorbance techniques,more » along with den-sity functional theory (DFT) calculations. Moreover, the measured stability constant value, log K1, of 16.5 for the UO 2 2+/PDA complex is very high compared to uranyl complexes with other dicarboxylic ligands. Moreover, PDA exhibits strong selectivity for uranyl over vanadium ions, since the determined sta-bility constant values of the PDA complexes of the vanadium ions are quite low (V(IV) log K1 = 7.4, V(V) = 7.3). Finally, the structures of the corresponding UO 2 2+, VO 2+, and VO 2+ complexes with PDA were identified by systematic DFT calculations, and helped to interpret the stronger binding affinity for uranium over the vanadium ions. Due to its high chemical stability, selectivity, and structural preor-ganization for UO 2 2+ complexation, PDA is a very promising candidate that can be potentially used in the development of novel adsorbent materials for the selective extraction of uranium from sea-water.« less

Highly Preorganized Ligand 1,10-Phenanthroline-2,9-dicarboxylic Acid for the Selective Recovery of Uranium from Seawater in the Presence of Competing Vanadium Species

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

Lashley, Mark A.; Ivanov, Alexander S.; Bryantsev, Vyacheslav S.

Studies of the complexation of new promising ligands with uranyl (UO 2 2+) and other seawater cations can aid the development of more efficient, selective, and robust sorbents for the recovery of uranium from seawater. Here, we propose that the ligand design principles based on structural preorganization can be successfully applied to obtain a dramatic enhancement in UO 2 2+ ion binding affinity and selectivity. This concept is exemplified through the investigation of the com-plexes of UO 2 2+, VO 2+, and VO 2+ with the highly preorganized ligand PDA (1,10-phenanthroline-2,9-dicarboxylic acid) using a combination of fluores-cence and absorbance techniques,more » along with den-sity functional theory (DFT) calculations. Moreover, the measured stability constant value, log K1, of 16.5 for the UO 2 2+/PDA complex is very high compared to uranyl complexes with other dicarboxylic ligands. Moreover, PDA exhibits strong selectivity for uranyl over vanadium ions, since the determined sta-bility constant values of the PDA complexes of the vanadium ions are quite low (V(IV) log K1 = 7.4, V(V) = 7.3). Finally, the structures of the corresponding UO 2 2+, VO 2+, and VO 2+ complexes with PDA were identified by systematic DFT calculations, and helped to interpret the stronger binding affinity for uranium over the vanadium ions. Due to its high chemical stability, selectivity, and structural preor-ganization for UO 2 2+ complexation, PDA is a very promising candidate that can be potentially used in the development of novel adsorbent materials for the selective extraction of uranium from sea-water.« less

Process for removing carbon from uranium

DOEpatents

Powell, George L.; Holcombe, Jr., Cressie E.

1976-01-01

Carbon contamination is removed from uranium and uranium alloys by heating in inert atmosphere to 700.degree.-1900.degree.C in effective contact with yttrium to cause carbon in the uranium to react with the yttrium. The yttrium is either in direct contact with the contaminated uranium or in indirect contact by means of an intermediate transport medium.

PRODUCTION OF URANIUM TETRACHLORIDE

DOEpatents

Calkins, V.P.

1958-12-16

A process is descrlbed for the production of uranium tetrachloride by contacting uranlum values such as uranium hexafluoride, uranlum tetrafluoride, or uranium oxides with either aluminum chloride, boron chloride, or sodium alumlnum chloride under substantially anhydrous condltlons at such a temperature and pressure that the chlorldes are maintained in the molten form and until the uranium values are completely converted to uranlum tetrachloride.

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.

Method for fabricating laminated uranium composites

DOEpatents

Chapman, L.R.

1983-08-03

The present invention is directed to a process for fabricating laminated composites of uranium or uranium alloys and at least one other metal or alloy. The laminated composites are fabricated by forming a casting of the molten uranium with the other metal or alloy which is selectively positioned in the casting and then hot-rolling the casting into a laminated plate in or around which the casting components are metallurgically bonded to one another to form the composite. The process of the present invention provides strong metallurgical bonds between the laminate components primarily since the bond disrupting surface oxides on the uranium or uranium alloy float to the surface of the casting to effectively remove the oxides from the bonding surfaces of the components.

Uranium dioxide electrolysis

DOEpatents

Willit, James L [Batavia, IL; Ackerman, John P [Prescott, AZ; Williamson, Mark A [Naperville, IL

2009-12-29

This is a single stage process for treating spent nuclear fuel from light water reactors. The spent nuclear fuel, uranium oxide, UO.sub.2, is added to a solution of UCl.sub.4 dissolved in molten LiCl. A carbon anode and a metallic cathode is positioned in the molten salt bath. A power source is connected to the electrodes and a voltage greater than or equal to 1.3 volts is applied to the bath. At the anode, the carbon is oxidized to form carbon dioxide and uranium chloride. At the cathode, uranium is electroplated. The uranium chloride at the cathode reacts with more uranium oxide to continue the reaction. The process may also be used with other transuranic oxides and rare earth metal oxides.

Sensitivity evaluation of the green alga Chlamydomonas reinhardtii to uranium by pulse amplitude modulated (PAM) fluorometry.

PubMed

Herlory, Olivier; Bonzom, Jean-Marc; Gilbin, Rodolphe

2013-09-15

Although ecotoxicological studies tend to address the toxicity thresholds of uranium in freshwaters, there is a lack of information on the effects of the metal on physiological processes, particularly in aquatic plants. Knowing that uranium alters photosynthesis via impairment of the water photo-oxidation process, we determined whether pulse amplitude modulated (PAM) fluorometry was a relevant tool for assessing the impact of uranium on the green alga Chlamydomonas reinhardtii and investigated how and to what extent uranium hampered photosynthetic performance. Photosynthetic activity and quenching were assessed from fluorescence induction curves generated by PAM fluorometry, after 1 and 5h of uranium exposure in controlled conditions. The oxygen-evolving complex (OEC) of PSII was identified as the primary action site of uranium, through alteration of the water photo-oxidation process as revealed by F0/Fv. Limiting re-oxidation of the plastoquinone pool, uranium impaired the electron flux between the photosystems until almost complete inhibition of the PSII quantum efficiency ( [Formula: see text] , EC50=303 ± 64 μg UL(-1) after 5h of exposure) was observed. Non-photochemical quenching (qN) was identified as the most sensitive fluorescence parameter (EC50=142 ± 98 μg UL(-1) after 5h of exposure), indicating that light energy not used in photochemistry was dissipated in non-radiative processes. It was shown that parameters which stemmed from fluorescence induction kinetics are valuable indicators for evaluating the impact of uranium on PSII in green algae. PAM fluorometry provided a rapid and reasonably sensitive method for assessing stress response to uranium in microalgae. Copyright © 2013 Elsevier B.V. All rights reserved.

75 FR 6068 - Extension of Public Comment Period on the Draft Environmental Impact Statement for the Lost Creek...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-02-05

... Environmental Impact Statement for the Lost Creek In-Situ Recovery Project in Sweetwater County, WY; Supplement to the Generic Environmental Impact Statement for In-Situ Leach Uranium Milling Facilities AGENCY... SEIS) for the Lost Creek [[Page 6069

77 FR 14837 - Bioassay at Uranium Mills

Federal Register 2010, 2011, 2012, 2013, 2014

2012-03-13

..., Revision 1 was issued. Comments related to the operation of in-situ recovery mills and associated... received on or before this date. Although a time limit is given, comments and suggestions in connection with items for inclusion in guides currently being developed or improvements in all published guides...

URANIUM SEPARATION PROCESS

DOEpatents

McVey, W.H.; Reas, W.H.

1959-03-10

The separation of uranium from an aqueous solution containing a water soluble uranyl salt is described. The process involves adding an alkali thiocyanate to the aqueous solution, contacting the resulting solution with methyl isobutyl ketons and separating the resulting aqueous and organic phase. The uranium is extracted in the organic phase as UO/sub 2/(SCN)/sub/.

PROCESS OF PRODUCING REFRACTORY URANIUM OXIDE ARTICLES

DOEpatents

Hamilton, N.E.

1957-12-01

A method is presented for fabricating uranium oxide into a shaped refractory article by introducing a uranium halide fluxing reagent into the uranium oxide, and then mixing and compressing the materials into a shaped composite mass. The shaped mass of uranium oxide and uranium halide is then fired at an elevated temperature so as to form a refractory sintered article. It was found in the present invention that the introduction of a uraninm halide fluxing agent afforded a fluxing action with the uranium oxide particles and that excellent cohesion between these oxide particles was obtained. Approximately 90% of uranium dioxide and 10% of uranium tetrafluoride represent a preferred composition.

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.

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.

High strength uranium-tungsten alloy process

DOEpatents

Dunn, Paul S.; Sheinberg, Haskell; Hogan, Billy M.; Lewis, Homer D.; Dickinson, James M.

1990-01-01

Alloys of uranium and tungsten and a method for making the alloys. The amount of tungsten present in the alloys is from about 4 wt % to about 35 wt %. Tungsten particles are dispersed throughout the uranium and a small amount of tungsten is dissolved in the uranium.

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.

Removal of uranium and fluorine from wastewater by double-functional microsphere adsorbent of SA/CMC loaded with calcium and aluminum

NASA Astrophysics Data System (ADS)

Wu, Liping; Lin, Xiaoyan; Zhou, Xingbao; Luo, Xuegang

2016-10-01

A novel dual functional microsphere adsorbent of alginate/carboxymethyl cellulose sodium composite loaded with calcium and aluminum (SA/CMC-Ca-Al) is prepared by an injection device to remove fluoride and uranium, respectively, from fluoro-uranium mixed aqueous solution. Batch experiments are performed at different conditions: pH, temperature, initial concentration and contact time. The results show that the maximum adsorption amount for fluoride is 35.98 mg/g at pH 2.0, 298.15 K concentration 100 mg/L, while that for uranium is 101.76 mg/g at pH 4.0, 298.15 K concentration 100 mg/L. Both of the adsorption process could be well described by Langmuir model. The adsorption kinetic data is fitted well with pseudo-first-order model for uranium and pseudo-second-order model for fluoride. Thermodynamic parameters are also evaluated, indicating that the adsorption of uranium on SA/CMC-Ca-Al is a spontaneous and exothermic process, while the removal of fluoride is non-spontaneous and endothermic process. The mechanism of modification and adsorption process on SA/CMC-Ca-Al is characterized by FT-IR, SEM, EDX and XPS. The results show that Ca (II) and Al (III) are loaded on SA/CMC through ion-exchange of sodium of SA/CMC. The coordination reaction and ion-exchange happen during the adsorption process between SA/CMC-Ca-Al and uranium, fluoride. Results suggest that the SA/CMC-Ca-Al adsorbent has a great potential in removing uranium and fluoride from aqueous solution.

Method of fabricating a uranium-bearing foil

DOEpatents

Gooch, Jackie G [Seymour, TN; DeMint, Amy L [Kingston, TN

2012-04-24

Methods of fabricating a uranium-bearing foil are described. The foil may be substantially pure uranium, or may be a uranium alloy such as a uranium-molybdenum alloy. The method typically includes a series of hot rolling operations on a cast plate material to form a thin sheet. These hot rolling operations are typically performed using a process where each pass reduces the thickness of the plate by a substantially constant percentage. The sheet is typically then annealed and then cooled. The process typically concludes with a series of cold rolling passes where each pass reduces the thickness of the plate by a substantially constant thickness amount to form the foil.

Determination of depleted uranium, pyridostigmine bromide and its metabolite in plasma and urine following combined administration in rats.

PubMed

Abu-Qare, A W; Abou-Donia, M B

2001-09-01

A simple and reliable method was developed for the quantification of depleted uranium, the anti nerve agent drug pyridostigmine bromide (PB;3-dimethylaminocarbonyloxy-N-methyl pyridinium bromide) and its metabolite N-methyl-3-hydroxypyridinium bromide in rat plasma and urine. The method involved using solid phase extraction and spectrophotometric determination of uranium, and high performance liquid chromatography (HPLC) with reversed phase C(18) column, and UV detection at 280 nm for PB and its metabolite. Uranium was derivatized using dibenzoylmethane (DBM) then the absorbance was measured at 405 nm. PB and its metabolite were separated using a gradient of 1--40% acetonitrile in 0.1% triflouroacetic acid water solution (pH 3.2) at a flow rate of 0.8 ml/min in a period of 14 min. Limits of detection were 2 ng/ml for uranium and 50 ng/ml for PB and its metabolite. Limits of quantitation were between 10 and 100 ng/ml for uranium and the other two analytes, respectively. Average percentage recovery of five spiked plasma samples were 83.7+/-8.6, 76.8+/-6.7, 79.1+/-7.1, and from urine 82.7+/-8.6, 79.3+/-9.5 and 78.0+/-6.2, for depleted uranium, PB and N-methyl-3-hydroxypyridinium bromide, respectively. The relationship between peak areas and concentration was linear for standards between 100 and 1000 ng/ml for all three analytes. This method was applied to analyze the above chemicals and metabolites following combined administration in rats.

ALPHA SPECTROMETRIC EVALUATION OF SRM-995 AS A POTENTIAL URANIUM/THORIUM DOUBLE TRACER SYSTEM FOR AGE-DATING URANIUM MATERIALS

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

Beals, D.

2011-12-06

Uranium-233 (t{sub 1/2} {approx} 1.59E5 years) is an artificial, fissile isotope of uranium that has significant importance in nuclear forensics. The isotope provides a unique signature in determining the origin and provenance of uranium-bearing materials and is valuable as a mass spectrometric tracer. Alpha spectrometry was employed in the critical evaluation of a {sup 233}U standard reference material (SRM-995) as a dual tracer system based on the in-growth of {sup 229}Th (t{sub 1/2} {approx} 7.34E3 years) for {approx}35 years following radiochemical purification. Preliminary investigations focused on the isotopic analysis of standards and unmodified fractions of SRM-995; all samples were separatedmore » and purified using a multi-column anion-exchange scheme. The {sup 229}Th/{sup 233}U atom ratio for SRM-995 was found to be 1.598E-4 ({+-} 4.50%) using recovery-corrected radiochemical methods. Using the Bateman equations and relevant half-lives, this ratio reflects a material that was purified {approx} 36.8 years prior to this analysis. The calculated age is discussed in contrast with both the date of certification and the recorded date of last purification.« less

31 CFR 540.316 - Uranium enrichment.

Code of Federal Regulations, 2010 CFR

2010-07-01

... 31 Money and Finance: Treasury 3 2010-07-01 2010-07-01 false Uranium enrichment. 540.316 Section 540.316 Money and Finance: Treasury Regulations Relating to Money and Finance (Continued) OFFICE OF... REGULATIONS General Definitions § 540.316 Uranium enrichment. The term uranium enrichment means the process of...

Uncertainty and variability in laboratory derived sorption parameters of sediments from a uranium in situ recovery site

NASA Astrophysics Data System (ADS)

Dangelmayr, Martin A.; Reimus, Paul W.; Johnson, Raymond H.; Clay, James T.; Stone, James J.

2018-06-01

This research assesses the ability of a GC SCM to simulate uranium transport under variable geochemical conditions typically encountered at uranium in-situ recovery (ISR) sites. Sediment was taken from a monitoring well at the SRH site at depths 192 and 193 m below ground and characterized by XRD, XRF, TOC, and BET. Duplicate column studies on the different sediment depths, were flushed with synthesized restoration waters at two different alkalinities (160 mg/l CaCO3 and 360 mg/l CaCO3) to study the effect of alkalinity on uranium mobility. Uranium breakthrough occurred 25% - 30% earlier in columns with 360 mg/l CaCO3 over columns fed with 160 mg/l CaCO3 influent water. A parameter estimation program (PEST) was coupled to PHREEQC to derive site densities from experimental data. Significant parameter fittings were produced for all models, demonstrating that the GC SCM approach can model the impact of carbonate on uranium in flow systems. Derived site densities for the two sediment depths were between 141 and 178 μmol-sites/kg-soil, demonstrating similar sorption capacities despite heterogeneity in sediment mineralogy. Model sensitivity to alkalinity and pH was shown to be moderate compared to fitted site densities, when calcite saturation was allowed to equilibrate. Calcite kinetics emerged as a potential source of error when fitting parameters in flow conditions. Fitted results were compared to data from previous batch and column studies completed on sediments from the Smith-Ranch Highland (SRH) site, to assess variability in derived parameters. Parameters from batch experiments were lower by a factor of 1.1 to 3.4 compared to column studies completed on the same sediments. The difference was attributed to errors in solid-solution ratios and the impact of calcite dissolution in batch experiments. Column studies conducted at two different laboratories showed almost an order of magnitude difference in fitted site densities suggesting that experimental methodology may play a bigger role in column sorption behavior than actual sediment heterogeneity. Our results demonstrate the necessity for ISR sites to remove residual pCO2 and equilibrate restoration water with background geochemistry to reduce uranium mobility. In addition, the observed variability between fitted parameters on the same sediments highlights the need to provide standardized guidelines and methodology for regulators and industry when the GC SCM approach is used for ISR risk assessments.

Uncertainty and variability in laboratory derived sorption parameters of sediments from a uranium in situ recovery site.

PubMed

Dangelmayr, Martin A; Reimus, Paul W; Johnson, Raymond H; Clay, James T; Stone, James J

2018-06-01

This research assesses the ability of a GC SCM to simulate uranium transport under variable geochemical conditions typically encountered at uranium in-situ recovery (ISR) sites. Sediment was taken from a monitoring well at the SRH site at depths 192 and 193 m below ground and characterized by XRD, XRF, TOC, and BET. Duplicate column studies on the different sediment depths, were flushed with synthesized restoration waters at two different alkalinities (160 mg/l CaCO 3 and 360 mg/l CaCO 3 ) to study the effect of alkalinity on uranium mobility. Uranium breakthrough occurred 25% - 30% earlier in columns with 360 mg/l CaCO 3 over columns fed with 160 mg/l CaCO 3 influent water. A parameter estimation program (PEST) was coupled to PHREEQC to derive site densities from experimental data. Significant parameter fittings were produced for all models, demonstrating that the GC SCM approach can model the impact of carbonate on uranium in flow systems. Derived site densities for the two sediment depths were between 141 and 178 μmol-sites/kg-soil, demonstrating similar sorption capacities despite heterogeneity in sediment mineralogy. Model sensitivity to alkalinity and pH was shown to be moderate compared to fitted site densities, when calcite saturation was allowed to equilibrate. Calcite kinetics emerged as a potential source of error when fitting parameters in flow conditions. Fitted results were compared to data from previous batch and column studies completed on sediments from the Smith-Ranch Highland (SRH) site, to assess variability in derived parameters. Parameters from batch experiments were lower by a factor of 1.1 to 3.4 compared to column studies completed on the same sediments. The difference was attributed to errors in solid-solution ratios and the impact of calcite dissolution in batch experiments. Column studies conducted at two different laboratories showed almost an order of magnitude difference in fitted site densities suggesting that experimental methodology may play a bigger role in column sorption behavior than actual sediment heterogeneity. Our results demonstrate the necessity for ISR sites to remove residual pCO2 and equilibrate restoration water with background geochemistry to reduce uranium mobility. In addition, the observed variability between fitted parameters on the same sediments highlights the need to provide standardized guidelines and methodology for regulators and industry when the GC SCM approach is used for ISR risk assessments. Copyright © 2018 Elsevier B.V. All rights reserved.

77 FR 62269 - Draft Tribal Protocol Manual and Scoping for Proposed Policy Statement

Federal Register 2010, 2011, 2012, 2013, 2014

2012-10-12

.... The NRC is committed to an open and collaborative regulatory environment in the development of its... NRC also seeks suggestions on the development of the proposed tribal consultation policy statement... uranium recovery development and new nuclear reactor construction has resulted in a significant increase...

77 FR 27254 - Sunshine Act Meeting

Federal Register 2010, 2011, 2012, 2013, 2014

2012-05-09

...-AI77) (Tentative). b. Strata Energy Inc. (Ross In Situ Recovery Uranium Project), Appeal by NRC Staff... participate in these public meetings, or need this meeting notice or the transcript or other information from the public meetings in another format (e.g. braille, large print), please notify Bill Dosch, Chief...

78 FR 65643 - Environmental Impact Statements; Notice of Availability

Federal Register 2010, 2011, 2012, 2013, 2014

2013-11-01

..., Gas Hills In-Situ Recovery Uranium Project, Review Period Ends: 12/02/2013, Contact: Tom Sunderland..., Tiering FEIS--U.S. Coast Guard Rulemaking for Dry Cargo Residue Discharges in the Great Lakes, Review... No. 20130290, Draft EIS, NPS, CA, Restoration of Native Species in High Elevation Aquatic Ecosystems...

78 FR 19261 - Environmental Impacts Statements; Notice of Availability

Federal Register 2010, 2011, 2012, 2013, 2014

2013-03-29

... No. 20130073, Draft Supplement, NRC, WY Ross In-Situ Leach Recovery (ISR) Project, Supplement to the Generic Environmental Impact Statement for In-Situ Leach Uranium Milling Facilities, Comment Period Ends... project. EIS No. 20130075, Draft Supplement, NMFS, AK, Effects of Oil and Gas Activities in the Arctic...

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

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)

METHOD FOR PURIFYING URANIUM

DOEpatents

Knighton, J.B.; Feder, H.M.

1960-04-26

A process is given for purifying a uranium-base nuclear material. The nuclear material is dissolved in zinc or a zinc-magnesium alloy and the concentration of magnesium is increased until uranium precipitates.

Urinary excretion of uranium in adult inhabitants of the Czech Republic.

PubMed

Malátová, Irena; Bečková, Věra; Kotík, Lukáš

2016-02-01

The main aim of this study was to determine and evaluate urinary excretion of uranium in the general public of the Czech Republic. This value should serve as a baseline for distinguishing possible increase in uranium content in population living near legacy sites of mining and processing uranium ores and also to help to distinguish the proportion of the uranium content in urine among uranium miners resulting from inhaled dust. The geometric mean of the uranium concentration in urine of 74 inhabitants of the Czech Republic was 0.091 mBq/L (7.4 ng/L) with the 95% confidence interval 0.071-0.12 mBq/L (5.7-9.6 ng/L) respectively. The geometric mean of the daily excretion was 0.15 mBq/d (12.4 ng/d) with the 95% confidence interval 0.12-0.20 mBq/d (9.5-16.1 ng/d) respectively. Despite the legacy of uranium mines and plants processing uranium ore in the Czech Republic, the levels of uranium in urine and therefore, also human body content of uranium, is similar to other countries, esp. Germany, Slovenia and USA. Significant difference in the daily urinary excretion of uranium was found between individuals using public supply and private water wells as a source of drinking water. Age dependence of daily urinary excretion of uranium was not found. Mean values and their range are comparable to other countries, esp. Germany, Slovenia and USA. Copyright © 2015 Elsevier Ltd. All rights reserved.

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.

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.

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.

PROCESSES OF RECLAIMING URANIUM FROM SOLUTIONS

DOEpatents

Zumwalt, L.R.

1959-02-10

A process is described for reclaiming residual enriched uranium from calutron wash solutions containing Fe, Cr, Cu, Ni, and Mn as impurities. The solution is adjusted to a pH of between 2 and 4 and is contacted with a metallic reducing agent, such as iron or zinc, in order to reduce the copper to metal and thereby remove it from the solution. At the same time the uranium present is reduced to the uranous state The solution is then contacted with a precipitate of zinc hydroxide or barium carbonate in order to precipitate and carry uranium, iron, and chromium away from the nickel and manganese ions in the solution. The uranium is then recovered fronm this precipitate.

Extraction of Uranium from Seawater: Design and Testing of a Symbiotic System

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

Slocum, Alex

The U.S. Department of Energy in October 2014 awarded the Massachusetts Institute of Technology (MIT) a Nuclear Energy University Program grant (DE-NE0008268) to investigate the design and testing of a symbiotic system to harvest uranium from seawater. As defined in the proposal, the goals for the project are: 1. Address the design of machines for seawater uranium mining. 2. Develop design rules for a uranium harvesting system that would be integrated into an offshore wind power tower. 3. Fabricate a 1/50th size scale prototype for bench and pool-testing to verify initial analysis and theory. 4. Design, build, and test amore » second 1/10th size scale prototype in the ocean for more comprehensive testing and validation. This report describes work done as part of DE-NE0008268 from 10/01/2014 to 11/30/2017 entitled, “Extraction of Uranium from Seawater: Design and Testing of a Symbiotic System.” This effort is part of the Seawater Uranium Recovery Program. This report details the publications and presentations to date on the project, an introduction to the project’s goals and background research into previous work done to achieve these goals thus far. From there, the report describes an algorithm developed during the project used to optimize the adsorption of uranium by changing mechanical parameters such as immersion time and adsorbent reuses is described. Next, a design tool developed as part of the project to determine the global feasibility of symbiotic uranium harvesting systems. Additionally, the report details work done on shell enclosures for uranium adsorption. Moving on, the results from the design, building, and testing of a 1/50th physical scale prototype of a highly feasible symbiotic uranium harvester is described. Then, the report describes the results from flume experiment used to determine the affect of enclosure shells on the uptake of uranium by the adsorbent they enclose. From there the report details the design of a Symbiotic Machine for Ocean uRanium Extraction (SMORE). Next, the results of the 1/10th scale physical scale prototype of a highly feasible symbiotic uranium harvester are presented. The report then details the design and results of an experiment to examine the hydrodynamic effects of a uranium harvester on the offshore wind turbine it is attached to using a 1/150th Froude scale tow tank test. Finally, the report details the results of an initial cost-analysis for the production of uranium from seawater from such a symbiotic device.« less

URANIUM DECONTAMINATION WITH RESPECT TO ZIRCONIUM

DOEpatents

Vogler, S.; Beederman, M.

1961-05-01

A process is given for separating uranium values from a nitric acid aqueous solution containing uranyl values, zirconium values and tetravalent plutonium values. The process comprises contacting said solution with a substantially water-immiscible liquid organic solvent containing alkyl phosphate, separating an organic extract phase containing the uranium, zirconium, and tetravalent plutonium values from an aqueous raffinate, contacting said organic extract phase with an aqueous solution 2M to 7M in nitric acid and also containing an oxalate ion-containing substance, and separating a uranium- containing organic raffinate from aqueous zirconium- and plutonium-containing extract phase.

FLUORINATION PROCESS

DOEpatents

McMillan, T.S.

1957-10-29

A process for the fluorination of uranium metal is described. It is known that uranium will react with liquid chlorine trifluoride but the reaction proceeds at a slow rate. However, a mixture of a halogen trifluoride together with hydrogen fluoride reacts with uranium at a significantly faster rate than does a halogen trifluoride alone. Bromine trifluoride is suitable for use in the process, but chlorine trifluoride is preferred. Particularly suitable is a mixture of ClF/sub 3/ and HF having a mole ratio (moles

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.

PROCESS OF PURIFYING URANIUM

DOEpatents

Seaborg, G.T.; Orlemann, E.F.; Jensen, L.H.

1958-12-23

A method of obtaining substantially pure uranium from a uranium composition contaminated with light element impurities such as sodium, magnesium, beryllium, and the like is described. An acidic aqueous solution containing tetravalent uranium is treated with a soluble molybdate to form insoluble uranous molybdate which is removed. This material after washing is dissolved in concentrated nitric acid to obtaln a uranyl nitrate solution from which highly purified uranium is obtained by extraction with ether.

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.

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.

Influence of uncertainties of isotopic composition of the reprocessed uranium on effectiveness of its enrichment in gas centrifuge cascades

NASA Astrophysics Data System (ADS)

Smirnov, A. Yu; Mustafin, A. R.; Nevinitsa, V. A.; Sulaberidze, G. A.; Dudnikov, A. A.; Gusev, V. E.

2017-01-01

The effect of the uncertainties of the isotopic composition of the reprocessed uranium on its enrichment process in gas centrifuge cascades while diluting it by adding low-enriched uranium (LEU) and waste uranium. It is shown that changing the content of 232U and 236U isotopes in the initial reprocessed uranium within 15% (rel.) can significantly change natural uranium consumption and separative work (up to 2-3%). However, even in case of increase of these parameters is possible to find the ratio of diluents, where the cascade with three feed flows (depleted uranium, LEU and reprocessed uranium) will be more effective than ordinary separation cascade with one feed point for producing LEU from natural uranium.

Investigations for the Recycle of Pyroprocessed Uranium

NASA Astrophysics Data System (ADS)

Westphal, B. R.; Price, J. C.; Chambers, E. E.; Patterson, M. N.

Given the renewed interest in uranium from the pyroprocessing of used nuclear fuel in a molten salt system, the two biggest hurdles for marketing the uranium are radiation levels and transuranic content. A radiation level as low as possible is desired so that handling operations can be performed directly with the uranium. The transuranic content of the uranium will affect the subsequent waste streams generated and, thus also should be minimized. Although the pyroprocessing technology was originally developed without regard to radiation and transuranic levels, adaptations to the process have been considered. Process conditions have been varied during the distillation and casting cycles of the process with increasing temperature showing the largest effect on the reduction of radiation levels. Transuranic levels can be reduced significantly by incorporating a pre-step in the salt distillation operation to remove a majority of the salt prior to distillation.

Preparation of uranium fuel kernels with silicon carbide nanoparticles using the internal gelation process

NASA Astrophysics Data System (ADS)

Hunt, R. D.; Silva, G. W. C. M.; Lindemer, T. B.; Anderson, K. K.; Collins, J. L.

2012-08-01

The US Department of Energy continues to use the internal gelation process in its preparation of tristructural isotropic coated fuel particles. The focus of this work is to develop uranium fuel kernels with adequately dispersed silicon carbide (SiC) nanoparticles, high crush strengths, uniform particle diameter, and good sphericity. During irradiation to high burnup, the SiC in the uranium kernels will serve as getters for excess oxygen and help control the oxygen potential in order to minimize the potential for kernel migration. The hardness of SiC required modifications to the gelation system that was used to make uranium kernels. Suitable processing conditions and potential equipment changes were identified so that the SiC could be homogeneously dispersed in gel spheres. Finally, dilute hydrogen rather than argon should be used to sinter the uranium kernels with SiC.

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

Chambers, Doug; Wiatzka, Gerd; Brown, Steve

This paper provides the life story of Canada's original radium/uranium mine. In addition to the history of operations, it discusses the unique and successful approach used to identify the key issues and concerns associated with the former radium, uranium and silver mining property and the activities undertaken to define the remedial actions and subsequent remedial plan. The Port Radium Mine site, situated approximately 275 km north of Yellowknife on the east shore of Great Bear Lake, Northwest Territories, was discovered in 1930 and underground mining began in 1932. The mine operated almost continuously from 1932 to 1982, initially for recoverymore » of radium, then uranium and finally, for recovery of silver. Tailings production totaled an estimated 900,000 tons and 800,000 tons from uranium and silver processing operations respectively. In the early days of mining, Port Radium miners were exposed to radon and associated decay product levels (in Working Level Months of exposure - WLM) hundreds of times greater than modern standards. The experience of the Port Radium miners provides important contribution to understanding the risks from radon. While the uranium mine was originally decommissioned in the early 1960's, to the standards of the day, the community of Deline (formerly Fort Franklin) had concerns about residual contamination at the mine site and the potential effects arising from use of traditional lands. The Deline people were also concerned about the possible risks to Deline Dene arising from their work as ore carriers. In the late 1990's, the community of Deline brought these concerns to national attention and consequently, the Government of Canada and the community of Deline agreed to move forward in a collaborative manner to address these concerns. The approach agreed to was to establish the Canada-Deline Uranium Table (CDUT) to provide a joint process by which the people of Deline could have their concerns expressed and addressed. A great deal of work was done through the CDUT, including efforts to assess site environment and safety issues in the context of modern reclamation standards. In addition to the environmental and remediation studies, an assessment of historic exposures of Deline ore carriers to radiation and a follow-up epidemiological feasibility study were performed. SENES Consultants Limited (SENES) carried out the dose reconstruction for the Port Radium miners in the 1990's, was the environmental consultant to the CDUT from 2000 to 2005, developed the Remedial Action Plan (RAP), engineering plans and specifications for decommissioning the Port Radium mine and vicinity sites in 2005/6, supervised the remedial works in 2007 and carried out the long term post closure monitoring from 2008 to 2012. Our firsthand experience from working cooperatively with the CDUT provides insights into effective decommissioning of historic contaminated sites. (authors)« less

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

Microbiological, Geochemical and Hydrologic Processes Controlling Uranium Mobility: An Integrated Field-Scale Subsurface Research Challenge Site at Rifle, Colorado, Quality Assurance Project Plan

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

Fix, N. J.

The U.S. Department of Energy (DOE) is cleaning up and/or monitoring large, dilute plumes contaminated by metals, such as uranium and chromium, whose mobility and solubility change with redox status. Field-scale experiments with acetate as the electron donor have stimulated metal-reducing bacteria to effectively remove uranium [U(VI)] from groundwater at the Uranium Mill Tailings Site in Rifle, Colorado. The Pacific Northwest National Laboratory and a multidisciplinary team of national laboratory and academic collaborators has embarked on a research proposed for the Rifle site, the object of which is to gain a comprehensive and mechanistic understanding of the microbial factors andmore » associated geochemistry controlling uranium mobility so that DOE can confidently remediate uranium plumes as well as support stewardship of uranium-contaminated sites. This Quality Assurance Project Plan provides the quality assurance requirements and processes that will be followed by the Rifle Integrated Field-Scale Subsurface Research Challenge Project.« less

Elution of Uranium and Transition Metals from Amidoxime-Based Polymer Adsorbents for Sequestering Uranium from Seawater

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

Pan, Horng-Bin; Kuo, Li-Jung; Wai, Chien M.

2015-11-30

High-surface-area amidoxime and carboxylic acid grafted polymer adsorbents developed at Oak Ridge National Laboratory were tested for sequestering uranium in a flowing seawater flume system at the PNNL-Marine Sciences Laboratory. FTIR spectra indicate that a KOH conditioning process is necessary to remove the proton from the carboxylic acid and make the sorbent effective for sequestering uranium from seawater. The alkaline conditioning process also converts the amidoxime groups to carboxylate groups in the adsorbent. Both Na 2CO 3-H 2O 2 and hydrochloric acid elution methods can remove ~95% of the uranium sequestered by the adsorbent after 42 days of exposure inmore » real seawater. The Na 2CO 3-H 2O 2 elution method is more selective for uranium than conventional acid elution. Iron and vanadium are the two major transition metals competing with uranium for adsorption to the amidoxime-based adsorbents in real seawater.« less

Hydrologic and Temporal Influences of Evaporite Minerals on the Vertical Distribution, Storage, and Mobility of Uranium

NASA Astrophysics Data System (ADS)

Roycroft, S. J.; Noel, V.; Boye, K.; Besancon, C.; Weaver, K. L.; Johnson, R. H.; Dam, W. L.; Fendorf, S. E.; Bargar, J.

2016-12-01

Uranium contaminated groundwater in Riverton, Wyoming persists despite anticipated natural attenuation outside of a former uranium ore processing facility. The inability of natural flushing to dilute the uranium below the regulatory threshold indicates that sediments act as secondary sources likely (re)supplying uranium to groundwater. Throughout the contaminated floodplain, uranium rich-evaporites are readily abundant in the upper 2 m of sediments and are spatially coincident with the location of the plume, which suggests a likely link between evaporites and increased uranium levels. Knowledge of where and how uranium is stored within evaporite-associated sediments is required to understand processes controlling the mobility of uranium. We expect that flooding and seasonal changes in hydrologic conditions will affect U phase partitioning, and thus largely control U mobility. The primary questions we are addressing in this project are: What is the relative abundance of uranium incorporated in various mineral complexes throughout the evaporite sediments? How do the factors of depth, location, and seasonality influence the relative incorporation, mobility and speciation of uranium?We have systematically sampled from two soil columns over three dates in Riverton. The sampling dates span before and after a significant flooding event, providing insight into the flood's impact on local uranium mobility. Sequential chemical extractions are used to decipher the reactivity of uranium and approximate U operationally defined within reactants targeting carbonate, silicate, organic, and metal oxide bound or water and exchangeable phases. Extractions throughout the entirety of the sediment cores provide a high-resolution vertical profile of the distribution of uranium in various extracted phases. Throughout the profile, the majority (50-60%) of uranium is bound within carbonate-targeted extracts, a direct effect of the carbonate-rich evaporite sediments. The sum of our analyses provide a dynamic model of uranium incorporation within evaporite sediments holding implications for the fate of uranium throughout contaminated sites across the Colorado River Basin.

Solubility testing of actinides on breathing-zone and area air samples

NASA Astrophysics Data System (ADS)

Metzger, Robert Lawrence

The solubility of inhaled radionuclides in the human lung is an important characteristic of the compounds needed to perform internal dosimetry assessments for exposed workers. A solubility testing method for uranium and several common actinides has been developed with sufficient sensitivity to allow profiles to be determined from routine breathing zone and area air samples in the workplace. Air samples are covered with a clean filter to form a filter-sample-filter sandwich which is immersed in an extracellular lung serum simulant solution. The sample is moved to a fresh beaker of the lung fluid simulant each day for one week, and then weekly until the end of the 28 day test period. The soak solutions are wet ashed with nitric acid and hydrogen peroxide to destroy the organic components of the lung simulant solution prior to extraction of the nuclides of interest directly into an extractive scintillator for subsequent counting on a Photon-Electron Rejecting Alpha Liquid Scintillation (PERALSsp°ler ) spectrometer. Solvent extraction methods utilizing the extractive scintillators have been developed for the isotopes of uranium, plutonium, and curium. The procedures normally produce an isotopic recovery greater than 95% and have been used to develop solubility profiles from air samples with 40 pCi or less of Usb3Osb8. This makes it possible to characterize solubility profiles in every section of operating facilities where airborne nuclides are found using common breathing zone air samples. The new method was evaluated by analyzing uranium compounds from two uranium mills whose product had been previously analyzed by in vitro solubility testing in the laboratory and in vivo solubility testing in rodents. The new technique compared well with the in vivo rodent solubility profiles. The method was then used to evaluate the solubility profiles in all process sections of an operating in situ uranium plant using breathing zone and area air samples collected during routine plant operations. The solubility profiles developed from this work showed excellent agreement with the results of the worker urine bioassay program at the plant and identified a significant error in existing internal dose assessments at this facility.

The inhalation of radioactive materials as related to hand contamination

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

Bailey, J.C.; Rohr, R.C.

1953-09-15

Tests performed to determine the hazard associated with the inhalation of radioactive materials as the result of smoking with contaminated hands indicate that for dry uranium compounds adhering to the palmar surfaces of the hands, approximately 1.0% of the material may be transferred to a cigarette, and that of this approximately 0.2% may appear in the smoke which is inhaled. Most of the contamination originally placed in a cigarette was found in the ash, and only 11% of the material was not recovered following burning; approximately half of this loss may be attributed to normal losses inherent in the analyticalmore » process, the recovery efficiency for which was found by supplementary experiments to be 95%.« less

ALKALINE CARBONATE LEACHING PROCESS FOR URANIUM EXTRACTION

DOEpatents

Thunaes, A.; Brown, E.A.; Rabbitts, A.T.

1957-11-12

A process for the leaching of uranium from high carbonate ores is presented. According to the process, the ore is leached at a temperature of about 200 deg C and a pressure of about 200 p.s.i.g. with a solution containing alkali carbonate, alkali permanganate, and bicarbonate ion, the bicarbonate ion functionlng to prevent premature formation of alkali hydroxide and consequent precipitation of a diuranate. After the leaching is complete, the uranium present is recovered by precipitation with NaOH.

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)

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.

DIRECT INGOT PROCESS FOR PRODUCING URANIUM

DOEpatents

Leaders, W.M.; Knecht, W.S.

1960-11-15

A process is given in which uranium tetrafluoride is reduced to the metal with magnesium and in the same step the uranium metal formed is cast into an ingot. For this purpose a mold is arranged under and connected with the reaction bomb, and both are filled with the reaction mixture. The entire mixture is first heated to just below reaction temperature, and thereafter heating is restricted to the mixture in the mold. The reaction starts in the mold whereby heat is released which brings the rest of the mixture to reaction temperature. Pure uranium metal settles in the mold while the magnesium fluoride slag floats on top of it. After cooling, the uranium is separated from the slag by mechanical means.

Sorption testing and generalized composite surface complexation models for determining uranium sorption parameters at a proposed in-situ recovery site

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

Johnson, Raymond H.; Truax, Ryan A.; Lankford, David A.

Solid-phase iron concentrations and generalized composite surface complexation models were used to evaluate procedures in determining uranium sorption on oxidized aquifer material at a proposed U in situ recovery (ISR) site. At the proposed Dewey Burdock ISR site in South Dakota, USA, oxidized aquifer material occurs downgradient of the U ore zones. Solid-phase Fe concentrations did not explain our batch sorption test results,though total extracted Fe appeared to be positively correlated with overall measured U sorption. Batch sorption test results were used to develop generalized composite surface complexation models that incorporated the full genericsorption potential of each sample, without detailedmore » mineralogiccharacterization. The resultant models provide U sorption parameters (site densities and equilibrium constants) for reactive transport modeling. The generalized composite surface complexation sorption models were calibrated to batch sorption data from three oxidized core samples using inverse modeling, and gave larger sorption parameters than just U sorption on the measured solidphase Fe. These larger sorption parameters can significantly influence reactive transport modeling, potentially increasing U attenuation. Because of the limited number of calibration points, inverse modeling required the reduction of estimated parameters by fixing two parameters. The best-fit models used fixed values for equilibrium constants, with the sorption site densities being estimated by the inversion process. While these inverse routines did provide best-fit sorption parameters, local minima and correlated parameters might require further evaluation. Despite our limited number of proxy samples, the procedures presented provide a valuable methodology to consider for sites where metal sorption parameters are required. Furthermore, these sorption parameters can be used in reactive transport modeling to assess downgradient metal attenuation, especially when no other calibration data are available, such as at proposed U ISR sites.« less

Sorption testing and generalized composite surface complexation models for determining uranium sorption parameters at a proposed in-situ recovery site

DOE PAGES

Johnson, Raymond H.; Truax, Ryan A.; Lankford, David A.; ...

2016-02-03

Solid-phase iron concentrations and generalized composite surface complexation models were used to evaluate procedures in determining uranium sorption on oxidized aquifer material at a proposed U in situ recovery (ISR) site. At the proposed Dewey Burdock ISR site in South Dakota, USA, oxidized aquifer material occurs downgradient of the U ore zones. Solid-phase Fe concentrations did not explain our batch sorption test results,though total extracted Fe appeared to be positively correlated with overall measured U sorption. Batch sorption test results were used to develop generalized composite surface complexation models that incorporated the full genericsorption potential of each sample, without detailedmore » mineralogiccharacterization. The resultant models provide U sorption parameters (site densities and equilibrium constants) for reactive transport modeling. The generalized composite surface complexation sorption models were calibrated to batch sorption data from three oxidized core samples using inverse modeling, and gave larger sorption parameters than just U sorption on the measured solidphase Fe. These larger sorption parameters can significantly influence reactive transport modeling, potentially increasing U attenuation. Because of the limited number of calibration points, inverse modeling required the reduction of estimated parameters by fixing two parameters. The best-fit models used fixed values for equilibrium constants, with the sorption site densities being estimated by the inversion process. While these inverse routines did provide best-fit sorption parameters, local minima and correlated parameters might require further evaluation. Despite our limited number of proxy samples, the procedures presented provide a valuable methodology to consider for sites where metal sorption parameters are required. Furthermore, these sorption parameters can be used in reactive transport modeling to assess downgradient metal attenuation, especially when no other calibration data are available, such as at proposed U ISR sites.« less

ANALYTICAL METHOD FOR THE DETERMINATION OF BORON IN URANYL NITRATE SOLUTIONS

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

None

1962-01-01

A method was developed for the determination of boron in uranyl nitrate solutions. The boron is separated from uranium and other impurities by distillation of methyl borate. It is determined absorptiometrically by means of curcumin in the presence of orthochlorophenol, perchloric acid, and acetic anhydride. The limit of detection is judged to be not greater than 0.05 mu g, but is dependent on the purity of the reagents used. The coefficient of variation on 210 results at the 0.2 mu g boron level was 26% with a bias of -25%. The method may be applied to depleted uranyl nitrate solutionsmore » and uranium slag recovery liquors. (auth)« less

A phase-field simulation of uranium dendrite growth on the cathode in the electrorefining process

NASA Astrophysics Data System (ADS)

Shibuta, Yasushi; Unoura, Seiji; Sato, Takumi; Shibata, Hiroki; Kurata, Masaki; Suzuki, Toshio

2011-07-01

The uranium dendrite growth on the cathode during the pyroprocessing of uranium is investigated using a novel phase-field model, in which electrodeposition of uranium and zirconium from the molten-salt is taken into account. The threshold concentration of zirconium in the molten salt demarcating the dendritic and planar growth is then estimated as a function of the current density. Moreover, the growth process of both the dendritic and planar electrodeposits has been demonstrated by way of varying the mobility of the phase field, which consists of the effect of attachment kinetics and diffusion.

Biogeochemical Processes Regulating the Mobility of Uranium in Sediments

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

Belli, Keaton M.; Taillefert, Martial

This book chapters reviews the latest knowledge on the biogeochemical processes regulating the mobility of uranium in sediments. It contains both data from the literature and new data from the authors.

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

URANIUM EXTRACTION PROCESS

DOEpatents

Baldwin, W.H.; Higgins, C.E.

1958-12-16

A process is described for recovering uranium values from acidic aqueous solutions containing hexavalent uranium by contacting the solution with an organic solution comprised of a substantially water-immiscible organlc diluent and an organic phosphate to extract the uranlum values into the organic phase. Carbon tetrachloride and a petroleum hydrocarbon fraction, such as kerosene, are sultable diluents to be used in combination with organlc phosphates such as dibutyl butylphosphonate, trlbutyl phosphine oxide, and tributyl phosphate.

Visualizing different uranium oxidation states during the surface alteration of uraninite and uranium tetrachloride.

PubMed

Grossmann, Kay; Arnold, Thuro; Steudtner, Robin; Weiss, Stefan; Bernhard, Gert

2009-08-01

Low-temperature alteration reactions on uranium phases may lead to the mobilization of uranium and thereby poses a potential threat to humans living close to uranium-contaminated sites. In this study, the surface alteration of uraninite (UO(2)) and uranium tetrachloride (UCl(4)) in air atmosphere was studied by confocal laser scanning microscopy (CLSM) and laser-induced fluorescence spectroscopy using an excitation wavelength of 408 nm. It was found that within minutes the oxidation state on the surface of the uraninite and the uranium tetrachloride changed. During the surface alteration process U(IV) atoms on the uraninite and uranium tetrachloride surface became stepwise oxidized by a one-electron step at first to U(V) and then further to U(VI). These observed changes in the oxidation states of the uraninite surface were microscopically visualized and spectroscopically identified on the basis of their fluorescence emission signal. A fluorescence signal in the wavelength range of 415-475 nm was indicative for metastable uranium(V), and a fluorescence signal in the range of 480-560 nm was identified as uranium(VI). In addition, the oxidation process of tetravalent uranium in aqueous solution at pH 0.3 was visualized by CLSM and U(V) was fluorescence spectroscopically identified. The combination of microscopy and fluorescence spectroscopy provided a very convincing visualization of the brief presence of U(V) as a metastable reaction intermediate and of the simultaneous coexistence of the three states U(IV), U(V), and U(VI). These results have a significant importance for fundamental uranium redox chemistry and should contribute to a better understanding of the geochemical behavior of uranium in nature.

Anaerobic bioremediation of hexavalent uranium in groundwater by reductive precipitation with methanogenic granular sludge.

PubMed

Tapia-Rodriguez, Aida; Luna-Velasco, Antonia; Field, Jim A; Sierra-Alvarez, Reyes

2010-04-01

Uranium has been responsible for extensive contamination of groundwater due to releases from mill tailings and other uranium processing waste. Past evidence has confirmed that certain bacteria can enzymatically reduce soluble hexavalent uranium (U(VI)) to insoluble tetravalent uranium (U(IV)) under anaerobic conditions in the presence of appropriate electron donors. This paper focuses on the evaluation of anaerobic granular sludge as a source of inoculum for the bioremediation of uranium in water. Batch experiments were performed with several methanogenic anaerobic granular sludge samples and different electron donors. Abiotic controls consisting of heat-killed inoculum and non-inoculated treatments confirmed the biological removal process. In this study, unadapted anaerobic granular sludge immediately reduced U(VI), suggesting an intrinsic capacity of the sludge to support this process. The high biodiversity of anaerobic granular sludge most likely accounts for the presence of specific microorganisms capable of reducing U(VI). Oxidation by O(2) was shown to resolubilize the uranium. This observation combined with X-ray diffraction evidence of uraninite confirmed that the removal during anaerobic treatment was due to reductive precipitation. The anaerobic removal activity could be sustained after several respikes of U(VI). The U(VI) removal was feasible without addition of electron donors, indicating that the decay of endogenous biomass substrates was contributing electron equivalents to the process. Addition of electron donors, such as H(2) stimulated the removal of U(VI) to varying degrees. The stimulation was greater in sludge samples with lower endogenous substrate levels. The present work reveals the potential application of anaerobic granular sludge for continuous bioremediation schemes to treat uranium-contaminated water. Copyright (c) 2009 Elsevier Ltd. All rights reserved.

Determination of trace uranium by resonance fluorescence method coupled with photo-catalytic technology and dual cloud point extraction

NASA Astrophysics Data System (ADS)

Li, Jiekang; Li, Guirong; Han, Qian

2016-12-01

In this paper, two kinds of salophens (Sal) with different solubilities, Sal1 and Sal2, have been respectively synthesized, and they all can combine with uranyl to form stable complexes: [UO22 +-Sal1] and [UO22 +-Sal2]. Among them, [UO22 +-Sal1] was used as ligand to extract uranium in complex samples by dual cloud point extraction (dCPE), and [UO22 +-Sal2] was used as catalyst for the determination of uranium by photocatalytic resonance fluorescence (RF) method. The photocatalytic characteristic of [UO22 +-Sal2] on the oxidized pyronine Y (PRY) by potassium bromate which leads to the decrease of RF intensity of PRY were studied. The reduced value of RF intensity of reaction system (ΔF) is in proportional to the concentration of uranium (c), and a novel photo-catalytic RF method was developed for the determination of trace uranium (VI) after dCPE. The combination of photo-catalytic RF techniques and dCPE procedure endows the presented methods with enhanced sensitivity and selectivity. Under optimal conditions, the linear calibration curves range for 0.067 to 6.57 ng mL- 1, the linear regression equation was ΔF = 438.0 c (ng mL- 1) + 175.6 with the correlation coefficient r = 0.9981. The limit of detection was 0.066 ng mL- 1. The proposed method was successfully applied for the separation and determination of uranium in real samples with the recoveries of 95.0-103.5%. The mechanisms of the indicator reaction and dCPE are discussed.

Removal of toxic uranium from synthetic nuclear power reactor effluents using uranyl ion imprinted polymer particles.

PubMed

Preetha, Chandrika Ravindran; Gladis, Joseph Mary; Rao, Talasila Prasada; Venkateswaran, Gopala

2006-05-01

Major quantities of uranium find use as nuclear fuel in nuclear power reactors. In view of the extreme toxicity of uranium and consequent stringent limits fixed by WHO and various national governments, it is essential to remove uranium from nuclear power reactor effluents before discharge into environment. Ion imprinted polymer (IIP) materials have traditionally been used for the recovery of uranium from dilute aqueous solutions prior to detection or from seawater. We now describe the use of IIP materials for selective removal of uranium from a typical synthetic nuclear power reactor effluent. The IIP materials were prepared for uranyl ion (imprint ion) by forming binary salicylaldoxime (SALO) or 4-vinylpyridine (VP) or ternary SALO-VP complexes in 2-methoxyethanol (porogen) and copolymerizing in the presence of styrene (monomer), divinylbenzene (cross-linking monomer), and 2,2'-azobisisobutyronitrile (initiator). The resulting materials were then ground and sieved to obtain unleached polymer particles. Leached IIP particles were obtained by leaching the imprint ions with 6.0 M HCl. Control polymer particles were also prepared analogously without the imprint ion. The IIP particles obtained with ternary complex alone gave quantitative removal of uranyl ion in the pH range 3.5-5.0 with as low as 0.08 g. The retention capacity of uranyl IIP particles was found to be 98.50 mg/g of polymer. The present study successfully demonstrates the feasibility of removing uranyl ions selectively in the range 5 microg - 300 mg present in 500 mL of synthetic nuclear power reactor effluent containing a host of other inorganic species.

Interlaboratory comparison program for nondestructive assay of prototype uranium reference materials

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

Trahey, N.M.; Smith, M.M.; Voeks, A.M.

The US Department of Energy (DOE), New Brunswick Laboratory (NBS), designed and administered an interlaboratory comparison program based on the measurement of NBL-produced prototype uranium nondestructive assay (NDA) reference materials for scrap and waste. The objectives of the program were to evaluate the reliability of NDA techniques as applied to nuclear safeguards materials control and accountability needs and to investigate the feasibility of providing practical NDA scrap and waste reference materials for use throughout the nuclear safeguards community. Fourteen facilities representing seven DOE contractors, four US Nuclear Regulatory Commission (NRC) licensees, one EURATOM Laboratory, and NBL, participated in this program.more » Three stable, well-characterized uranium reference materials were developed and certified for this program. Synthetic calcined ash, cellulose fiber, and ion-exchange resin simulate selected uranium scrap and waste forms which are often encountered in fabrication and recovery operations. The synthetic calcined ash represents an intermediate density inorganic matrix while the cellulose fiber and ion-exchange resin are representative of low-density organic matrices. The materials, containing from 0 to 13% uranium enriched at 93% /sup 235/U, were sealed in specially selected containers. Nineteen prototype reference samples, plus three empty containers, one to accompany each set, was circulated to the participants between August 1979 and May 1984. Triplicate measurements for /sup 235/U on each of the 19 filled containers were required. In addition, participants could opt to perform modular configuration measurements using containers from Sets IIA and IIB to simulate non-homogeneously dispersed uranium in waste containers. All data were reported to NBL for evaluation.« less

303-K Storage Facility closure plan. Revision 2

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

Not Available

1993-12-15

Recyclable scrap uranium with zircaloy-2 and copper silicon alloy, uranium-titanium alloy, beryllium/zircaloy-2 alloy, and zircaloy-2 chips and fines were secured in concrete billets (7.5-gallon containers) in the 303-K Storage Facility, located in the 300 Area. The beryllium/zircaloy-2 alloy and zircaloy-2 chips and fines are designated as mixed waste with the characteristic of ignitability. The concretion process reduced the ignitability of the fines and chips for safe storage and shipment. This process has been discontinued and the 303-K Storage Facility is now undergoing closure as defined in the Resource Conservation and Recovery Act (RCRA) of 1976 and the Washington Administrative Codemore » (WAC) Dangerous Waste Regulations, WAC 173-303-040. This closure plan presents a description of the 303-K Storage Facility, the history of materials and waste managed, and the procedures that will be followed to close the 303-K Storage Facility. The 303-K Storage Facility is located within the 300-FF-3 (source) and 300-FF-5 (groundwater) operable units, as designated in the Hanford Federal Facility Agreement and Consent Order (Tri-Party Agreement) (Ecology et al. 1992). Contamination in the operable units 300-FF-3 and 300-FF-5 is scheduled to be addressed through the Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) of 1980 remedial action process. Therefore, all soil remedial action at the 304 Facility will be conducted as part of the CERCLA remedial action of operable units 300-FF-3 and 300-FF-5.« less

Chemical Technology Division annual technical report, 1990

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

Not Available

1991-05-01

Highlights of the Chemical Technology (CMT) Division's activities during 1990 are presented. In this period, CMT conducted research and development in the following areas: (1) electrochemical technology, including advanced batteries and fuel cells; (2) technology for coal- fired magnetohydrodynamics and fluidized-bed combustion; (3) methods for recovery of energy from municipal waste and techniques for treatment of hazardous organic waste; (4) the reaction of nuclear waste glass and spent fuel under conditions expected for a high-level waste repository; (5) processes for separating and recovering transuranic elements from nuclear waste streams, concentrating plutonium solids in pyrochemical residues by aqueous biphase extraction, andmore » treating natural and process waters contaminated by volatile organic compounds; (6) recovery processes for discharged fuel and the uranium blanket in the Integral Fast Reactor (IFR); (7) processes for removal of actinides in spent fuel from commercial water-cooled nuclear reactors and burnup in IFRs; and (8) physical chemistry of selected materials in environments simulating those of fission and fusion energy systems. The Division also has a program in basic chemistry research in the areas of fluid catalysis for converting small molecules to desired products; materials chemistry for superconducting oxides and associated and ordered solutions at high temperatures; interfacial processes of importance to corrosion science, high-temperature superconductivity, and catalysis; and the geochemical processes responsible for trace-element migration within the earth's crust. The Analytical Chemistry Laboratory in CMT provides a broad range of analytical chemistry support services to the scientific and engineering programs at Argonne National Laboratory (ANL). 66 refs., 69 figs., 6 tabs.« less

40 CFR 192.00 - Applicability.

Code of Federal Regulations, 2012 CFR

2012-07-01

... AND ENVIRONMENTAL PROTECTION STANDARDS FOR URANIUM AND THORIUM MILL TAILINGS Standards for the Control of Residual Radioactive Materials from Inactive Uranium Processing Sites § 192.00 Applicability. This... sites under section 108 of the Uranium Mill Tailings Radiation Control Act of 1978 (henceforth...

40 CFR 192.10 - Applicability.

Code of Federal Regulations, 2013 CFR

2013-07-01

... AND ENVIRONMENTAL PROTECTION STANDARDS FOR URANIUM AND THORIUM MILL TAILINGS Standards for Cleanup of Land and Buildings Contaminated with Residual Radioactive Materials from Inactive Uranium Processing... radioactive materials at which all or substantially all of the uranium was produced for sale to any Federal...

40 CFR 192.00 - Applicability.

Code of Federal Regulations, 2011 CFR

2011-07-01

... AND ENVIRONMENTAL PROTECTION STANDARDS FOR URANIUM AND THORIUM MILL TAILINGS Standards for the Control of Residual Radioactive Materials from Inactive Uranium Processing Sites § 192.00 Applicability. This... sites under section 108 of the Uranium Mill Tailings Radiation Control Act of 1978 (henceforth...

40 CFR 192.10 - Applicability.

Code of Federal Regulations, 2014 CFR

2014-07-01

... AND ENVIRONMENTAL PROTECTION STANDARDS FOR URANIUM AND THORIUM MILL TAILINGS Standards for Cleanup of Land and Buildings Contaminated with Residual Radioactive Materials from Inactive Uranium Processing... radioactive materials at which all or substantially all of the uranium was produced for sale to any Federal...

40 CFR 192.00 - Applicability.

Code of Federal Regulations, 2013 CFR

2013-07-01

... AND ENVIRONMENTAL PROTECTION STANDARDS FOR URANIUM AND THORIUM MILL TAILINGS Standards for the Control of Residual Radioactive Materials from Inactive Uranium Processing Sites § 192.00 Applicability. This... sites under section 108 of the Uranium Mill Tailings Radiation Control Act of 1978 (henceforth...

40 CFR 192.00 - Applicability.

Code of Federal Regulations, 2010 CFR

2010-07-01

... AND ENVIRONMENTAL PROTECTION STANDARDS FOR URANIUM AND THORIUM MILL TAILINGS Standards for the Control of Residual Radioactive Materials from Inactive Uranium Processing Sites § 192.00 Applicability. This... sites under section 108 of the Uranium Mill Tailings Radiation Control Act of 1978 (henceforth...

Process for reducing beta activity in uranium

DOEpatents

Briggs, Gifford G.; Kato, Takeo R.; Schonegg, Edward

1986-10-07

This invention is a method for lowering the beta radiation hazards associated with the casting of uranium. The method reduces the beta radiation emitted from the as-cast surfaces of uranium ingots. The method also reduces the amount of beta radiation emitters retained on the interiors of the crucibles that have been used to melt the uranium charges and which have undergone cleaning in a remote handling facility. The lowering of the radioactivity is done by scavenging the beta emitters from the molten uranium with a molten mixture containing the fluorides of magnesium and calcium. The method provides a means of collection and disposal of the beta emitters in a manner that reduces radiation exposure to operating personnel in the work area where the ingots are cast and processed.

Process for reducing beta activity in uranium

DOEpatents

Briggs, Gifford G.; Kato, Takeo R.; Schonegg, Edward

1986-01-01

This invention is a method for lowering the beta radiation hazards associated with the casting of uranium. The method reduces the beta radiation emitted from the as-cast surfaces of uranium ingots. The method also reduces the amount of beta radiation emitters retained on the interiors of the crucibles that have been used to melt the uranium charges and which have undergone cleaning in a remote handling facility. The lowering of the radioactivity is done by scavenging the beta emitters from the molten uranium with a molten mixture containing the fluorides of magnesium and calcium. The method provides a means of collection and disposal of the beta emitters in a manner that reduces radiation exposure to operating personnel in the work area where the ingots are cast and processed.

Process for reducing beta activity in uranium

DOEpatents

Briggs, G.G.; Kato, T.R.; Schonegg, E.

1985-04-11

This invention is a method for lowering the beta radiation hazards associated with the casting of uranium. The method reduces the beta radiation emitted from the as-cast surfaces of uranium ingots. The method also reduces the amount of beta radiation emitters retained on the interiors of the crucibles that have been used to melt the uranium charges and which undergone cleaning in a remote handling facility. The lowering of the radioactivity is done by scavenging the beta emitters from the molten uranium with a molten mixture containing the fluorides of magnesium and calcium. The method provides a means of collection and disposal of the beta emitters in a manner that reduces radiation exposure to operating personnel in the work area where the ingots are cast and processed. 5 tabs.

High loading uranium fuel plate

DOEpatents

Wiencek, Thomas C.; Domagala, Robert F.; Thresh, Henry R.

1990-01-01

Two embodiments of a high uranium fuel plate are disclosed which contain a meat comprising structured uranium compound confined between a pair of diffusion bonded ductile metal cladding plates uniformly covering the meat, the meat having a uniform high fuel loading comprising a content of uranium compound greater than about 45 Vol. % at a porosity not greater than about 10 Vol. %. In a first embodiment, the meat is a plurality of parallel wires of uranium compound. In a second embodiment, the meat is a dispersion compact containing uranium compound. The fuel plates are fabricated by a hot isostatic pressing process.

URANIUM SEPARATION PROCESS

DOEpatents

Hyde, E.K.; Katzin, L.I.; Wolf, M.J.

1959-07-14

The separation of uranium from a mixture of uranium and thorium by organic solvent extraction from an aqueous solution is described. The uranium is separrted from an aqueous mixture of uranium and thorium nitrates 3 N in nitric acid and containing salting out agents such as ammonium nitrate, so as to bring ihe total nitrate ion concentration to a maximum of about 8 N by contacting the mixture with an immiscible aliphatic oxygen containing organic solvent such as diethyl carbinol, hexone, n-amyl acetate and the like. The uranium values may be recovered from the organic phase by back extraction with water.

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.

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.

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

Chemical treatment of low-grade uranium ores. Extraction of uranium from tricalcium phosphate; TRAITEMENT CHIMIQUE DES MINERAIS PAUVRES D'URANIUM. EXTRACTION DE L'URANIUM DU PHOSPHATE TRICALCIQUE (in French)

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

Mechelynck, Ph.

1958-07-15

After an examination of the different processes for the treatment of uranium minerals, it is concluded that the extraction of uranium by ion exchange is not applicable to hydrochloric acid solutions of phosphates. A sulfuric or phosphoric solution can be used. For solvent extraction of uranium, sulfuric or phosphoric solutions are the best, but hydrochloric solutions can be used. The cost of the solvents used would determine the cost of the operation. It is necessary, in the case of liquid-liquid extraction, to filter or decant the solution before extraction. (tr-auth)

PROCESS FOR PRODUCING URANIUM TETRAFLUORIDE

DOEpatents

Harvey, B.G.

1954-09-14

>This patent relates to improvements in the method for producing uranium tetrafluoride by treating an aqueous solutlon of a uranyl salt at an elevated temperature with a reducing agent effective in acld solutlon in the presence of hydrofluoric acid. Uranium tetrafluoride produced this way frequentiy contains impurities in the raw material serving as the source of uranium. Uranium tetrafluoride much less contaminated with impurities than when prepared by the above method can be prepared from materials containing such impurities by first adding a small proportion of reducing agent so as to cause a small fraction, for example 1 to 5% of the uranium tetrafluoride to be precipitated, rejecting such precipitate, and then precipitating and recovering the remainder of the uranium tetrafluoride.

FUEL ELEMENTS FOR NUCLEAR REACTORS AND PROCESS OF MAKING

DOEpatents

Roake, W.E.

1958-08-19

A process is described for producing uranium metal granules for use in reactor fuel elements. The granules are made by suspending powdered uramiunn metal or uranium hydride in a viscous, non-reactive liquid, such as paraffin oil, aad pouring the resulting suspension in droplet, on to a bed of powdered absorbent. In this manner the liquid vehicle is taken up by the sorbent and spherical pellets of uranium metal are obtained. The

PROCESS OF TREATING URANIUM HEXAFLUORIDE AND PLUTONIUM HEXAFLUORIDE MIXTURES WITH SULFUR TETRAFLUORIDE TO SEPARATE SAME

DOEpatents

Steindler, M.J.

1962-07-24

A process was developed for separating uranium hexafluoride from plutonium hexafluoride by the selective reduction of the plutonium hexafluoride to the tetrafluoride with sulfur tetrafluoride at 50 to 120 deg C, cooling the mixture to --60 to -100 deg C, and volatilizing nonreacted sulfur tetrafluoride and sulfur hexafluoride formed at that temperature. The uranium hexafluoride is volatilized at room temperature away from the solid plutonium tetrafluoride. (AEC)

75 FR 34219 - Revision of Fee Schedules; Fee Recovery for FY 2010

Federal Register 2010, 2011, 2012, 2013, 2014

2010-06-16

....8 $6.3 $7.5 Spent Fuel Storage/Reactor Decommissioning..... -- -- 2.7 0.2 0.2 Test and Research... 2009 fee is also shown for comparative purposes. Table V--Rebaselined Annual Fees FY2009 Annual FY 2010... Decommissioning Test and Research Reactors (Non-power 87,600 81,700 Reactors) High Enriched Uranium Fuel Facility...

78 FR 9945 - Crow Butte Resources, Inc.; Establishment of Atomic Safety and Licensing Board

Federal Register 2010, 2011, 2012, 2013, 2014

2013-02-12

... Material License, SUA-1534, to obtain authority to construct and operate an in situ uranium recovery satellite facility at its Marsland site in Dawes County, Nebraska. In response to a ``Notice of Opportunity for a Hearing'' published in the Federal Register, see 77 Fed. Reg. 71,454 (Nov. 30, 2012), hearing...

77 FR 68814 - Notice of Availability of the Draft Environmental Impact Statement for the Gas Hills In Situ...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-11-16

... DEPARTMENT OF THE INTERIOR Bureau of Land Management [LLWYR05000.L51100000.GN0000.LVEMK10CW370-WYW... Recovery Uranium Project, Fremont and Natrona Counties, WY AGENCY: Bureau of Land Management, Interior... (NEPA), the Bureau of Land Management (BLM) has prepared a Draft Environmental Impact Statement (EIS...

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)

40 CFR 421.326 - Pretreatment standards for new sources.

Code of Federal Regulations, 2014 CFR

2014-07-01

... GUIDELINES AND STANDARDS NONFERROUS METALS MANUFACTURING POINT SOURCE CATEGORY Secondary Uranium Subcategory... wastewater pollutants in secondary uranium process wastewater introduced into a POTW shall not exceed the following values: (a) Refinery sump filtrate. PSNS for the Secondary Uranium Subcategory Pollutant or...

40 CFR 421.326 - Pretreatment standards for new sources.

Code of Federal Regulations, 2012 CFR

2012-07-01

... GUIDELINES AND STANDARDS NONFERROUS METALS MANUFACTURING POINT SOURCE CATEGORY Secondary Uranium Subcategory... wastewater pollutants in secondary uranium process wastewater introduced into a POTW shall not exceed the following values: (a) Refinery sump filtrate. PSNS for the Secondary Uranium Subcategory Pollutant or...

40 CFR 421.326 - Pretreatment standards for new sources.

Code of Federal Regulations, 2013 CFR

2013-07-01

... GUIDELINES AND STANDARDS NONFERROUS METALS MANUFACTURING POINT SOURCE CATEGORY Secondary Uranium Subcategory... wastewater pollutants in secondary uranium process wastewater introduced into a POTW shall not exceed the following values: (a) Refinery sump filtrate. PSNS for the Secondary Uranium Subcategory Pollutant or...

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.

Multistage bioassociation of uranium onto an extremely halophilic archaeon revealed by a unique combination of spectroscopic and microscopic techniques

DOE PAGES

Bader, Miriam; Müller, Katharina; Foerstendorf, Harald; ...

2016-12-27

The interactions of two extremely halophilic archaea with uranium were investigated in this paper at high ionic strength as a function of time, pH and uranium concentration. Halobacterium noricense DSM-15987 and Halobacterium sp. putatively noricense, isolated from the Waste Isolation Pilot Plant repository, were used for these investigations. The kinetics of U(VI) bioassociation with both strains showed an atypical multistage behavior, meaning that after an initial phase of U(VI) sorption, an unexpected interim period of U(VI) release was observed, followed by a slow reassociation of uranium with the cells. By applying in situ attenuated total reflection Fourier-transform infrared spectroscopy, themore » involvement of phosphoryl and carboxylate groups in U(VI) complexation during the first biosorption phase was shown. Differences in cell morphology and uranium localization become visible at different stages of the bioassociation process, as shown with scanning electron microscopy in combination with energy dispersive X-ray spectroscopy. Finally, our results demonstrate for the first time that association of uranium with the extremely halophilic archaeon is a multistage process, beginning with sorption and followed by another process, probably biomineralization.« less

Multistage bioassociation of uranium onto an extremely halophilic archaeon revealed by a unique combination of spectroscopic and microscopic techniques

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

Bader, Miriam; Müller, Katharina; Foerstendorf, Harald

The interactions of two extremely halophilic archaea with uranium were investigated in this paper at high ionic strength as a function of time, pH and uranium concentration. Halobacterium noricense DSM-15987 and Halobacterium sp. putatively noricense, isolated from the Waste Isolation Pilot Plant repository, were used for these investigations. The kinetics of U(VI) bioassociation with both strains showed an atypical multistage behavior, meaning that after an initial phase of U(VI) sorption, an unexpected interim period of U(VI) release was observed, followed by a slow reassociation of uranium with the cells. By applying in situ attenuated total reflection Fourier-transform infrared spectroscopy, themore » involvement of phosphoryl and carboxylate groups in U(VI) complexation during the first biosorption phase was shown. Differences in cell morphology and uranium localization become visible at different stages of the bioassociation process, as shown with scanning electron microscopy in combination with energy dispersive X-ray spectroscopy. Finally, our results demonstrate for the first time that association of uranium with the extremely halophilic archaeon is a multistage process, beginning with sorption and followed by another process, probably biomineralization.« less

Multistage bioassociation of uranium onto an extremely halophilic archaeon revealed by a unique combination of spectroscopic and microscopic techniques.

PubMed

Bader, Miriam; Müller, Katharina; Foerstendorf, Harald; Drobot, Björn; Schmidt, Matthias; Musat, Niculina; Swanson, Juliet S; Reed, Donald T; Stumpf, Thorsten; Cherkouk, Andrea

2017-04-05

The interactions of two extremely halophilic archaea with uranium were investigated at high ionic strength as a function of time, pH and uranium concentration. Halobacterium noricense DSM-15987 and Halobacterium sp. putatively noricense, isolated from the Waste Isolation Pilot Plant repository, were used for these investigations. The kinetics of U(VI) bioassociation with both strains showed an atypical multistage behavior, meaning that after an initial phase of U(VI) sorption, an unexpected interim period of U(VI) release was observed, followed by a slow reassociation of uranium with the cells. By applying in situ attenuated total reflection Fourier-transform infrared spectroscopy, the involvement of phosphoryl and carboxylate groups in U(VI) complexation during the first biosorption phase was shown. Differences in cell morphology and uranium localization become visible at different stages of the bioassociation process, as shown with scanning electron microscopy in combination with energy dispersive X-ray spectroscopy. Our results demonstrate for the first time that association of uranium with the extremely halophilic archaeon is a multistage process, beginning with sorption and followed by another process, probably biomineralization. Copyright © 2016. Published by Elsevier B.V.

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)

Weathering and evaporation controls on dissolved uranium concentrations in groundwater - A case study from northern Burundi.

PubMed

Post, V E A; Vassolo, S I; Tiberghien, C; Baranyikwa, D; Miburo, D

2017-12-31

The potential use of groundwater for potable water supply can be severely compromised by natural contaminants such as uranium. The environmental mobility of uranium depends on a suite of factors including aquifer lithology, redox conditions, complexing agents, and hydrological processes. Uranium concentrations of up to 734μg/L are found in groundwater in northern Burundi, and the objective of the present study was to identify the causes for these elevated concentrations. Based on a comprehensive data set of groundwater chemistry, geology, and hydrological measurements, it was found that the highest dissolved uranium concentrations in groundwater occur near the shores of Lake Tshohoha South and other smaller lakes nearby. A model is proposed in which weathering and evapotranspiration during groundwater recharge, flow and discharge exert the dominant controls on the groundwater chemical composition. Results of PHREEQC simulations quantitatively confirm this conceptual model and show that uranium mobilization followed by evapo-concentration is the most likely explanation for the high dissolved uranium concentrations observed. The uranium source is the granitic sand, which was found to have a mean elemental uranium content of 14ppm, but the exact mobilization process could not be established. Uranium concentrations may further be controlled by adsorption, especially where calcium-uranyl‑carbonate complexes are present. Water and uranium mass balance calculations for Lake Tshohoha South are consistent with the inferred fluxes and show that high‑uranium groundwater represents only a minor fraction of the overall water input to the lake. These findings highlight that the evaporation effects that cause radionuclide concentrations to rise to harmful levels in groundwater discharge areas are not only confined to arid regions, and that this should be considered when selecting suitable locations for water supply wells. Copyright © 2017 Elsevier B.V. All rights reserved.

10 CFR 765.21 - Procedures for processing reimbursement claims.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 4 2011-01-01 2011-01-01 false Procedures for processing reimbursement claims. 765.21 Section 765.21 Energy DEPARTMENT OF ENERGY REIMBURSEMENT FOR COSTS OF REMEDIAL ACTION AT ACTIVE URANIUM... uranium or thorium processing site licensees for approved costs of remedial action will be made...

10 CFR 765.21 - Procedures for processing reimbursement claims.

Code of Federal Regulations, 2012 CFR

2012-01-01

... 10 Energy 4 2012-01-01 2012-01-01 false Procedures for processing reimbursement claims. 765.21 Section 765.21 Energy DEPARTMENT OF ENERGY REIMBURSEMENT FOR COSTS OF REMEDIAL ACTION AT ACTIVE URANIUM... uranium or thorium processing site licensees for approved costs of remedial action will be made...

10 CFR 765.21 - Procedures for processing reimbursement claims.

Code of Federal Regulations, 2014 CFR

2014-01-01

... 10 Energy 4 2014-01-01 2014-01-01 false Procedures for processing reimbursement claims. 765.21 Section 765.21 Energy DEPARTMENT OF ENERGY REIMBURSEMENT FOR COSTS OF REMEDIAL ACTION AT ACTIVE URANIUM... uranium or thorium processing site licensees for approved costs of remedial action will be made...

Characteristics and adaptability of iron- and sulfur-oxidizing microorganisms used for the recovery of metals from minerals and their concentrates

PubMed Central

Rawlings, Douglas E

2005-01-01

Microorganisms are used in large-scale heap or tank aeration processes for the commercial extraction of a variety of metals from their ores or concentrates. These include copper, cobalt, gold and, in the past, uranium. The metal solubilization processes are considered to be largely chemical with the microorganisms providing the chemicals and the space (exopolysaccharide layer) where the mineral dissolution reactions occur. Temperatures at which these processes are carried out can vary from ambient to 80°C and the types of organisms present depends to a large extent on the process temperature used. Irrespective of the operation temperature, biomining microbes have several characteristics in common. One shared characteristic is their ability to produce the ferric iron and sulfuric acid required to degrade the mineral and facilitate metal recovery. Other characteristics are their ability to grow autotrophically, their acid-tolerance and their inherent metal resistance or ability to acquire metal resistance. Although the microorganisms that drive the process have the above properties in common, biomining microbes usually occur in consortia in which cross-feeding may occur such that a combination of microbes including some with heterotrophic tendencies may contribute to the efficiency of the process. The remarkable adaptability of these organisms is assisted by several of the processes being continuous-flow systems that enable the continual selection of microorganisms that are more efficient at mineral degradation. Adaptability is also assisted by the processes being open and non-sterile thereby permitting new organisms to enter. This openness allows for the possibility of new genes that improve cell fitness to be selected from the horizontal gene pool. Characteristics that biomining microorganisms have in common and examples of their remarkable adaptability are described. PMID:15877814

Characteristics and adaptability of iron- and sulfur-oxidizing microorganisms used for the recovery of metals from minerals and their concentrates.

PubMed

Rawlings, Douglas E

2005-05-06

Microorganisms are used in large-scale heap or tank aeration processes for the commercial extraction of a variety of metals from their ores or concentrates. These include copper, cobalt, gold and, in the past, uranium. The metal solubilization processes are considered to be largely chemical with the microorganisms providing the chemicals and the space (exopolysaccharide layer) where the mineral dissolution reactions occur. Temperatures at which these processes are carried out can vary from ambient to 80 degrees C and the types of organisms present depends to a large extent on the process temperature used. Irrespective of the operation temperature, biomining microbes have several characteristics in common. One shared characteristic is their ability to produce the ferric iron and sulfuric acid required to degrade the mineral and facilitate metal recovery. Other characteristics are their ability to grow autotrophically, their acid-tolerance and their inherent metal resistance or ability to acquire metal resistance. Although the microorganisms that drive the process have the above properties in common, biomining microbes usually occur in consortia in which cross-feeding may occur such that a combination of microbes including some with heterotrophic tendencies may contribute to the efficiency of the process. The remarkable adaptability of these organisms is assisted by several of the processes being continuous-flow systems that enable the continual selection of microorganisms that are more efficient at mineral degradation. Adaptability is also assisted by the processes being open and non-sterile thereby permitting new organisms to enter. This openness allows for the possibility of new genes that improve cell fitness to be selected from the horizontal gene pool. Characteristics that biomining microorganisms have in common and examples of their remarkable adaptability are described.

Development of an extractive spectrophotometric method for estimation of uranium in ore leach solutions using 2-ethylhexyl phosphonic acid-mono-2-ethylhexyl ester (PC88A) and tri-n-octyl phosphine oxide (TOPO) mixture as extractant and 2-(5-bromo-2-pyridylozo)-5-diethyl aminophenol (Br-PADAP) as chromophore

NASA Astrophysics Data System (ADS)

Biswas, Sujoy; Pathak, P. N.; Roy, S. B.

2012-06-01

An extractive spectrophotometric analytical method has been developed for the determination of uranium in ore leach solution. This technique is based on the selective extraction of uranium from multielement system using a synergistic mixture of 2-ethylhexyl phosphonic acid-mono-2-ethylhexyl ester (PC88A) and tri-n-octyl phosphine oxide (TOPO) in cyclohexane and color development from the organic phase aliquot using 2-(5-Bromo-2-pyridylazo)-5-diethyl aminophenol (Br-PADAP) as chromogenic reagent. The absorption maximum (λmax) for UO22+-Br-PADAP complex in organic phase samples, in 64% (v/v) ethanol containing buffer solution (pH 7.8) and 1,2-cyclohexylenedinitrilotetraacetic acid (CyDTA) complexing agent, has been found to be at 576 nm (molar extinction coefficient, ɛ: 36,750 ± 240 L mol-1 cm-1). Effects of various parameters like stability of complex, ethanol volume, ore matrix, interfering ions etc. on the determination of uranium have also been evaluated. Absorbance measurements as a function of time showed that colored complex is stable up to >24 h. Presence of increased amount of ethanol in colored solution suppresses the absorption of a standard UO22+-Br-PADAP solution. Analyses of synthetic standard as well as ore leach a solution show that for 10 determination relative standard deviation (RSD) is <2%. The accuracy of the developed method has been checked by determining uranium using standard addition method and was found to be accurate with a 98-105% recovery rate. The developed method has been applied for the analysis of a number of uranium samples generated from uranium ore leach solutions and results were compared with standard methods like inductively coupled plasma emission spectrometry (ICPAES). The determined values of uranium concentrations by these methods are within ±2%. This method can be used to determine 2.5-250 μg mL-1 uranium in ore leach solutions with high accuracy and precision.

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

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)

Heat-induced redistribution of surface oxide in uranium

NASA Astrophysics Data System (ADS)

Swissa, Eli; Shamir, Noah; Mintz, Moshe H.; Bloch, Joseph

1990-09-01

The redistribution of oxygen and uranium metal at the vicinity of the metal-oxide interface of native and grown oxides due to vacuum thermal annealing was studied for uranium and uranium-chromium alloy using Auger depth profiling and metallographic techniques. It was found that uranium metal is segregating out through the uranium oxide layer for annealing temperatures above 450°C. At the same time the oxide is redistributed in the metal below the oxide-metal interface in a diffusion like process. By applying a diffusion equation of a finite source, the diffusion coefficients for the process were obtained from the oxygen depth profiles measured for different annealing times. An Arrhenius like behavior was found for the diffusion coefficient between 400 and 800°C. The activation energy obtained was Ea = 15.4 ± 1.9 kcal/mole and the pre-exponential factor, D0 = 1.1 × 10 -8cm2/ s. An internal oxidation mechanism is proposed to explain the results.

Absorption of Thermal Neutrons in Uranium

DOE R&D Accomplishments Database

Creutz, E. C.; Wilson, R. R.; Wigner, E. P.

1941-09-26

A knowledge of the absorption processes for neutrons in uranium is important for planning a chain reaction experiment. The absorption of thermal neutrons in uranium and uranium oxide has been studied. Neutrons from the cyclotron were slowed down by passage through a graphite block. A uranium or uranium oxide sphere was placed at various positions in the block. The neutron intensity at different points in the sphere and in the graphite was measured by observing the activity induced in detectors or uranium oxide or manganese. It was found that both the fission activity in the uranium oxide and the activity induced in manganese was affected by non-thermal neutrons. An experimental correction for such effects was made by making measurements with the detectors surrounded by cadmium. After such corrections the results from three methods of procedure with the uranium oxide detectors and from the manganese detectors were consistent to within a few per cent.

Process for producing an aggregate suitable for inclusion into a radiation shielding product

DOEpatents

Lessing, Paul A.; Kong, Peter C.

2000-01-01

The present invention is directed to methods for converting depleted uranium hexafluoride to a stable depleted uranium silicide in a one-step reaction. Uranium silicide provides a stable aggregate material that can be added to concrete to increase the density of the concrete and, consequently, shield gamma radiation. As used herein, the term "uranium silicide" is defined as a compound generically having the formula U.sub.x Si.sub.y, wherein the x represents the molecules of uranium and the y represent the molecules of silicon. In accordance with the present invention, uranium hexafluoride is converted to a uranium silicide by contacting the uranium hexafluoride with a silicon-containing material at a temperature in a range between about 1450.degree. C. and about 1750.degree. C. The stable depleted uranium silicide is included as an aggregate in a radiation shielding product, such as a concrete product.

Uranium lines in the spectra of peculiar A stars - A search for recent r-process events

NASA Technical Reports Server (NTRS)

Cowley, C. R.; Adelman, S. J.

1975-01-01

Uranium wavelengths in the spectra of Ap stars are studied to see if they give any indication of a recent r-process event. It is concluded that there is no credible evidence for an admixture of uranium-235 in these stars, which would imply such an event. The evidence, though negative, is badly confused by blending of lines, and a final judgement must wait for an observational clarification of the situation.

Quarterly Report 24.1.3.1

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

Shusterman, Jennifer A.

We are measuring freshly separated uranium samples using modern list mode (event-by-event) electronics with high resolution HPGe detectors to study the in-growth behaviors of uranium daughters’ gamma-rays. These data will show how we can use gamma-ray spectroscopy to determine the separation date for processed uranium. With this knowledge, one can obtain proper uranium isotope ratios using standard safeguards accountability software such as U-235 or MGAU.

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

Biosorption and biomineralization of uranium(VI) by Saccharomyces cerevisiae-Crystal formation of chernikovite.

PubMed

Zheng, Xin-Yan; Wang, Xiao-Yu; Shen, Yang-Hao; Lu, Xia; Wang, Tie-Shan

2017-05-01

Biosorption of heavy metal elements including radionuclides by microorganisms is a promising and effective method for the remediation of the contaminated places. The responses of live Saccharomyces cerevisiae in the toxic uranium solutions during the biosorption process and the mechanism of uranium biomineralization by cells were investigated in the present study. A novel experimental phenomenon that uranium concentrations have negative correlation with pH values and positive correlation with phosphate concentrations in the supernatant was observed, indicating that hydrogen ions, phosphate ions and uranyl ions were involved in the chernikovite precipitation actively. During the biosorption process, live cells desorb deposited uranium within the equilibrium state of biosorption system was reached and the phosphorus concentration increased gradually in the supernatant. These metabolic detoxification behaviours could significantly alleviate uranium toxicity and protect the survival of the cells better in the environment. The results of microscopic and spectroscopic analysis demonstrated that the precipitate on the cell surface was a type of uranium-phosphate compound in the form of a scale-like substance, and S. cerevisiae could transform the uranium precipitate into crystalline state-tetragonal chernikovite [H 2 (UO 2 ) 2 (PO 4 ) 2 ·8H 2 O]. Copyright © 2017 Elsevier Ltd. All rights reserved.

Equilibrium, kinetic and thermodynamic studies of uranium biosorption by calcium alginate beads.

PubMed

Bai, Jing; Fan, Fangli; Wu, Xiaolei; Tian, Wei; Zhao, Liang; Yin, Xiaojie; Fan, Fuyou; Li, Zhan; Tian, Longlong; Wang, Yang; Qin, Zhi; Guo, Junsheng

2013-12-01

Calcium alginate beads are potential biosorbent for radionuclides removal as they contain carboxyl groups. However, until now limited information is available concerning the uptake behavior of uranium by this polymer gel, especially when sorption equilibrium, kinetics and thermodynamics are concerned. In present work, batch experiments were carried out to study the equilibrium, kinetics and thermodynamics of uranium sorption by calcium alginate beads. The effects of initial solution pH, sorbent amount, initial uranium concentration and temperature on uranium sorption were also investigated. The determined optimal conditions were: initial solution pH of 3.0, added sorbent amount of 40 mg, and uranium sorption capacity increased with increasing initial uranium concentration and temperature. Equilibrium data obtained under different temperatures were fitted better with Langmuir model than Freundlich model, uranium sorption was dominated by a monolayer way. The kinetic data can be well depicted by the pseudo-second-order kinetic model. The activation energy derived from Arrhenius equation was 30.0 kJ/mol and the sorption process had a chemical nature. Thermodynamic constants such as ΔH(0), ΔS(0) and ΔG(0) were also evaluated, results of thermodynamic study showed that the sorption process was endothermic and spontaneous. Copyright © 2013 Elsevier Ltd. All rights reserved.

40 CFR 421.324 - Standards of performance for new sources.

Code of Federal Regulations, 2013 CFR

2013-07-01

...) EFFLUENT GUIDELINES AND STANDARDS NONFERROUS METALS MANUFACTURING POINT SOURCE CATEGORY Secondary Uranium... Uranium Subcategory Pollutant or pollutant property Maximum for any 1 day Maximum for monthly average mg/kg (pounds per million pounds) of uranium processed in the refinery Chromium (total) 27.14 11.00...

40 CFR 421.324 - Standards of performance for new sources.

Code of Federal Regulations, 2014 CFR

2014-07-01

...) EFFLUENT GUIDELINES AND STANDARDS NONFERROUS METALS MANUFACTURING POINT SOURCE CATEGORY Secondary Uranium... Uranium Subcategory Pollutant or pollutant property Maximum for any 1 day Maximum for monthly average mg/kg (pounds per million pounds) of uranium processed in the refinery Chromium (total) 27.14 11.00...

40 CFR 421.324 - Standards of performance for new sources.

Code of Federal Regulations, 2012 CFR

2012-07-01

...) EFFLUENT GUIDELINES AND STANDARDS NONFERROUS METALS MANUFACTURING POINT SOURCE CATEGORY Secondary Uranium... Uranium Subcategory Pollutant or pollutant property Maximum for any 1 day Maximum for monthly average mg/kg (pounds per million pounds) of uranium processed in the refinery Chromium (total) 27.14 11.00...

METHOD OF SEPARATING URANIUM SUSPENSIONS

DOEpatents

Wigner, E.P.; McAdams, W.A.

1958-08-26

A process is presented for separating colloidally dissed uranium oxides from the heavy water medium in upwhich they are contained. The method consists in treating such dispersions with hydrogen peroxide, thereby converting the uranium to non-colloidal UO/sub 4/, and separating the UO/sub 4/ sfter its rapid settling.

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.

The adsorption of methyl iodide on uranium and uranium dioxide: Surface characterization using X-ray photoelectron spectroscopy (XPS) and Auger electron spectroscopy (AES)

NASA Astrophysics Data System (ADS)

Dillard, J. G.; Moers, H.; Klewe-Nebenius, H.; Kirch, G.; Pfennig, G.; Ache, H. J.

1984-09-01

The adsorption of methyl iodide on uranium and on uranium dioxide has been studied at 25 °C. Surfaces of the substrates were characterized before and after adsorption by X-ray photoelectron spectroscopy (XPS) and Auger electron spectroscopy (AES). The XPS binding energy results indicate that CH 3I adsorption on uranium yields a carbide-type carbon, UC, and uranium iodide, UI 3. On uranium dioxide the carbon electron binding energy measurements are consistent with the formation of a hydrocarbon, —CH 3-type moiety. The interpretation of XPS and AES spectral features for CH 3I adsorption on uranium suggest that a complex dissociative adsorption reaction takes place. Adsorption of CH 3I on UO 2 occurs via a dissociative process. Saturation coverage occurs on uranium at approximately two langmuir (1 L = 10 -6 Torr s) exposure whereas saturation coverage on uranium dioxide is found at about five langmuir.

Uranium provinces of North America; their definition, distribution, and models

USGS Publications Warehouse

Finch, Warren Irvin

1996-01-01

Uranium resources in North America are principally in unconformity-related, quartz-pebble conglomerate, sandstone, volcanic, and phosphorite types of uranium deposits. Most are concentrated in separate, well-defined metallogenic provinces. Proterozoic quartz-pebble conglomerate and unconformity-related deposits are, respectively, in the Blind River–Elliot Lake (BRELUP) and the Athabasca Basin (ABUP) Uranium Provinces in Canada. Sandstone uranium deposits are of two principal subtypes, tabular and roll-front. Tabular sandstone uranium deposits are mainly in upper Paleozoic and Mesozoic rocks in the Colorado Plateau Uranium Province (CPUP). Roll-front sandstone uranium deposits are in Tertiary rocks of the Rocky Mountain and Intermontane Basins Uranium Province (RMIBUP), and in a narrow belt of Tertiary rocks that form the Gulf Coastal Uranium Province (GCUP) in south Texas and adjacent Mexico. Volcanic uranium deposits are concentrated in the Basin and Range Uranium Province (BRUP) stretching from the McDermitt caldera at the Oregon-Nevada border through the Marysvale district of Utah and Date Creek Basin in Arizona and south into the Sierra de Peña Blanca District, Chihuahua, Mexico. Uraniferous phosphorite occurs in Tertiary sediments in Florida, Georgia, and North and South Carolina and in the Lower Permian Phosphoria Formation in Idaho and adjacent States, but only in Florida has economic recovery been successful. The Florida Phosphorite Uranium Province (FPUP) has yielded large quantities of uranium as a byproduct of the production of phosphoric acid fertilizer. Economically recoverable quantities of copper, gold, molybdenum, nickel, silver, thorium, and vanadium occur with the uranium deposits in some provinces.Many major epochs of uranium mineralization occurred in North America. In the BRELUP, uranium minerals were concentrated in placers during the Early Proterozoic (2,500–2,250 Ma). In the ABUP, the unconformity-related deposits were most likely formed initially by hot saline formational water related to diagenesis (»1,400 to 1,330 Ma) and later reconcentrated by hydrothermal events at »1,280–»1,000, »575, and »225 Ma. Subsequently in North America, only minor uranium mineralization occurred until after continental collision in Permian time (255 Ma). Three principal epochs of uranium mineralization occurred in the CPUP: (1) » 210–200 Ma, shortly after Late Triassic sedimentation; (2) »155–150 Ma, in Late Jurassic time; and (3) » 135 Ma, after sedimentation of the Upper Jurassic Morrison Formation. The most likely source of the uranium was silicic volcaniclastics for the three epochs derived from a volcanic island arc at the west edge of the North American continent. Uranium mineralization occurred during Eocene, Miocene, and Pliocene times in the RMIBUP, GCUP, and BRUP. Volcanic activity took place near the west edge of the continent during and shortly after sedimentation of the host rocks in these three provinces. Some volcanic centers in the Sierra de Peña Blanca district within the BRUP may have provided uranium-rich ash to host rocks in the GCUP.Most of the uranium provinces in North America appear to have a common theme of close associations to volcanic activity related to the development of the western margin of the North American plate. The south and west margin of the Canadian Shield formed the leading edge of the progress of uranium source development and mineralization from the Proterozoic to the present. The development of favorable hosts and sources of uranium is related to various tectonic elements developed over time. Periods of major uranium mineralization in North America were Early Proterozoic, Middle Proterozoic, Late Triassic–Early Jurassic, Early Cretaceous, Oligocene, and Miocene. Tertiary mineralization was the most pervasive, covering most of Western and Southern North America. 

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.

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.

Process for Transition of Uranium Mill Tailings Radiation Control Act Title II Disposal Sites to the U.S. Department of Energy Office of Legacy Management for Long-Term Surveillance and Maintenance

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

None

This document presents guidance for implementing the process that the U.S. Department of Energy (DOE) Office of Legacy Management (LM) will use for assuming perpetual responsibility for a closed uranium mill tailings site. The transition process specifically addresses sites regulated under Title II of the Uranium Mill Tailings Radiation Control Act (UMTRCA) but is applicable in principle to the transition of sites under other regulatory structures, such as the Formerly Utilized Sites Remedial Action Program.

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.

EXTRACTION OF URANIUM

DOEpatents

Kesler, R.D.; Rabb, D.D.

1959-07-28

An improved process is presented for recovering uranium from a carnotite ore. In the improved process U/sub 2/O/sub 5/ is added to the comminuted ore along with the usual amount of NaCl prior to roasting. The amount of U/sub 2/O/ sub 5/ is dependent on the amount of free calcium oxide and the uranium in the ore. Specifically, the desirable amount of U/sub 2/O/sub 5/ is 3.2% for each 1% of CaO, and 5 to 6% for each 1% of uranium. The mixture is roasted at about 1560 deg C for about 30 min and then leached with a 3 to 9% aqueous solution of sodium carbonate.